[Federal Register Volume 72, Number 166 (Tuesday, August 28, 2007)]
[Notices]
[Pages 49568-49586]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-16766]
[[Page 49567]]
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Part III
Nuclear Regulatory Commission
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Biweekly Notice; Applications and Amendments to Facility Operating
Licenses Involving No Significant Hazards Considerations; Notice
Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 /
Notices
[[Page 49568]]
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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 August 2, 2007, to August 15, 2007. The last
biweekly notice was published on August 14, 2007 (72 FR 45454).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the basis 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
[[Page 49569]]
fact. Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, PA
Date of amendment request: June 29, 2007.
Description of amendment request: The proposed license amendment
would revise the TMI-1 Technical Specifications 3.3.1.3, 3.3.2.1 and
4.1, to reflect a change to the Reactor Building spray system buffering
agent from sodium hydroxide to trisodium phosphate dodecahydrate. This
proposed change is designed to minimize the potential for exacerbating
sump screen blockage under post loss of coolant event conditions by
limiting potential adverse chemical interactions between the buffering
agent and certain insulation materials used in the TMI-1 containment.
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.
For the proposed change, trisodium phosphate dodecahydrate (TSP)
will be used as a buffer for post-accident pH control and will
replace the existing buffer. The buffer material and means of
storage and delivery are not initiators for previously analyzed
accidents. The accident mitigation function of the replacement
buffer is the same as the existing buffer. The pH of the water in
the emergency sump following a loss of coolant accident (LOCA) will
be adjusted with TSP rather than sodium hydroxide (NaOH) to be
within a range that will reduce the potential for elemental iodine
re-evolution and long term stress corrosion during the recirculation
mode of emergency core cooling system (ECCS) operation. In addition,
the replacement buffer will reduce the formation of precipitates
resulting from chemical reactions between the recirculating spray
solution and insulating materials in the Reactor Building (RB), thus
reducing the potential for ECCS emergency sump intake screen
blockage. The proposed sump pH range will not result in an increase
in post-LOCA hydrogen generation. The proposed isolation of the
sodium hydroxide tank, and the installation of TSP in baskets has
been evaluated for impacts on accident effects and the safety
functions of required systems, structures, and components (SSCs).
The RB emergency sump solution pH profile resulting from the
proposed change has been evaluated for impacts on environmental
qualification of SSCs. The accident mitigation functions of required
SSCs will not be affected by the proposed change.
As a part of the proposed change, the radiological consequences
of a postulated LOCA have been reanalyzed using Standard Review Plan
(SRP) 6.5.2, ``Containment Spray as a Fission Product Cleanup
System,'' and the Alternate Source Term (AST) guidance in Regulatory
Guide 1.183. The analysis considered the use of a plain borated
water spray during the post-LOCA injection phase and a spray mixture
with a minimum pH of 7.3 during the recirculation phase. The results
of the reanalysis show that the consequences of the accident are not
increased. The calculated doses at the Exclusion Area Boundary, Low
Population Zone boundary, and in the Control Room remain within 10
CFR 50.67 AST dose limits.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change will replace the existing spray additive
design using sodium hydroxide solution stored in a tank with TSP
contained in baskets located on the floor of the RB. The TSP storage
and delivery method is passive. The baskets are constructed of
stainless steel to resist corrosion and are seismically qualified.
The existing sodium hydroxide tank, associated piping, and valves
will no longer be used and will be permanently isolated, but their
structural integrity will be maintained. The RB spray system will
perform the same function and operate in the same manner for the
proposed change; however, the sodium hydroxide tank isolation valves
will no longer be required to open on an engineered safeguards
actuation
[[Page 49570]]
signal. The accident mitigation function of TSP will be the same as
the existing buffer, sodium hydroxide. The TSP will act as a
buffering agent to raise the pH of the water in the containment
emergency sump to greater than 7.3 for long-term post-LOCA RB spray
recirculation. The SSCs required for post-LOCA accident mitigation
have been evaluated for the proposed change including the effects of
the modified emergency sump solution pH profile. No new accident
scenarios, failure mechanisms, or single failures are introduced as
a result of the proposed change.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change from sodium hydroxide to TSP will not reduce
the effectiveness of the post-LOCA pH control buffer. The TSP will
buffer the sump water sufficiently to assure that the resulting
mixture pH is > 7.3 and < 8.0. This pH level will be effective in
reducing the potential for iodine re-evolution during the
recirculation phase of a LOCA, preventing long-term stress corrosion
cracking of austenitic stainless steel, and minimizing post-LOCA
hydrogen generation. In addition, the use of TSP will reduce the
formation of precipitates resulting from chemical reactions between
the recirculating spray solution and insulating materials in the RB,
thus reducing the potential for ECCS emergency sump intake screen
blockage. The proposed use of SRP 6.5.2 guidance, which is an NRC-
approved methodology, for post-LOCA dose calculations does not
result in a reduction in a margin of safety. The proposed change
does not adversely affect the performance of SSCs required for post-
LOCA mitigation, and does not affect an operating parameter or
setpoint used in the accident analyses to establish a margin of
safety. Also, the proposed change does not affect a margin of safety
associated with containment functional performance.
Therefore, the proposed change does not involve a significant
reduction in any 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. Bradley Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, SC
Date of amendment request: July 17, 2007.
Description of amendment request: A change is proposed to the
standard technical specifications (STS) (NUREGs 1430 through 1434) and
plant specific technical specifications (TS), to strengthen TS
requirements regarding control room envelope (CRE) habitability by
changing the action and surveillance requirements associated with the
limiting condition for operation operability requirements for the CRE
emergency ventilation system, and by adding a new TS administrative
controls program on CRE habitability. Accompanying the proposed TS
change are appropriate conforming technical changes to the TS Bases.
The proposed revision to the Bases also includes editorial and
administrative changes to reflect applicable changes to the
corresponding STS Bases, which were made to improve clarity, conform
with the latest information and references, correct factual errors, and
achieve more consistency among the STS NUREGs. The proposed revision to
the TS and associated Bases is consistent with STS as revised by TSTF-
448, Revision 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, WI
Date of amendment request: June 12, 2007.
[[Page 49571]]
Description of amendment request: The proposed amendment would
revise the nuclear instrumentation system permissive setpoints in
Technical Specification (TS) Table 3.5-2, ``Instrument Operation
Conditions for Reactor Trip,'' revise the Table format, and revise TS
2.3, ``Instrumentation System,'' to make consistent with other proposed
changes to the TS.
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 amendment does not change the probability or
consequences of any previously evaluated accidents in the KPS
[Kewaunee Power Station] updated safety analysis report (USAR). The
proposed amendment would modify the TS setpoint values for the P-7
and P-10 permissives. The actual plant settings will continue to be
approximately 10% of rated reactor power. The reactor protection
system (RPS) is designed to monitor various plant parameters and
initiate a reactor trip in the event these parameters are outside
predetermined limits. The RPS is not an accident initiator and
therefore, changing the setpoints for these permissives will not
increase the probability of an accident previously evaluated.
The proposed amendment would add a setpoint band to the current
TS required settings for permissive P-7 and P-10 to accommodate
proper setting of the permissives. The only previously evaluated
accident that is potentially affected by the proposed changes is the
Uncontrolled Rod Cluster Assembly Rod Withdrawal At-Power (RWAP)
accident analysis. The effects of these setpoint changes have been
evaluated and determined not to have a significant effect on the
consequences of the RWAP accident analysis results. The acceptance
criteria for the RWAP accident analysis continue to be met.
Therefore the proposed changes would not increase the consequences
of an accident previously evaluated.
Therefore the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment modifies the TS setpoint values for
permissives P-7 and P-10. The actual plant settings will continue to
be approximately 10% power. The proposed changes affect the power
level at which RPS trip functions are enabled or blocked to ensure
proper operation of the RPS. The changes do not add any new systems,
structures or components (SSCs) or physically modify any existing
SSCs with the possibility of creating a new accident.
The proposed amendment does not functionally affect the
operation of any SSC important to safety or its ability to perform
its design function. Additionally, the proposed amendment does not
create the possibility of a new or different kind of accident due to
credible new failure mechanisms, malfunctions, or accident
initiators not considered in the design and licensing bases.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment would add a setpoint band to the current
TS required settings for permissivies P-7 and P-10 to accommodate
proper setting of the permissives. The safety function of the
nuclear instrumentation system and the affected permissives are not
affected by this proposed change.
The only safety analysis in the KPS USAR potentially affected by
these proposed changes is the Uncontrolled Rod Cluster Assembly Rod
Withdrawal At-Power (RWAP) event analysis. Evaluation of the RWAP
event analysis results demonstrated that the RWAP would not have a
significant effect on a margin of safety.
The effects of the proposed change have been evaluated and all
safety analysis acceptance criteria will continue to be met.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Acting Branch Chief: Travis L. Tate.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station (KPS), Kewaunee County, Wisconsin
Date of amendment request: July 2, 2007.
Description of amendment request: The proposed amendment would
delete operating license (OL) condition 2.C (5), ``Fuel Burnup,'' which
restricts maximum rod average burnup to 60 giga-watt days per metric
ton uranium (GWD/MTU). Deletion of the OL condition will provide the
opportunity to increase maximum rod average burnup to as high as 62
GWD/MTU and allow fuel management flexibility.
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.
Deletion of KPS OL condition 2.C (5) does not add, delete, or
modify any KPS systems, structures, or components (SSCs). The
proposed amendment would effectively allow future increases in the
KPS maximum rod average burnup limit using currently existing fuel
management methods and models that have been reviewed and approved
by the NRC [Nuclear Regulatory Commission].
Maximum average rod burnup limits will continue to be maintained
within safe and acceptable limits using these fuel management
methods and models. Nuclear fuel is the only plant component
potentially affected by increasing the maximum rod average burnup
limit. Increasing the KPS maximum rod average burnup limit does not
affect the thermal hydraulic response or the radiological
consequences of any previously evaluated accident. The fuel rod
design criteria will continue to be met at the maximum burnup limits
allowed under the current fuel management and evaluation processes.
An increase to the maximum rod average burnup limit will not
increase the likelihood of a malfunction of nuclear fuel since the
fuel currently used at KPS has been designed to support a maximum
rod average burnup well in excess of 62 GWD/MTU.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment would delete a KPS OL condition that
limits maximum rod average burnup. The proposed amendment would
effectively allow future increases in the KPS maximum rod average
burnup limit using currently existing fuel management methods and
models that have been reviewed and approved by the NRC. Nuclear fuel
is the only component potentially affected by changes to the maximum
rod average burnup limit. The proposed amendment does not change the
design function of the nuclear fuel or create any credible new
failure mechanisms or malfunctions for nuclear fuel. Fuel rod design
criteria will continue to be met at the maximum burnup limits
allowed under the fuel management methods and models that have been
previously reviewed and approved by the NRC. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
[[Page 49572]]
Response: No.
The proposed amendment deletes a KPS OL condition that limits
maximum rod average burnup. The proposed amendment would effectively
allow future increases in the KPS maximum rod average burnup limit
using currently existing methods and models that have been reviewed
and approved by the NRC. The proposed amendment does not result in
altering or exceeding a design basis or safety limit for the plant.
All current fuel design criteria will continue to be satisfied, and
the safety analysis of record, including evaluations of the
radiological consequences of design basis accidents, will remain
applicable.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Acting Branch Chief: Travis L. Tate.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, WA
Date of amendment request: July 26, 2007, as superseded by letter
dated August 8, 2007.
Description of amendment request: The proposed changes revise the
requirements of Technical Specification (TS) 3.3.5.2, ``Reactor Core
Isolation Cooling (RCIC) System Instrumentation,'' and TS 3.5.2, ``ECCS
[Emergency Core Cooling System]--Shutdown,'' to increase the Condensate
Storage Tank (CST) level.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the 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 operation of Columbia in accordance with the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated. Neither of
these changes affects the probability of any accident previously
evaluated as they do not involve or impact accident initiators.
The proposed change to TS 3.3.5.2 would ensure that the
consequences would remain the same as that previously evaluated for
during any event in which the RCIC pump was utilized. Adequate
volume would be maintained in the CST whenever the RCIC pump was
aligned to it to ensure that it did not experience loss of suction
due to vortexing.
The proposed changes to TS 3.5.2.2 would ensure that the
previously assumed volume of water in the CST would still be
available to inject into the reactor vessel during Modes 4 and 5
should the suppression pool not meet minimum volume requirements.
Therefore, operation of Columbia in accordance with the proposed
amendment will 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 operation of Columbia in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
change will not create a new or different kind of accident since it
only affects the amount of water held in reserve to support reactor
vessel inventory loss. The proposed change does not introduce any
credible mechanisms for unacceptable radiation release nor does it
require physical modification to the plant. The plant has operated
well within the existing allowable values. The increased margin
provided by the increased level will assure no new or different
kinds of accidents result from the proposed change. Therefore, the
operation of Columbia in accordance with the proposed amendment will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The operation of Columbia in accordance with the proposed
amendment will not involve a significant reduction in the margin of
safety. The proposed amendment provides assurance that the RCIC pump
suction will be transferred without loss of suction and that 135,000
gallons of CST inventory will continue to be available for injection
into the RPV [reactor pressure vessel] under worst case conditions.
Therefore, operation of Columbia in accordance with the proposed
amendment will 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.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, WA
Date of amendment request: July 30, 2007.
Description of amendment request: The proposed changes revise
Technical Specifications (TSs) 1.4, ``Frequency,'' 3.1.5, ``Control Rod
Scram Accumulators,'' 3.4.1, ``Recirculation Loops Operating,'' 3.5.1,
``ECCS [Emergency Core Cooling System]--Operating,'' 3.5.2, ``ECCS--
Shutdown,'' 3.7.1, ``Standby Service Water (SW) System and Ultimate
Heat Sink (UHS),'' 3.8.1, ``AC [Alternating Current] Sources--
Operating,'' 3.8.2, ``AC Sources--Shutdown,'' and 5.5.6, ``Inservice
Testing Program.'' The proposed changes include updates to adopt
approved TS Task Force (TSTF) Travelers 284, Revision 3, ``Add `Met'
vs. `Perform' to Specification 1.4, Frequency,'' TSTF-479, Revision 0,
``Changes to Reflect Revision of 10 CFR 50.55a,'' and TSTF-485,
Revision 0, ``Correct Example 1.4-1.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment is administrative in nature and does not
affect analysis inputs or mitigation of analyzed accidents and
transients. Therefore, the proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design. The
proposed change does not introduce any new modes of plant operation
or make any changes to system setpoints. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed amendment is administrative in nature and does not
involve physical changes to plant SSCs [structures, systems, or
components], or the manner in which these SSCs are operated,
maintained, modified, tested, or inspected. The proposed amendment
does not involve a change to any
[[Page 49573]]
safety limit, limiting safety system setting, limiting condition for
operation, or design parameters for any SSC. The only minor
alteration to the plant design basis is relative to the application
of TS 3.4.1. However, as discussed in Section 4 [of the licensee's
submittal], this alteration biases the operation of the plant in the
direction of safety. The proposed amendment does not impact any
safety analysis assumptions and does not involve a change in initial
conditions, system response times, or other parameters affecting any
accident analysis. For these reasons, the proposed amendment 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.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, WA
Date of amendment request: July 30, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) to establish more effective
and appropriate action, surveillance, and administrative TS
requirements related to ensuring the habitability of the control room
envelope (CRE) in accordance with Nuclear Regulatory Commission (NRC)-
approved TS Task Force (TSTF) Standard Technical Specification change
traveler TSTF-448, Revision 3, ``Control Room Habitability.''
Specifically, the proposed amendment would modify TS 3.7.3, ``Control
Room Emergency Filtration (CREF) System,'' and add new TS 5.5.14,
``Control Room Envelope Habitability Program,'' to Section 5.5,
``Programs and Manuals.''
The NRC staff issued a ``Notice of Availability of Technical
Specification Improvement to Modify Requirements Regarding Control Room
Envelope Habitability Using the Consolidated Line Item Improvement
Process'' associated with TSTF-448, Revision 3, in the Federal Register
on January 17, 2007 (72 FR 2022). The notice included a model safety
evaluation, a model no significant hazards consideration (NSHC)
determination, and a model license amendment request. In its
application dated July 30, 2007, the licensee affirmed the
applicability of the model NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE [control room
envelope] emergency ventilation system, which is a mitigation system
designed to minimize unfiltered air leakage into the CRE and to
filter the CRE atmosphere to protect the CRE occupants in the event
of accidents previously analyzed. An important part of the CRE
emergency ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation.
The proposed change does not alter any safety analysis
assumptions and is consistent with current plant operating practice.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based upon this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendment involves NSHC.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Energy Northwest, Docket No.50-397, Columbia Generating Station, Benton
County, WA
Date of amendment request: July 30, 2007.
Description of amendment request: The proposed changes revise
Technical Specifications (TSs) 3.3.3.1, ``Post Accident Monitoring
(PAM) Instrumentation,'' 3.3.6.1, ``Primary Containment Isolation
Instrumentation,'' 3.6.1.3, ``Primary Containment Isolation Valves
(PCIVs),'' and 3.6.4.2, ``Secondary Containment Isolation Valves
(SCIVs).'' The proposed changes adopt the following TS Task Force
(TSTF) Travelers that have been previously approved by the Nuclear
Regulatory Commission (NRC): TSTF-45-A, Revision 2, ``Exempt
Verification of CIVs [containment isolation valves] that are Not
Locked, Sealed or Otherwise Secured,'' TSTF-46-A, Revision 1, ``Clarify
the CIV Surveillance to Apply Only to Automatic Isolation Valves,''
TSTF-207-A, Revision 5, ``Completion Time for Restoration of Various
Excessive Leakage Rates,'' TSTF-269-A, Revision 2, ``Allow
Administrative Means of Position Verification for Locked or Sealed
Valves,'' TSTF-295-A, Revision 0, ``Modify Note 2 to Actions of PAM
Table to Allow Separate Condition Entry for Each Penetration,'' TSTF-
306-A, Revision 2, ``Add Action to LCO
[[Page 49574]]
[limiting condition for operation] 3.3.6.1 to Give Option to Isolate
the Penetration,'' and TSTF-323-A, Revision 0, ``EFCV [excess flow
check valve] Completion Time to 72 Hours.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below. The licensee addressed each
proposed TSTF separately in its analysis:
TSTF-45-A, Revision 2
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change would exempt manual isolation valves and
blind flanges located inside and outside the primary containment and
in the secondary containment that are locked, sealed, or otherwise
secured in position from the periodic verification of valve position
required by SRs [surveillance requirements] 3.6.1.3.2 and 3.6.1.3.3,
and SR 3.6.4.2.1. The exempted valves are verified to be in the
correct position upon being locked, sealed, or secured. Because the
valves are in the condition assumed in the accident analysis, the
proposed change will not affect the initiators or mitigation of any
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change replaces the periodic verification of valve
position with verification of valve position followed by locking,
sealing, or otherwise securing the valve in position. Periodic
verification is also effective in detecting valve mispositioning.
However, verification followed by securing the valve in position is
effective in preventing valve mispositioning.
TSTF-46-A, Revision 1
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change would revise the verification of PCIV and
SCIV closure time to clarify that only power operated, automatic
valves are required to be tested. PCIVs and SCIVs are not an
initiator of any accident previously evaluated; rather, they serve
to mitigate the consequences of evaluated accidents. The proposed
change does not change the requirement to verify that power
operated, automatic PCIVs and SCIVs close within the time assumed in
the accident analysis, but rather, clarifies that non-automatic
valves, which the accident analysis does not assume close within a
specified time, are not required to be tested to verify the closure
time. As a result, the mitigating action of the PCIVs and SCIVs is
not affected by this change.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change would revise the verification of PCIV and
SCIV closure time to clarify that only power operated, automatic
valves are required to be tested, and not all power operated valves.
There is no closure time assumed in the accident analysis for power
operated PCIVs and SCIVs that are not automatic.
TSTF-207-A, Revision 5
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change revises the Actions of TS 3.6.1.3 to make
the presentation consistent with similar Conditions in the ISTS
[Improved Standard TSs]. Part of this change would extend the CT
[completion time] for hydrostatically tested lines on a closed
system to 72 hours for
Condition D. Most of the proposed changes do not affect the
requirements in the TS and have no effect on the initiation or
mitigation of any accident previously evaluated. Leakage of
hydrostatically tested lines on a closed system is not an initiator
of any accident previously evaluated. The consequences of a
previously evaluated accident during the extended CT are the same as
the consequences during the existing CT.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed changes are editorial in nature and do not affect
the requirements of the TS. Extension of the CT for hydrostatically
tested lines on a closed system to 72 hours does not represent a
significant reduction in safety given the reliability of closed
systems. Nonetheless, leakage can be isolated restored by isolating
the penetration with a valve not exceeding the leakage limits.
TSTF-269-A, Revision 2
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The proposed change modifies TS 3.6.1.3 and TS
3.6.4.2. Both TS 3.6.1.3 and TS 3.6.4.2 require penetrations with an
inoperable isolation valve to be isolated and periodically verified
to be isolated. A Note is added to TS 3.6.1.3, Actions A and C, and
TS 3.6.4.2, Action A, to allow isolation devices that are locked,
sealed, or otherwise secured to be verified by use of administrative
means. The proposed change does not affect any plant equipment, test
methods, or plant operation, and are not initiators of any analyzed
accident sequence. The inoperable containment penetrations will
continue to be isolated, and hence perform their isolation function.
Operation in accordance with the proposed TS will ensure that all
analyzed accidents will continue to be mitigated as previously
analyzed.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will not affect the operation of plant
equipment or the function of any equipment assumed in the accident
analysis. The PCIVs and SCIVs will continue to be operable or will
be isolated as required by the existing specifications.
TSTF-295-A, Revision 0
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change clarifies the separate condition entry Note
in TS 3.3.3.1 for Function 7, ``PCIV Position.'' The proposed change
does not affect any plant equipment, test methods, or plant
operation, and are not initiators of any analyzed accident sequence.
The actions taken for inoperable PAM channels are not changed.
Operation in accordance with the proposed TS will ensure that all
analyzed accidents will continue to be mitigated as previously
analyzed.
[[Page 49575]]
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will not affect the operation of plant
equipment or the function of any equipment assumed in the accident
analysis. The PAM channels will continue to be operable or the
existing, appropriate actions will be followed.
TSTF-306-A, Revision 2
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change revises TS 3.3.6.1 by adding an Actions Note
that would allow penetration flow paths to be unisolated
intermittently under administrative controls. Furthermore, the TIP
[traversing incore probe] isolation system is segregated into a
separate Function, allowing 24 hours to isolate the penetration. The
proposed change does not affect any plant equipment, test methods,
or plant operation, and are not initiators of any analyzed accident
sequence. The allowance to unisolate a penetration flow path will
not have a significant effect on the mitigation of any accident
previously evaluated because the penetration flow path can be
isolated, if needed, by a dedicated operator. The option to isolate
a TIP penetration will ensure the penetration will perform as
assumed in the accident analysis. Operation in accordance with the
proposed TS will ensure that all analyzed accidents will continue to
be mitigated as previously analyzed.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will not affect the operation of plant
equipment or the function of any equipment assumed in the accident
analysis. The allowance to unisolate a penetration flow path will
not have a significant effect on a margin of safety because the
penetration flow path can be isolated manually, if needed. The
option to isolate a TIP penetration will ensure the penetration will
perform as assumed in the accident analysis.
TSTF-323-A, Revision 0
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change would revise Action C of TS 3.6.1.3 to
provide a 72-hour CT instead of a 12 hour CT to isolate an
inoperable EFCV. PCIVs are not an initiator of any accident
previously evaluated. The consequences of a previously evaluated
accident during the extended CT are the same as the consequences
during the existing CT.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The PCIVs serve to mitigate the potential for radioactive
release from the primary containment following an accident. The
design and response of the PCIVs to an accident are not affected by
this change. The revised CT is appropriate given the EFCVs are on
penetrations that have been found to have acceptable barrier(s) in
the event that the single isolation valve failed.
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: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, MI
Date of amendment request: May 22, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 5.5.7, ``Inservice Testing
Program'' to: (1) Delete reference to American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code), Section
XI and incorporate reference to the ASME Code for Operation and
Maintenance of Nuclear Power Plants (ASME OM Code), and (2) address the
applicability of Surveillance Requirement (SR) 3.0.2 to other normal
and accelerated frequencies specified as two years or less in the
inservice testing (IST) program.
The proposed amendment incorporates changes based on U.S. Nuclear
Regulatory Commission (NRC)--approved Technical Specification Task
Force (TSTF) TSTF-479-A, ``Changes to Reflect Revision of 10 CFR
50.55a,'' Revision 0, as modified by NRC-approved TSTF-497, ``Limit
Inservice Testing Program SR 3.0.2 Application to Frequencies of Two
Years or Less,'' Revision 0. The proposed changes include two
deviations from the NRC-approved TSTFs that are administrative in
nature: (1) Addition of ``ASME'' to TS 5.5.7 to make references to
``ASME OM Code'' and (2) use of the term ``intervals'' instead of
``frequencies.'' 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 amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed changes do not have any impact on the
integrity of any plant system, structure, or component that
initiates an analyzed event. The proposed changes would not alter
the operation of, or otherwise increase the failure probability of
any plant equipment that initiates an analyzed accident. Thus, the
probability of any accident previously evaluated is not
significantly increased.
The proposed changes do not affect the ability to mitigate
previously evaluated accidents, and do not affect radiological
assumptions used in the evaluations. The proposed changes do not
change or alter the design criteria for the systems or components
used to mitigate the consequences of any design basis accident. The
proposed amendment does not involve operation of the required
structures, systems, or components (SSCs) in a manner or
configuration different from those previously recognized or
evaluated. Thus, the radiological consequences of any accident
previously evaluated are not increased.
Therefore, operation of the facility in accordance with the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
[[Page 49576]]
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 amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed amendment does not involve a physical
alteration of any SSC or a change in the way any SSC is operated.
The proposed amendment does not involve operation of any required
SSCs in a manner or configuration different from those previously
recognized or evaluated. No new failure mechanisms would be
introduced by the changes being requested.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The amendment does not involve a significant reduction in a
margin of safety. The proposed amendment does not affect the
acceptance criteria for any safety analysis analyzed accidents or
anticipated operational occurrences. The proposed amendment does not
alter the limiting values and acceptance criteria used to judge the
continued acceptability of components tested by the IST Program. The
safety function of the affected pumps and valves will be maintained.
Therefore, the proposed amendment would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Acting Branch Chief: Travis L. Tate.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, AR
Date of amendment request: July 31, 2007.
Description of amendment request: The proposed amendment will
revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification
(TS) 6.6.5, Core Operating Limits. The proposed change will add new
analytical methods to support the implementation of Next Generation
Fuel (NGF).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the COLR [Core Operating Limits Report]
TS are administrative in nature and have no impact on any plant
configuration or system performance relied upon to mitigate the
consequences of an accident. Changes to the calculated core
operating limits may only be made using NRC-approved methodologies,
must be consistent with all applicable safety analysis limits, and
are controlled by the 10 CFR 50.59 process.
The proposed change will add the following topical reports to
the list of referenced core operating analytical methods.
WCAP-16500-P and Final Safety Evaluation (SE)
Westinghouse topical report WCAP-16500-P describes the methods
and models that will be used to evaluate the acceptability of CE
[Combustion Engineering] 16 x 16 NGF at CE plants. Entergy has
demonstrated that the Limitations and Conditions associated with the
NRC SE will be met. Prior to implementation of NGF, the new core
design will be analyzed with applicable NRC staff-approved codes and
methods.
WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A
The proposed change allows the use of methods required for the
implementation of Optimized ZIRLO\TM\ clad fuel rods. Entergy has
demonstrated that the Limitations and Conditions associated with the
NRC SE will be met.
WCAP-16523-P and Final Safety Evaluation
This topical report describes the departure from nucleate
boiling [DNB] correlations that will be used to account for the
impact of the CE 16 x 16 NGF fuel assembly design. Entergy has
demonstrated that the Limitations and Conditions associated with the
NRC SE will be met. Prior to implementation of NGF, the new core
design will be analyzed with applicable NRC staff-approved codes and
methods.
CENPD-387-P-A
The proposed addition of this topical report provides the [DNB]
correlation that will be used to evaluate the DNB impact of non-
mixing vane grid spans for CE 16 x 16 standard and NGF assemblies.
Entergy has demonstrated that the Limitations and Conditions
associated with the NRC SE will be met.
CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety Evaluation
The addendum provides an optional steam cooling model that can
be used for Emergency Core Cooling System (ECCS) Performance
analyses to support the implementation of the CE 16 x 16 NGF fuel
assembly design. The optional steam cooling model is not being used
to support implementation of CE 16 x 16 NGF assemblies in ANO-2 at
this time. However, Entergy has demonstrated that the Limitations
and Conditions associated with the NRC SE will be met.
Assumptions used for accident initiators and/or safety analysis
acceptance criteria are not altered by the addition of these topical
reports.
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 identifies changes in the codes used to
confirm the values of selected cycle-specific reactor physics
parameter limits. The proposed change allows the use of methods
required for the implementation of CE 16 x 16 NGF. The proposed
addition of the referenced topical reports has no impact on any
plant configurations or on system performance that is relied upon to
mitigate the consequences of an accident. These changes are
administrative in nature and do not result in a change to the
physical plant or to the modes of operation defined in the facility
license.
WCAP-16500-P and Final Safety Evaluation
The proposed change adds Westinghouse topical report WCAP-16500-
P, which describes the methods and models that will be used to
evaluate the acceptability of CE 16 x 16 NGF at CE plants. Entergy
has demonstrated that the Limitations and Conditions associated with
the NRC SE will be met. Prior to implementation of NGF, the new core
design will be analyzed with applicable NRC staff-approved codes and
methods.
WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A
The proposed change allows the use of methods required for the
implementation of Optimized ZIRLO\TM\ clad fuel rods. Entergy has
demonstrated that the Limitations and Conditions associated with the
NRC SE will be met.
WCAP-16523-P and Final Safety Evaluation
This topical report describes the [DNB] correlations that will
be used to account for the impact of the CE 16 x 16 NGF fuel
assembly design. Entergy has demonstrated that the Limitations and
Conditions associated with the SE will be met.
CENPD-387-P-A
The proposed addition of this topical report provides the [DNB]
correlation that will be used to evaluate the DNB impact of non-
mixing vane grid spans for CE 16 x 16 standard and NGF assemblies.
Entergy has demonstrated that the Limitations and Conditions
associated with the NRC SE will be met.
CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety Evaluation
The addendum provides an optional steam cooling model that can
be used for ECCS Performance analyses to support the implementation
of the CE 16 x 16 NGF fuel assembly design. The optional steam
cooling model is not being used to support implementation of CE 16 x
16 NGF
[[Page 49577]]
assemblies in ANO-2 at this time. However, Entergy has demonstrated
that the Limitations and Conditions associated with the NRC SE will
be met.
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 changes do not amend the cycle-specific parameter
limits located in the COLR from the values presently required by the
TS. The individual specifications continue to require operation of
the plant within the bounds of the limits specified in COLR.
The addition of the following topical reports to the list of
analytical methods referenced in the COLR is administrative in
nature:
[squ] WCAP-16500-P and Final Safety Evaluation for Westinghouse
Electric Company (Westinghouse) Topical Report (TR) WCAP-16500-P,
Revision 0, ``CE [Combustion Engineering] 16 x 16 Next Generation
Fuel [(NGF)] Core Reference Report''
[squ] WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A
[squ] WCAP-16523-P and Final Safety Evaluation for Westinghouse
Electric Company (Westinghouse) Topical Report (TR), WCAP-16523-P,
``Westinghouse Correlations WSSV and WSSV-T for Predicting Critical
Heat Flux in Rod Bundles with Side-Supported Mixing Vanes''
[squ] CENPD-387-P-A
[squ] CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety
Evaluation for Westinghouse Electric Company (Westinghouse) Topical
Report (TR) CENPD-132 Supplement 4-P-A, Addendum 1-P, ``Calculative
Methods for the CE [Combustion Engineering] Nuclear Power Large
Break LOCA Evaluation Model--Improvement to 1999 Large Break LOCA EM
Steam Cooling Model for Less Than 1 in/sec Core Reflood''
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3,York
and Lancaster Counties, PA
Date of application for amendments: November 17, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement 3.3.1.1.8 to
increase the frequency interval between Local Power Range Monitor
(LPRM) calibrations from 1000 megawatt days per ton (MWD/T) average
core exposure to 2000 MWD/T average core exposure. The LPRM system
provides signals to associated nuclear instrumentation systems that
serve to detect conditions in the core that have the potential to
threaten the overall integrity of the fuel barrier. The LPRM system
also incorporates features designed to diagnose and display various
system trip and inoperative conditions.
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 amendment revises the surveillance interval for the
LPRM calibration from 1000 MWD/T average core exposure to 2000 MWD/T
average core exposure. Increasing the frequency interval between
required LPRM calibrations is acceptable due to improvements in core
monitoring processes and nuclear instrumentation and therefore, the
revised surveillance interval continues to ensure that the LPRM
detector signal is adequately calibrated.
This change will not alter the operation of process variables,
structures, systems, or components as described in the PBAPS Updated
Final Safety Analysis Report (UFSAR). The proposed change does not
alter the initiation conditions or operational parameters for the LPRM
system and there is no new equipment introduced by the extension of the
LPRM calibration interval. The performance of the APRM, OPRM and RBM
systems is not significantly affected by the proposed surveillance
interval increase. As such, the probability of occurrence of a
previously evaluated accident is not increased.
The radiological consequences of an accident can be affected by the
thermal limits existing at the time of the postulated accident;
however, LPRM chamber exposure has no significant effect on the
calculated thermal limits since LPRM accuracy does not significantly
deviate with exposure. For the LPRM extended calibration interval, the
total nodal power uncertainty remains less than the uncertainty assumed
in the thermal analysis basis safety limit, maintaining the accuracy of
the thermal limit calculation. Therefore, the thermal limit calculation
is not significantly affected by LPRM calibration frequency, and thus
the radiological consequences of any accident previously evaluated are
not increased.
Therefore, based on the above information, 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 performance of the APRM, OPRM and RBM systems is not
significantly affected by the proposed LPRM surveillance interval
increase. The proposed change does not affect the control parameters
governing unit operation or the response of plant equipment to
transient conditions. The proposed change does not change or introduce
any new equipment, modes of system operation or failure mechanisms.
Therefore, based on the above information, the proposed change does
not create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No
The proposed change has no impact on equipment design or
fundamental operation, and there are no changes being made to safety
limits or safety system allowable values that would adversely affect
plant safety as a result of the proposed LPRM surveillance interval
increase. The performance of the APRM, OPRM and RBM systems is not
significantly affected by the proposed change. The margin of safety can
be affected by the thermal limits existing at the time of the
postulated accident; however, uncertainties associated with LPRM
chamber exposure have no significant effect on the calculated thermal
limits. The thermal limit calculation is not significantly affected
since LPRM sensitivity with exposure is well defined. LPRM accuracy
remains within the total nodal power uncertainty assumed in the thermal
analysis basis; thereby maintaining thermal limits and the safety
margin. The proposed change does not affect safety analysis assumptions
or initial conditions and
[[Page 49578]]
therefore, the margin of safety in the original safety analyses are
maintained.
Therefore, based on the above information, the proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: Mr. J. Bradley Fewell, Associate General
Counsel, Exelon Generation Company LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Florida Power and Light Company (FPL), Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Units 1 and 2, St. Lucie County, FL
Date of amendment request: July 16, 2007.
Description of amendment request: The proposed amendment would
modify the technical specification (TS) requirements related to control
room envelope (CRE) habitability in accordance with Technical
Specification Task Force (TSTF) Traveler TSTF-448, Revision 3,
``Control Room Habitability,'' published in the Federal Register on
January 17, 2007 (Volume 72, Number 10), as part of the consolidated
line item improvement process. Specifically by modifying Unit 1 TS
3.7.7.1, ``Control Room Emergency Ventilation System (CREVS),'' and
Unit 2 TS 3.7.7,'' Control Room Emergency Air Cleanup System
(CREACS),'' and adding a new Unit 1 and Unit 2 TS Section 6.8.4.m.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
Florida Power and Light Company (FPL), Docket Nos. 50-335, St. Lucie
Plant, Unit 1, St. Lucie County, FL
Date of amendment request: July 16, 2007.
Description of amendment request: The proposed amendment would
modify the facilities operating licensing bases to adopt the
alternative source term (AST) as allowed in 10 CFR 50.67 and described
in Regulatory Guide (RG) 1.183. The licensee proposes to revise the
plant licensing basis through reanalysis of the following radiological
consequences of the Updated Final Safety Analysis Report (UFSAR)
Chapter 15 accidents: Loss-of-Coolant Accident, Fuel Handling Accident,
Main Steam Line Break, Steam Generator Tube Rupture, Reactor Coolant
Pump Shaft Seizure, Control Element Assembly Ejection, and Inadvertent
Opening of a Main Steam Safety Valve.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Alternative source term calculations have been
performed for St. Lucie Unit No. 1 which demonstrate that the dose
consequences remain below limits specified in NRC Regulatory Guide
1.183 and 10 CFR 50.67. The proposed changes do not modify the
design or operation of the plant. The use of the AST only changes
the regulatory assumptions regarding the analytical treatment of the
design basis accidents and has no direct effect on the probability
of any accident. The AST has been utilized in the analysis of the
limiting design basis accidents listed above. The results of the
analyses, which include the proposed changes to the Technical
Specifications [TSs], demonstrate that the dose consequences of
these limiting events are all within the regulatory limits.
With the exception of the deletion of SRs 4.6.6.1.c.[3].b and
4.7.8.1.c.[3].b, the proposed Technical Specification changes are
consistent with, or more restrictive than, the current TS
requirements. The proposed filter testing requirements continue to
ensure
[[Page 49579]]
that the associated filtration systems function as described in the
UFSAR and as assumed in the accident analyses. None of the affected
systems, components or programs are related to accident initiators.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not affect any plant structures,
systems, or components. The operation of plant systems and equipment
will not be affected by this proposed change. Neither implementation
of the alternative source term methodology, establishing more
restrictive TS requirements, nor deleting SRs 4.6.6.1.c.[3].b and
4.7.8.1.c.[3].b have the capability to introduce any new failure
mechanisms or cause any analyzed accident to progress in a different
manner.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed implementation of the alternative source term
methodology is consistent with NRC Regulatory Guide 1.183. With the
exception of the deletion of SRs 4.6.6.1.c.[3].b and
4.7.8.1.c.[3].b, the proposed Technical Specification changes are
consistent with, or more restrictive than, the current TS
requirements. The proposed TS requirements support the AST revisions
to the limiting design basis accidents. The proposed filter testing
requirements continue to ensure that the associated filtration
systems function as described in the UFSAR and as assumed in the
accident analyses. As such, the current plant margin of safety is
preserved. Conservative methodologies, per the guidance of RG 1.183,
have been used in performing the accident analyses. The radiological
consequences of these accidents are all within the regulatory
acceptance criteria associated with use of the alternative source
term methodology.
The proposed changes continue to ensure that the doses at the
exclusion area and low population zone boundaries and in the Control
Room are within the corresponding regulatory limits of RG 1.183 and
10 CFR 50.67. The margin of safety for the radiological consequences
of these accidents is considered to be that provided by meeting the
applicable regulatory limits, which are set at or below the 10 CFR
50.67 limits. An acceptable margin of safety is inherent in these
limits.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
Florida Power and Light Company (FPL), Docket No. 50-389, St. Lucie
Plant, Unit 2, St. Lucie County, FL
Date of amendment request: July 16, 2007.
Description of amendment request: The proposed amendment would
modify the facilities operating licensing bases to adopt the
alternative source term (AST) as allowed in 10 CFR 50.67 and described
in Regulatory Guide (RG) 1.183. The licensee proposes to revise the
plant licensing basis through reanalysis of the following radiological
consequences of the Updated Final Safety Analysis Report Chapter 15
accidents: Loss-of-Coolant Accident, Fuel Handling Accident, Main Steam
Line Break, Steam Generator Tube Rupture, Reactor Coolant Pump Shaft
Seizure, Control Element Assembly Ejection, Letdown Line Break, and
Feedwater Line Break.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Alternative source term calculations have been performed for St.
Lucie Unit No. 2 which demonstrate that the dose consequences remain
below limits specified in NRC Regulatory Guide 1.183 and 10 CFR
50.67. The proposed changes do not modify the design or operation of
the plant. The use of the AST only changes the regulatory
assumptions regarding the analytical treatment of the design basis
accidents and has no direct effect on the probability of any
accident. The AST has been utilized in the analysis of the limiting
design basis accidents listed above. The results of the analyses,
which include the proposed changes to the Technical Specifications
[TSs], demonstrate that the dose consequences of these limiting
events are all within the regulatory limits.
The proposed Technical Specification Changes are consistent
with, or more restrictive than, the current TS requirements. None of
the affected systems, components or programs are related to accident
initiators.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not affect any plant structures,
systems, or components. The operation of plant systems and equipment
will not be affected by this proposed change. Neither implementation
of the alternative source term methodology nor establishing more
restrictive TS requirements have the capability to introduce any new
failure mechanisms or cause any analyzed accident to progress in a
different manner.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed implementation of the alternative source term
methodology is consistent with NRC Regulatory Guide 1.183. The
proposed Technical Specification changes are consistent with, or
more restrictive than, the current TS requirements. These TS
requirements support the AST revisions to the limiting design basis
accidents. As such, the current plant margin of safety is preserved.
Conservative methodologies, per the guidance of RG 1.183, have been
used in performing the accident analyses. The radiological
consequences of these accidents are all within the regulatory
acceptance criteria associated with use of the alternative source
term methodology.
The proposed changes continue to ensure that the doses at the
exclusion area and low population zone boundaries and in the Control
Room are within the corresponding regulatory limits of RG 1.183 and
10 CFR 50.67. The margin of safety for the radiological consequences
of these accidents is considered to be that provided by meeting the
applicable regulatory limits, which are set at or below the 10 CFR
50.67 limits. An acceptable margin of safety is inherent in these
limits.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
Based on the above discussion, FP&L has determined that the
proposed change does not involve a significant hazards
consideration.
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.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, MN
Date of amendment request: July 3, 2007.
Description of amendment request: The proposed amendments would
revise the Technical Specifications
[[Page 49580]]
(TSs) for the Prairie Island Nuclear Generating Plant (PINGP), Units 1
and 2 to:
1. Revise TS 1.4, ``Frequency'' to modify the second paragraph of
Example 1.4-1 to be consistent with the requirements of Surveillance
Requirement (SR) 3.0.4 and incorporate the changes in Technical
Specification Task Force (TSTF) industry traveler TSTF-485, ``Correct
Example 1.4-1.''
2. Revise TS 5.5.7.a, to modify references to Section XI of the
American Society of Mechanical Engineers (ASME) Code with references to
the ASME Code for Operation and Maintenance of Nuclear Power Plants
(ASME OM Code), to be consistent with TSTF-479, ``Changes to Reflect
Revision of 10 CFR [Code of Federal Regulations] 50.55a.
3. Revise TS 5.5.7.b, to restrict extension of Frequencies to those
Frequencies specified as 2 years or less, and take exception to the
limitation in SR 3.0.2 which does not apply the 1.25 times extension to
Frequencies of 24 months, to be consistent with TSTF-479 and TSTF-497,
``Limit Inservice Testing Program SR 3.0.2 Application to Frequencies
of 2 Years or Less.''
4. Revise TS 5.5.7.d, to modify the referenced ASME Code to be
consistent with TSTF-479.
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:
TSTF-479
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 Improved Standard Technical
Specification (ISTS) Inservice Testing Program for consistency with
the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which
are classified as American Society of Mechanical Engineers (ASME)
Code Class 1, Class 2 and Class 3. 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 change does 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, this
proposed change does not represent 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 revises the improved Standard Technical
Specification (ISTS) Inservice Testing Program for consistency with
the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which
are classified as American Society of Mechanical Engineers (ASME)
Code Class 1, Class 2 and Class 3. 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 change does 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. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the improved Standard Technical
Specification (ISTS) Inservice Testing Program for consistency with
the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which
are classified as American Society of Mechanical Engineers (ASME)
Code Class 1, Class 2 and Class 3. 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.
TSTF-485
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 Section 1.4, Frequency, Example 1.4-
1, to be consistent with Surveillance Requirement (SR) 3.0.4 and
Limiting Condition for Operation (LCO) 3.0.4. This change is
considered administrative in that it modifies the example to
demonstrate the proper application of SR 3.0.4 and LCO 3.0.4. The
requirements of SR 3.0.4 and LCO 3.0.4 are clear and are clearly
explained in the associated Bases. As a result, modifying the
example will not result in a change in usage of the Technical
Specifications (TS). The proposed change does not adversely affect
accident initiators or precursors, the ability of structures,
systems, and components (SSCs) to perform their intended function to
mitigate the consequences of an initiating event within the assumed
acceptance limits, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Therefore, this change is considered administrative and
will have no effect on the probability or consequences of any
accident previously evaluated. 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.
No new or different accidents result from utilizing the proposed
change. The change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the change does not impose any new or
different requirements or eliminate any existing requirements. The
change does not alter assumptions made in the safety analysis. The
proposed change is 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.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is administrative and will have no effect on
the application of the Technical Specification requirements.
Therefore, the margin of safety provided by the Technical
Specification requirements is unchanged. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
TSTF-497
This Traveler is considered an administrative change to the ISTS
NUREGs. Therefore, a regulatory analysis is not provided.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Travis L. Tate.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Rivers, Manitowoc
County, WI
Date of amendment request: July 12, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.6.3, ``Containment Isolation
Valves.'' The revision would delete Surveillance
[[Page 49581]]
Requirement (SR) 3.6.3.1, which is no longer required due to the
containment purge supply and exhaust valve isolation function being
replaced with blind flanges. The proposed amendment would also support
a change to the Final Safety Analysis Report (FSAR) to revise the
requirement to leak check the purge supply and exhaust valves.
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 to the containment purge supply penetration
and the containment exhaust penetration presents no change in the
probability or the consequence of an accident. The penetrations
continue to conform to the TS requirements for containment and will
be appropriately tested as required by 10 CFR 50 Appendix J. The
blind flanges are passive devices not susceptible to an active
failure or malfunction that could result in a loss of isolation or
leakage that exceeds the limits assumed in the safety analyses. The
blind flanges are leak rate tested in accordance with the
containment leakage rate testing program. Containment isolation is
not lessened by this change.
The change to the containment purge system does not affect the
design basis limit for any fission product barrier.
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 to the containment purge supply penetration
and the containment exhaust penetration does not change the function
of the system and does not alter containment isolation. The
penetrations continue to conform to the TS requirements for
containment isolation and will be appropriately tested as required
by 10 CFR 50 Appendix J. No new accident scenarios, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed changes.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The proposed change will not alter any assumptions, initial
conditions or results specified in any accident analysis. The
containment purge supply and exhaust penetrations will continue to
conform to the TS requirements for containment and will be
appropriately tested as required by 10 CFR 50 Appendix J. The blind
flanges are passive devices not susceptible to an active failure or
malfunction that could result in a loss of isolation or leakage that
exceeds limits assumed in the safety analysis. The blind flanges are
leak rate tested in accordance with the containment leakage rate
testing program. Containment isolation is not lessened by this
change. Therefore, there is no 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.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Travis L. Tate.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, NE
Date of amendment request: July 30, 2007.
Description of amendment request: The proposed amendment by Omaha
Public Power District requests changes to the Fort Calhoun Station Unit
No.1 Operating License No. DPR-40 to modify the containment spray
system actuation logic to preclude automatic start of the containment
spray pumps for a loss-of-coolant accident.
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 containment spray (CS) system and the containment air
cooling and filtering system (CACFS) are not initiators of any
accident previously evaluated at the Fort Calhoun Station (FCS).
Both systems are accident mitigation systems. Their licensing basis
functions are to limit the containment pressure rise and reduce the
leakage of airborne radioactivity from the containment by providing
a means for cooling the containment following a loss-of-coolant
accident (LOCA) or main steam line break (MSLB) inside containment.
The proposed modification to the CS system logic shifts the function
of containment pressure and temperature control during a LOCA from
the [CS] system to the equally capable and reliable containment air
coolers. The change in the CS actuation logic does not impact the
containment response to the MSLB analysis of record (AOR). The CACFS
provides the design heat removal capabilities for the containment
during the postulated LOCA. The system is operated to remove
atmospheric heat loads from the containment during normal plant
operation. Since system components are only lightly loaded during
normal operation, system availability and reliability are enhanced.
In the unlikely event that normal power sources are lost and one
emergency diesel generator fails to operate, one containment air
cooling and filtering unit and one containment air cooling unit will
operate.
The component cooling water (CCW) system, on which the CACFS is
dependent, has sufficient capacity for all normal and shutdown
operating modes. In addition, the system is capable of satisfying
the design criteria under post design-basis accident (DBA)
conditions with the single failure of an active component and a loss
of instrument air. Analyses demonstrate that CCW flowrates to
essential equipment would be adequate for removing post accident
design-basis heat loads.
Following implementation of the proposed change, at least one
train of containment air coolers will be available to mitigate a
LOCA. Analyses show that one train of coolers can maintain the
containment pressure and temperature below the design values;
therefore, the proposed change will have no adverse effect on the
containment pressure analysis following a LOCA.
Analyses have also shown that one train of containment high-
efficiency particulate air (HEPA) filters maintains the radiological
consequences doses within regulatory limits; therefore, the proposed
change will have no adverse effect on the radiological consequences
following a LOCA.
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 CACFS was designed to remove heat released to containment
atmosphere during the [DBA] to the extent necessary to maintain the
structure below the design pressure. The proposed modification to
the CS system logic shifts the function of containment pressure and
temperature control from the [CS] system to the equally capable and
reliable containment air coolers. The use of CACFS, as a means of
containment pressure control, has been evaluated for the LOCA event
and found to result in an acceptable peak containment pressure (peak
pressure less than 60 psig [pounds per square inch gauge]).
Radiological consequences were evaluated for the use of CACFS in
this application using the guidance provided in Regulatory Guide
(RG) 1.183. This radiological analysis demonstrates that the dose
consequences are in compliance with applicable regulatory
requirements. The estimated dose consequences at the exclusion area
boundary (EAB), low population zone (LPZ), and control room (CR)
remain within the acceptance criteria of 10 CFR 50.67 as
supplemented by RG 1.183 and the standard review plan (SRP) 15.0.1.
The assessment also demonstrates that the dose consequences in the
technical support center (TSC) remain compliant with regulatory
guidance provided in Supplement 1 of NUREG-0737.
[[Page 49582]]
No credible new failure mechanisms, malfunctions, or accident
initiators not considered in the design and licensing basis have
been created and none of the initial condition assumptions of any
accident evaluated in the safety analysis are impacted.
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 containment building and associated penetrations are
designed to withstand an internal pressure of 60 psig at 305 [deg]F
[degrees Fahrenheit], including all thermal loads resulting from the
temperature associated with this pressure, with a leakage rate of
0.1 percent by weight or less of the contained volume per 24 hours.
The containment air coolers are credited for maintaining containment
pressure and temperatures within design limitations, and assure that
the release of fission products to the environment following a [DBA]
will not exceed regulatory guidelines for a large break (LB) LOCA.
The [CS] system and containment air coolers continue to be
credited for limiting peak containment pressure for an MSLB.
Adequate NPSH [net positive suction head] margin is maintained
for the HPSI [high-pressure safety injection] pumps during the
recirculation phase of a[n] LBLOCA due to the reduction in ECCS
[emergency core cooling system] sump strainer pressure drop.
The CACFS operates independently of the CS system to remove heat
from the containment atmosphere. The CACFS consists of two redundant
trains, each train with one air cooling and filtering unit and one
air cooling unit, for a total of four cooling units. Operation of
the CACFS, in accordance with analyses completed for the 2006 steam
generator replacement, is and will continue to be credited in the
MSLB containment pressure analysis. The operation and maintenance of
the CACFS are not impacted by this proposed change. Therefore, the
containment heat removal licensing basis is not adversely affected
by the proposed change. The ability to maintain containment peak
pressure and temperature, as well as long-term containment pressure
and temperature, is maintained.
The LBLOCA 10 CFR 50.46 analysis assumes that there will be
three CS pumps operating when evaluating the effects of containment
pressure on ECCS performance. This assumption minimizes containment
pressure, to conservatively evaluate ECCS performance in response to
a LOCA. Eliminating operation of the CS pumps improves ECCS
performance and thus increases margin to 10 CFR 50.46 limits on peak
clad temperature, therefore, the existing analysis remains bounding
as is.
In summary, following implementation of the proposed change:
[squ] Peak containment pressure for analyzed DBAs remains within
design limits;
[squ] Radiological releases remain within the limits of 10 CFR
50.67; and
[squ] The currently calculated peak clad temperature following a
LOCA remains bounding.
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: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, NE
Date of amendment request: July 31, 2007.
Description of amendment request: The proposed amendment will
modify Technical Specification (TS) requirements to support a planned
inverter modification to be installed during the 2008 refueling outage.
The inverter modification will require revisions to TS 2.7(1), 2.7(2),
and 3.7(5), and the associated Bases sections to allow for the addition
of two safety-related swing inverters.
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 addition of two safety-related swing inverters to the 120 V a-c
[Volts alternating current] vital instrument buses is not an initiator
of any previously evaluated accidents. The swing inverters will not
prevent safety systems from performing any of the accident mitigation
functions assumed in the safety analysis. The revisions proposed for
the Technical Specifications (TS) take advantage of the operational
flexibility provided by the swing inverters yet maintain current TS
requirements that four inverters be operable.
Similarly, the change maintains the current TS allowance for one of
the required inverters to be inoperable for up to twenty-four hours
provided all current TS requirements for operability are met.
Although continued operation for up to twenty-four hours with one
of the required inverters inoperable is allowed, the addition of the
two safety-related swing inverters is expected to decrease the amount
of time that the station must operate with less than four inverters.
This is because the design allows the inoperable inverter to be
replaced by its associated swing (or non-swing) inverter. Reducing the
need to shut the station down due to an inoperable inverter also
reduces the risk associated with mode transition to shutdown.
The correction of two typographical errors and correcting spacing
inconsistencies in the text are administrative changes that do not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any 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 design function of the safety-related inverters is unchanged.
The addition of the safety-related swing inverters and their bypass
sources to the 120 volt a-c vital instrument distribution system allows
preventative maintenance, repair and for testing to be performed
online. If a safety-related inverter becomes inoperable or is otherwise
out-of-service, its instrument bus is manually transferred to the
associated swing inverter. If a required inverter should fail, the time
that the station will operate with less than the four inverters
required by TS 2.7(1)j should, in most cases, be less due to the
ability to place an associated inverter online. Reducing the need to
shut the station down due to an inoperable inverter also reduces the
risk associated with mode transition to shutdown.
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 involve a significant reduction in
the margin of safety?
Response: No
The design function of the safety-related inverters is unchanged.
The addition of the safety-related swing inverters to the 120 volt a-c
vital instrument distribution system allows preventative maintenance or
repair of a safety-related inverter to be performed online since its
instrument bus can be manually transferred to the associated swing
inverter. Installation of the safety-related swing inverters does not
require changes to accident analyses or results. The revisions proposed
for the TS
[[Page 49583]]
maintain current TS requirements that four inverters be operable.
Should a required inverter fail, the time that the station will operate
with less than the four inverters required by TS 2.7(1)j should, in
most cases, be less due to the ability to place an associated inverter
online. Reducing the need to shut the station down due to an inoperable
inverter also reduces the risk associated with mode transition to
shutdown. In addition, administrative controls are in place to ensure
the current station battery capacity is not degraded and to ensure
battery margin is adequately maintained as a result of the inverter
modification.
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: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, TN
Date of amendment request: July 26, 2007.
Description of amendment request: The proposed amendment would add
a new reference to Technical Specification 6.9.1.14.a, which lists
documents that have been approved by the U.S. Nuclear Regulatory
Commission for use in determining the core operating limits. The new
reference is the Areva NP, Inc. topical report EMF-2103P-A, ``Realistic
Large Break LOCA Methodology for Pressurized Water Reactors.''
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 adds an approved analytical method for
evaluating large break loss of coolant accidents (LOCAs). The
proposed change will not affect previously evaluated accidents
because they continue to be analyzed by NRC approved methodologies
to ensure required safety limits are maintained. The acceptance
criteria of the SQN Final Safety Analysis Report analyzed accidents
and anticipated operational occurrences are not affected by the
proposed addition of the realistic large break LOCA methodology. As
the evaluations for accidents and operation occurrences are not
adversely affected, the proposed change will not increase the
consequences of a postulated event. The proposed change does not
result in any modification of the plant equipment or operating
practices and therefore, does not alter plant conditions or plant
response prior to or after postulated events. 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.
As previously noted, the proposed change does not result in any
modification of the plant equipment or operating practices and
therefore, does not alter plant conditions or plant response prior
to or after postulated events. 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 alter plant equipment including the
automatic accident mitigation setpoints designed to mitigate the
affects of a postulated accident. The accident analyses and plant
safety limits continue to be acceptable as evaluated by NRC approved
methodologies. The proposed application of the realistic large break
LOCA methodology ensures acceptable margins and limits for fuel core
designs. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(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].
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, NJ
Date of application for amendment: November 27, 2006.
Brief description of amendment: The amendment revised the required
submittal date for the Annual Radioactive Effluent Release Report.
Specifically, the required submittal date is revised from ``within 60
days after January 1, each year,'' to ``prior to May 1 of each year.''
Date of Issuance: August 8, 2007.
[[Page 49584]]
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No: 264.
Facility Operating License No. DPR-16: The amendment revised the
license and the Technical Specifications.
Date of initial notice in Federal Register: May 8, 2007 (72 FR
26174).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated August 8, 2007.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi, Unit 2, Monroe
County, MI
Date of application for amendment: July 12, 2006, as supplemented
by letters dated April 25, May 23, June 15, June 20, and June 29, 2007.
Brief description of amendment: The amendment modifies Conditions,
Required Actions and Completion Times in Technical Specification (Ts)
3.8.1, ``AC Sources-Operating,'' associated with the Required Actions
when emergency diesel generators are declared inoperable.
Date of issuance: August 1, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No: 175.
Facility Operating License No. NPF-43: Amendment revised the TSs
and License.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51225). The April 25, May 23, June 15, June 20, and June 29, 2007,
supplements, contained clarifying information and did not change the
NRC staff's initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 1, 2007.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, MI
Date of application for amendment: January 26, 2007.
Brief description of amendment: The amendment adds a Limiting
Condition for Operation (LCO) 3.0.9 to the Technical Specifications
(TS), allowing a delay time for entering a supported system TS, when
the inoperability is due solely to an unavailable barrier, if risk is
assessed and managed. Additionally, the amendment makes editorial
changes to LCO 3.0.8 to be consistent with terminology of LCO 3.0.9.
Date of issuance: August 1, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 176.
Facility Operating License No. NPF-43: Amendment revised the TS and
License.
Date of initial notice in Federal Register: April 10, 2007 (72 FR
17945).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 1, 2007.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, OH
Date of application for amendment: May 30, 2006, as supplemented by
letters dated April 24, 2007, and June 27, 2007.
Brief description of amendment: This amendment revises the existing
SG tube surveillance program to be consistent with the Nuclear
Regulatory Commission's approved TS Task Force (TSTF) Standard TS
Change Traveler, TSTF-449, ``Steam Generator Tube Integrity.'' A notice
of availability for this TS improvement using the consolidated line
item improvement process was published in the Federal Register on May
6, 2005 (70 FR 24126). The amendment is also the modification of the SG
portion of the TSs requested in NRC Generic Letter (GL) 2006-01,
``Steam Generator Tube Integrity and Associated Technical
Specification.''
Date of issuance: July 31, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 276.
Facility Operating License No. NPF-3: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: October 10, 2006 (71 FR
59531). The April 24, 2007, and June 27, 2007 supplements, contained
clarifying information and did not change the NRC staff's initial
proposed finding of no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 31, 2007.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No. 2, Oswego County, NY
Date of application for amendment: March 8, 2007.
Brief description of amendment: The amendment revises the Technical
Specification requirements for inoperable snubbers by adding Limiting
Condition for Operation 3.0.8 using the Consolidated Line Item
Improvement Process.
Date of issuance: July 30, 2007.
Effective date: As of the date of issuance to be implemented within
180 days.
Amendment No.: 118.
Renewed Facility Operating License No. NPF-69: Amendment revises
the License and Technical Specifications.
Date of initial notice in Federal Register: April 24, 2007 (72 FR
20384).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 30, 2007.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear
Plant, Unit 1, Limestone County, AL
Date of application for amendment: June 25, 2007, as supplemented
by letters dated July 3 and 26, 2007 (TS-461).
Brief description of amendment: The amendment deletes License
Condition 2.G.(2) as the result of completion of power uprate large
transient testing.
Date of issuance: August 14, 2007.
Effective date: The date of issuance, to be implemented within 30
days.
Amendment No.: 272.
Renewed Facility Operating License No. DPR-33: Amendment revised
the renewed operating license.
Date of initial notice in Federal Register: July 13, 2007 (72 FR
38627). The July 3 and 26, 2007, supplemental letters provided
clarifying information that did not expand the scope of the application
or change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 14, 2007.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the
[[Page 49585]]
Atomic Energy Act of 1954, as amended (the Act), and the Commission's
rules and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397 4209, (301) 415-4737 or by email to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\
[[Page 49586]]
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
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Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by email to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, IL
Date of amendment request: June 29, 2007, as supplemented by
letters dated August 1, 2007 and August 2, 2007.
Description of amendment request: The amendments revised the
maximum allowed Technical Specification (TS) temperature limit,
contained in TS Surveillance Requirement 3.7.3.1, of the cooling water
supplied to the plant from the Core Standby Cooling System (CSCS) pond
(i.e., the Ultimate Heat Sink) from 100 [deg]F to 101.25 [deg]F.
Date of issuance: August 2, 2007.
Effective date: August 2, 2007.
Amendment Nos.: 183 and 170.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications and License.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated August 2,
2007.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Dated at Rockville, Maryland, this 20th day of August 2007.
For The Nuclear Regulatory Commission.
John W. Lubinski,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-16766 Filed 8-27-07; 8:45 am]
BILLING CODE 7590-01-P