[Federal Register Volume 70, Number 79 (Tuesday, April 26, 2005)]
[Notices]
[Pages 21449-21470]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-8166]
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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 April 1, 2005, through April 14, 2005. The
last biweekly notice was published on April 12, 2005 (70 FR 19110).
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, Rules and
Directives 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
[[Page 21450]]
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: February 25, 2005.
Description of amendment request: The proposed change would delete
Section 2.G of the Clinton's Facility Operating License (FOL), NPF-62,
which requires AmerGen Energy Company, LLC, to report violations of the
requirements contained in Section 2.C of this license. The proposed
change will reduce unnecessary regulatory burden and will allow AmerGen
to take full advantage of the revisions to Title 10, Code of Federal
Regulations (10
[[Page 21451]]
CFR), Section 50.72, ``Immediate notification requirements for
operating nuclear power reactors,'' and 10 CFR 50.73, ``Licensee event
report system.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change involves an administrative change only. The
proposed change does not involve the modification of any plant
equipment or affect plant operation. The proposed change will have
no impact on any safety related structures, systems or components.
The reporting requirement section of the FOL is not required because
the requirements are either adequately addressed by 10 CFR 50.72 and
10 CFR 50.73, or other regulatory requirements, or are not required
based on the nature of the Condition.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change has no impact on the design, function or
operation of any plant structure, system or component. The proposed
change is administrative in nature and does not affect plant
equipment or accident analyses. The reporting requirement section of
the FOL is not required because the requirements are either
adequately addressed by 10 CFR 50.72 and 10 CFR 50.73, or other
regulatory requirements, or are not required based on the nature of
the Condition.
Therefore, the proposed change 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?
The proposed change is administrative in nature, does not negate
any existing requirement, and does not adversely affect existing
plant safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there is no change being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by deletion of the
reporting requirement that is adequately addressed elsewhere.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: March 25, 2005.
Description of amendment request: The proposed change would revise
Technical Specification Surveillance Requirement (SR) 3.6.1.3.8 to add
a note excluding leakage through primary containment penetrations 1MC-
101 and 1MC-102 from the secondary containment bypass leakage total
specified in the SR.
Implementation of this proposed change will provide operational
flexibility by allowing Clinton Power Station (CPS) to utilize the
additional margin in the regulatory dose limit analysis that supports
the implementation of the alternative source term.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment adds a note excluding the leakage through
the primary containment purge lines from the secondary containment
bypass leakage based on separate analysis of these paths using the
assumptions in the alternative source term (AST) revision to the
loss of coolant accident (LOCA) analysis.
The proposed change does not require modification to the
facility. The proposed change in secondary containment bypass
leakage does not affect the operation of any facility equipment, the
interface between facility systems, or the reliability of any
equipment. In addition, secondary containment bypass leakage does
not constitute an initiator of any previously evaluated accidents.
Therefore, the proposed amendment does not involve a significant
increase in the probability of an accident previously evaluated.
The radiological consequences of the LOCA analysis using the
primary containment purge line leakage as separate from the
secondary containment bypass leakage, has been evaluated as part of
the application of AST assumptions. The results conclude that the
radiological consequences remain within applicable regulatory
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 does not affect the design, functional
performance or operation of the facility. No new equipment is being
introduced and installed equipment is not being operated in a new or
different manner. Similarly, the proposed change does not affect the
design or operation of any structures, systems or components
involved in the mitigation of any accidents, nor does it affect the
design or operation of any component in the facility such that new
equipment failure modes are created. There are no set points at
which protective or mitigative actions are initiated that are
affected by this proposed action. No change is being made to
procedures relied upon to respond to an off-normal event.
As such 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 amendment involve a significant reduction
in a margin of safety?
Response: No.
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the establishment of set points to initiate
alarms or actions. The proposed change adds a note excluding the
leakage through the primary containment purge lines from the
secondary containment bypass leakage based on separate analysis of
these paths using the assumptions in the AST revision to the LOCA
analysis. There is no change in the design of the affected systems,
no alteration of the set points at which alarms or actions are
initiated, and no change in plant configuration from original
design.
The margin of safety is considered to be that provided by
meeting the applicable regulatory limits. The AST analysis indicates
that the doses following a LOCA remain within the regulatory limits,
and therefore, there is not a significant reduction in a margin of
safety. The AST analysis confirms the change continues to ensure
that the doses at the exclusion area and low population zone
boundaries, as well as the control room, are within the
corresponding regulatory limits.
Therefore, operation of CPS in accordance with the proposed
change will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel,
[[Page 21452]]
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Section Chief: Gene Y. Suh.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: April 1, 2005.
Description of amendment request: The proposed changes would
incorporate into the Technical Specifications (TSs) the Oscillation
Power Range Monitor (OPRM) instrumentation that will be declared
operable within 30 days after completion of the February 2006 refueling
outage. The proposed changes would add TS Section 3.3.1.3,
``Oscillation Power Range Monitor (OPRM) Instrumentation,'' and would
revise TS Sections 3.4.1, ``Recirculation Loops Operating,'' and 5.6.5,
``Core Operating Limits Report (COLR).'' In addition, the changes would
insert a new TS section for the OPRM instrumentation, delete the
current thermal-hydraulic instability administrative requirements, and
add the appropriate references for the OPRM trip set points and
methodology. Clinton Power Station (CPS) will activate the automatic
reactor protection system (i.e., scram) outputs of the OPRM
instrumentation upon implementation of these proposed TS changes.
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 specify limiting conditions for operation,
required actions and surveillance requirements for the OPRM system,
and allows operation in regions of the power to flow map currently
restricted by the requirements of the Interim Corrective Actions
(ICAs) and certain limiting conditions of operation of TS Section
3.4.1, ``Recirculation Loops Operating.'' The restrictions of the
ICAs and TS Section 3.4.1 were imposed to ensure adequate capability
to detect and suppress conditions consistent with the onset of
thermal-hydraulic oscillations that may develop into a thermal-
hydraulic instability event. A thermal-hydraulic instability event
has the potential to challenge the Minimum Critical Power Ratio
(MCPR) safety limit. The OPRM system can automatically detect and
suppress conditions necessary for thermal-hydraulic instability.
With the activation of the OPRM system, the restrictions of the ICAs
and TS Section 3.4.1 will no longer be required.
This proposed change has no impact on any of the existing
neutron monitoring functions. When the OPRM is operable with
operating limits as specified in the Core Operating Limits Report
(COLR), the OPRM can automatically detect the imminent onset of
local power oscillations and generate a trip signal. Actuation of a
Reactor Protection System (RPS) trip (i.e., scram) will suppress
conditions necessary for thermal-hydraulic instability and decrease
the probability of a thermal-hydraulic instability event. In the
event the trip capability of the OPRM is not maintained, the
proposed changes limit the period of time before an alternate method
to detect and suppress thermal-hydraulic oscillations is required.
CPS intends to utilize the ICAs as the alternative method for
ensuring thermal-hydraulic oscillations do not occur. Since the
duration of this period of time is limited, the increase in the
probability of a thermal-hydraulic instability event is not
significant.
Activation of the OPRM scram function will replace the current
methods that require operators to insert an immediate manual reactor
scram in certain reactor operating regions where thermal hydraulic
instabilities could potentially occur. While these regions will
continue to be avoided during normal operation, certain transients,
such as a reduction in reactor recirculation flow, could place the
reactor in these regions. During these transient conditions, with
the OPRM instrumentation scram function activated; an immediate
manual scram will no longer be required. This may potentially cause
a marginal increase in the probability of occurrence of an
instability event. This potential increase in probability is
acceptable because the OPRM function will automatically detect the
instability condition and initiate a reactor scram before the
Minimum Critical Power Ratio (MCPR) Safety Limit is reached.
Consequences of the potential instability event are reduced because
of the more reliable automatic detection and suppression of an
instability event, and the elimination of dependence on the manual
operator actions. Operators monitor for indications of thermal
hydraulic instability when the reactor is operating in regions of
potential instability as a backup to the OPRM instrumentation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes replace procedural actions that were
established to avoid operating conditions where reactor
instabilities might occur with an NRC approved automatic detect and
suppress function (i.e., OPRM).
Potential failures in the OPRM trip function could result in
either failure to take the required mitigating action or an
unintended reactor scram. These are the same potential effects of
failure of the operator to take the correct appropriate action under
the current procedural actions. The effects of failure of the OPRM
equipment are limited to reduced or failed mitigation, but such
failure cannot cause an instability event or other type of accident.
The OPRM system uses input signals shared with the Average Power
Range Monitor (APRM) system and rod block functions to monitor core
conditions and generate a Reactor Protection System (RPS) trip when
required. Quality requirements for software design, testing,
implementation and module self-testing of the OPRM system provide
assurance that no new equipment malfunctions due to software errors
are created. The design of the OPRM system also ensures that neither
operation nor malfunction of the OPRM system will adversely impact
the operation of the other systems and no accident or equipment
malfunction of these other systems could cause the OPRM system to
malfunction or cause a different kind of accident. No new failure
modes of either the new OPRM equipment or of the existing APRM
equipment have been introduced.
Operation in regions currently restricted by the ICAs and TS
Section 3.4.1 is within the nominal operating domain and ranges of
plant systems and components for which postulated equipment and
accidents have been evaluated. Therefore, operation within these
regions does not create the possibility of a new or different kind
of accident from any previously evaluated.
These proposed changes which specify limiting conditions for
operations, required actions and surveillance requirements of the
OPRM system and allow operation in certain regions of the power-to-
flow map do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The OPRM system monitors small groups of Local Power Range
Monitor (LPRM) signals for indication of local variations of core
power consistent with thermal-hydraulic oscillations and generates
an RPS trip when conditions consistent with the onset of
oscillations are detected. An unmitigated thermal-hydraulic
instability event has the potential to result in a challenge to the
MCPR safety limit. The OPRM system provides the capability to
automatically detect and suppress conditions that might result in a
thermal-hydraulic instability event and thereby maintains the margin
of safety by providing automatic protection for the MCPR safety
limit while reducing the burden on the control room operators
significantly. The OPRM trip provides a trip output of the same type
as currently used for the APRM. Its failure modes and types are
similar to those for the present APRM output. Since the MCPR Safety
Limit will not be exceeded as a result of an instability event
following implementation of the OPRM trip function, it is concluded
that the proposed change does not reduce the margin of safety.
Operation in regions currently restricted by the requirements of
the ICAs and TS Section 3.4.1 is within the nominal operating domain
assumed for identifying the range of initial
[[Page 21453]]
conditions considered in the analysis of anticipated operational
occurrences and postulated accidents. Therefore, operation in these
regions does not involve a significant reduction in the margin of
safety.
The proposed changes, which specify limiting conditions for
operations, required actions and surveillance requirements of the
OPRIVI system and allow operation in certain regions of the power to
flow map, do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: February 25, 2005.
Description of amendment request: The proposed change would delete
Section 2.E of the Oyster Creek's Facility Operating License (FOL),
DPR-16, which requires AmerGen Energy Company, LLC, to report
violations of the requirements contained in Section 2.C of this
license. The proposed change will reduce unnecessary regulatory burden
and will allow AmerGen to take full advantage of the revisions to Title
10, Code of Federal Regulations (10 CFR), Section 50.72, ``Immediate
notification requirements for operating nuclear power reactors,'' and
10 CFR 50.73, ``Licensee event report system.''
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 involves an administrative change only. The
proposed change does not involve the modification of any plant
equipment or affect plant operation. The proposed change will have
no impact on any safety related structures, systems or components.
The reporting requirement section of the FOL is not required because
the requirements are either adequately addressed by 10 CFR 50.72 and
10 CFR 50.73, or other regulatory requirements, or are not required
based on the nature of the Condition.
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 has no impact on the design, function or
operation of any plant structure, system or component. The proposed
change is administrative in nature and does not affect plant
equipment or accident analyses. The reporting requirement section of
the FOL is not required because the requirements are either
adequately addressed by 10 CFR 50.72 and 10 CFR 50.73, or other
regulatory requirements, or are not required based on the nature of
the Condition.
Therefore, the proposed change 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 proposed change is administrative in nature, does not negate
any existing requirement, and does not adversely affect existing
plant safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there is no change being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by deletion of the
reporting requirement that is adequately addressed elsewhere.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Richard J. Laufer.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: March 17, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.4.10, ``Reactor Coolant System (RCS)
Pressure and Temperature (P/T) Limits,'' to replace the combination
figure with separate P/T limit figures for each one of the three
categories of operation: hydrostatic pressure test [Curve A], non-
nuclear heatup and cooldown [Curve B], and nuclear (core critical)
operation [Curve C]. The new curves also provide composite limits for
all reactor pressure vessel (RPV) regions including core beltline
region. RPV bottom head individual limit curves are superimposed on
Curves A and B. In addition, two sets of curves are calculated; one for
32 effective full power years (EFPY) which represents the end of the
current 40-year plant license and the other one is for 24 EFPY which
has been selected as an intermediate point between the current EFPY and
32 EFPY.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The revised P/T curves are based on the 1998 Edition of the
American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel (B&PV) Code, Section XI, including the 2000 Addenda. This
edition of the Code has been approved for use in both 10 CFR 50.55a
and Regulatory Guide (RG) 1.147. The revised curves are also based
on updated fluence calculations performed utilizing NRC-approved
methodology consistent with RG 1.190 for calculating Reactor
Pressure Vessel (RPV) neutron fluence. Revised fluence calculations
are applicable for 24 and for 32 Effective Full Power Years (EFPY).
The 32 EFPY represents a conservative exposure level at the end of
the current 40-year plant operating license. The proposed change
incorporates adjustment of the reference temperature for all
beltline material to account for irradiation effects and provide a
comparable level of protection as previously evaluated and approved.
The adjusted reference temperature calculations were performed in
accordance with the requirements of 10 CFR 50 Appendix G using the
guidance contained in RG 1.99, Revision 2, to provide operating
limits for up to 32 EFPY.
There are no changes being made to the RCS pressure boundary or
to RCS material, design or construction standards. The proposed P/T
curves define limits that continue to ensure the prevention of
nonductile failure of the RCS pressure boundary. The revision of the
P/T curves does not alter any assumptions previously made in the
radiological consequence evaluations since the integrity of the RCS
pressure boundary is unaffected. Therefore, the proposed changes
will not significantly increase the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
[[Page 21454]]
The revised P/T curves are based on a later edition and addenda
of the ASME Code that incorporates current industry standards for
the curves. The revised curves are also based on an RPV fluence that
has been recalculated in accordance with the methodology of RG
1.190. The proposed change does not involve a modification to plant
structures, systems or components. There is no effect on the
function of any plant system, and no newly introduced system
interactions. The proposed change does not create new failure modes
or cause any systems, structures or components to be operated beyond
their design bases. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed P/T curves define the limits of operation to
prevent nonductile failure of the RPV upper vessel, bottom head and
beltline region. The new curves conform to the guidance contained in
RG 1. 190, ``Calculational and Dosimetry Methods for Determining
Pressure Vessel Neutron Fluence,'' and RG 1.99, Revision 2,
``Radiation Embrittlement of Reactor Vessel Materials,'' and
maintain the safety margins specified in 10 CFR 50 Appendix G.
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 G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279
NRC Section Chief: L. Raghavan.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: March 17, 2005. This amendment request
supercedes, in its entirety, a previous application dated March 19,
2004, published in the Federal Register on June 22, 2004 (69 FR 34698).
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.6.1, ``Primary Containment
Isolation Instrumentation,'' to correct a formatting error introduced
during conversion to Improved Technical Specifications (ITS) by
replacing ``1 per room'' with ``2'' for the required channels per trip
system for the reactor water cleanup (RWCU) area ventilation
differential temperature--high primary containment isolation
instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change restores the number of Required Channels Per
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1,
Primary Containment Isolation Instrumentation, to its pre-ITS value
and adds a note to Table 3.3.6.1-1 of TS 3.3.6.1, Primary
Containment Isolation Instrumentation, that ensures, during
surveillance testing and normal operation, there will always be at
least one instrument monitoring for a small leak in all RWCU
locations. No changes in operating practices or physical plant
equipment are created as a result of this change. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different type of accident from any accident previously evaluated?
Response: No.
The proposed change restores the number of Required Channels Per
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1,
Primary Containment Isolation Instrumentation, to its pre-ITS value
and adds a note to Table 3.3.6.1-1 of TS 3.3.6.1, Primary
Containment Isolation Instrumentation, that ensures, during
surveillance testing and normal operation, there will always be at
least one instrument monitoring for a small leak in all RWCU
locations. No physical change in plant equipment will result from
this proposed change. Therefore, the proposed change does not create
the possibility of a new or different type 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 restores the number of Required Channels Per
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1,
Primary Containment Isolation Instrumentation, to its pre-ITS value
and adds a note to Table 3.3.6.1-1 of TS 3.3.6.1, Primary
Containment Isolation Instrumentation, that ensures, during
surveillance testing and normal operation, there will always be at
least one instrument monitoring for a small leak in all RWCU
locations. 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 G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279
NRC Section Chief: L. Raghavan.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 16, 2004.
Description of amendment request: The amendments would revise
Technical Specifications (TS) 3.5.2, ``Emergency Core Cooling System,''
TS 3.6.6, ``Containment Spray System,'' TS 3.6.17, ``Containment Valve
Injection Water System,'' TS 3.7.5, ``Auxiliary Feedwater System,'' TS
3.7.7, ``Component Cooling Water System,'' TS 3.7.8, ``Nuclear Service
Water System (NSWS),'' TS 3.7.10, ``Control Room Area Ventilation
System'' TS 3.7.12, ``Auxiliary Building Filtered Ventilation Exhaust
System,'' and TS 3.8.1, ``AC Sources-Operating'' for Catawba, Units 1
and 2. The revisions would allow for the ``A'' and ``B'' NSWS headers
to be take out of service for up to 14 days each for system upgrades.
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 operation of the facility in accordance with the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
Response: No.
The pipe repair project for the [nuclear service water system]
NSWS and proposed [technical specifications] TS changes have been
evaluated to assess their impact on normal operation of the systems
affected and to ensure that the design basis safety functions are
preserved. During the pipe repair the other NSWS train will be
operable and no major maintenance or testing will be done on the
operable train. The operable train will be protected to help ensure
it would be available if called upon.
This pipe repair project will enhance the long term structural
integrity in the NSWS system. This will ensure that the NSWS headers
maintain their integrity to ensure its ability to comply with design
basis requirements and increase the overall reliability for many
years.
The increased NSWS train unavailability as a result of the
implementation of this
[[Page 21455]]
amendment does involve a one time increase in the probability or
consequences of an accident previously evaluated during the time
frame the NSWS headers are out of service for pipe repair.
Considering this small time frame for the NSWS train outages with
the increased reliability and the decrease in unavailability of the
NSWS system in the future because of this project, the overall
probability or consequences of an accident previously evaluated will
decrease.
Therefore, because this is a temporary and not a permanent
change, the time averaged risk increase is acceptable. The increase
in the overall reliability of the NSWS along with the decreased
unavailability in the future because of the pipe repair project will
result in an overall increase in the safety of both Catawba units.
Therefore, the consequences of an accident previously evaluated
remains unaffected and there will be minimal impact on any accident
consequences.
2. Does operation of the facility in accordance with the
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated?
Response: No.
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed temporary TS changes do not
affect the basic operation of the [emergency core cooling system]
ECCS, [containment spray system] CSS, [containment valve injection
water system] CVIWS, NSWS, [auxiliary feedwater] AFW, [component
cooling water] CCW, [control room area ventilation system] [sic]
CRAVS, [auxiliary building filtered ventilation exhaust system]
ABFVES, or [emergency diesel generator] EDG systems. The only change
is increasing the required action time frame from 72 hours (ECCS,
CSS, NSWS, AFW, CCW, and EDG) or 168 hours (CVIWS, CRAVS and ABFVES)
to 336 hours. The train not undergoing maintenance will be operable
and capable of meeting its design requirements. Therefore, only the
redundancy of the above systems is affected by the extension of the
required action to 336 hours. During the project, contingency
measures will be in place to provide additional assurance that the
affected systems will be able to complete their design functions.
No new accident causal mechanisms are created as a result of NRC
approval of this amendment request. No changes are being made to the
plant, which will introduce any new accident causal mechanisms.
3. Does operation of the facility in accordance with the
proposed amendment involve a significant reduction in the margin of
safety?
Response: No.
Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The performance of these fission
product barriers will not be impacted by implementation of this
proposed temporary TS amendment. During the NSWS train outages, the
affected systems will still be capable of performing their required
functions and contingency measures will be in place to provide
additional assurance that the affected systems will be maintained in
a condition to be able to complete their design functions. No safety
margins will be impacted.
The probabilistic risk analysis conducted for this proposed
amendment demonstrated that the [core damage probability] CDP
associated with the outage extension is judged to be acceptable for
a one-time or rare evolution. Therefore, there is not 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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: March 8, 2005.
Description of amendment request: The proposed amendment would
enable the licensee to make changes to the Updated Safety Analysis
Report (USAR) to reflect the use of the non-single-failure-proof Fuel
Building Cask Handling Crane (FBCHC) for dry spent fuel cask component
lifting and handling operations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
Response: No.
The proposed amendment introduces no new mode of plant
operations and does not affect Structures, Systems, and Components
(SSCs) associated with power production, accident mitigation, or
safe plant shutdown. The SSCs affected by this proposed amendment
are the Fuel Building Cask Handling Crane (FBCHC), the spent fuel
storage canister, the spent fuel transfer cask, and the spent fuel
inside the storage canister. A hypothetical 30 ft. drop of a loaded
spent fuel shipping cask from the FBCHC is part of the River Bend
Station (RBS) current licensing basis. With the proposed spent fuel
transfer cask design and procedural changes implemented, the FCHC
will be used to lift and handle a fuel-loaded spent fuel transfer
cask of the same maximum weight and approximately the same
dimensions as previously evaluated in the RBS USAR. The proposed
amendment involves the use of redundant crane rigging during most
lateral moves with a loaded spent fuel transfer cask, which provides
temporary single-failure proof design features to provide protection
against an uncontrolled lowering of the load or load drop. In those
cases where the spent fuel transfer cask is not supported with
redundant rigging, certain hypothetical, non-mechanistic load drops
have been postulated and evaluated, with due consideration of the
use of impact limiters in some locations.
With this amendment, the probability of a loaded spent fuel
transfer cask drop is actually less likely than previously evaluated
because the capacity of the spent fuel multi-purpose canister [MPC]
(68 fuel assemblies) is larger than the capacity of the shipping
cask described in the current licensing basis (18 fuel assemblies),
which means that fewer casks will be required to be loaded, lifted,
and handled for a given population of spent fuel assemblies. The
consequences of the hypothetical spent fuel transfer cask load drops
on plant SSCs are bounded by those previously evaluated for a
shipping cask. That is, there is no significant damage to the Fuel
Building structure or any SSCs used for safe plant shutdown. New
analyses of hypothetical drops of a loaded transfer cask or canister
confirm that there is no release of radioactive material from the
storage canister and no unacceptable damage to the fuel, MPC, or
transfer cask.
The hypothetical drop of a spent fuel canister lid into an open,
fuel-filled canister in the spent fuel pool during fuel loading has
also been evaluated. Again, this hypothetical accident is no more
likely to occur than previously considered due to the higher
capacity of the spent fuel transfer cask over the spent fuel
shipping cask (i.e., fewer casks will need to be loaded for a given
number of fuel assemblies). The radiological consequences of this
event due to the potential damage of spent fuel assemblies in the
canister onto which the lid could be dropped have been evaluated.
While more total fuel assemblies could potentially be damaged from a
spent fuel canister lid drop compared to that assumed for the fuel
handling accident described in the RBS current licensing basis, the
significantly longer decay time of the spent fuel assemblies in the
canister results in a much smaller source term, such that the
existing fuel handling accident described in USAR Section 15.7.4
provides a bounding evaluation for the radiological consequences MPC
lid drop. There is no rearrangement of the fuel or deformation of
the fuel basket in the canister such that a critical geometry is
created as a result of an MPC lid drop.
The likelihood of a spent fuel canister lid drop due to the
failure of a crane component due to overload is very unlikely
because the rated load of the crane (250,000 lbs) is
[[Page 21456]]
approximately 16 times the weight of components lifted to install
the canister lid.
2. Will operation of the facility in accordance with this
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated?
Response: No.
The proposed amendment introduces no new mode of plant
operations and does not affect SSCs associated with power
production, accident mitigation, or safe plant shutdown. The SSCs
affected by this proposed amendment are the non-single-failure-proof
FBCHC, the spent fuel canister, the spent fuel transfer cask, and
the spent fuel inside the canister. The design function of the FBCHC
is not changed. 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. The
proposed amendment creates a new initiator of two accidents
previously evaluated and caused by the non-mechanistic single
failure of a component in the FBCHC load path.
The current licensing basis accidents for which new initiators
are created by this amendment are the spent fuel shipping cask drop
and the fuel handling accident. The RBS current licensing basis
includes evaluations of the consequences of a spent fuel shipping
cask drop and the consequences of the drop of a spent fuel assembly
into the reactor core shortly after shutdown and reactor head
removal. The new initiators include the drop of a spent fuel
transfer cask of the same maximum weight and approximately the same
dimensions as the shipping cask, and the drop of a spent fuel
canister lid into an open, fuel filled canister in the spent fuel
pool. Both of these new initiators create hypothetical accidents
that are comparable in consequences to those previously evaluated.
For the drop of a spent fuel transfer cask, the consequences are
bounded by the current licensing basis analysis of the spent fuel
shipping cask drop. That is, there is no significant damage to the
Fuel Building structure or any SSCs used for safe plant shutdown,
and there is no release of radioactive material. New analyses of the
drop of a loaded transfer cask confirm that there is no release of
radioactive material from the storage canister and no unacceptable
damage to the fuel, MPC, or transfer cask.
For the drop of the spent fuel canister lid, the significantly
longer decay time of the spent fuel assemblies in the canister
compared to a spent fuel assembly in a recently shutdown reactor
results in doses to the public that are less than the previously
analyzed fuel handling accident. There is no rearrangement of the
fuel in the canister such that a critical geometry is created as a
result of an MPC lid drop.
3. Will operation of the facility in accordance with this
proposed amendment involve a significant reduction in a margin of
safety?
Response: No.
The proposed amendment introduces no new mode of plant
operations and does not affect SSCs associated with power
production, accident mitigation, or safe plant shutdown. The SSCs
affected by this proposed amendment are the non-single-failure-proof
FBCHC, the spent fuel storage canister, the spent fuel transfer
cask, and the spent fuel inside the canister. Therefore, this
amendment does not affect the reactor or fuel during power
operations, the reactor coolant pressure boundary, or primary or
secondary containment. All activities associated with this amendment
occur in the Fuel Building or in the adjacent outdoor truck bay
area. The design function of the FBCHC is not changed. The proposed
changes to plant operating procedures needed to implement dry spent
fuel storage at RBS do not exceed or alter a design basis or safety
limit associated with plant operation, accident mitigation, or safe
shutdown. The FBCHC is used to lift and handle the spent fuel
canister lid over spent fuel in the canister while in the spent fuel
pool, and to lift and handle the spent fuel transfer cask, both when
it is empty and after it is loaded with spent fuel in the spent fuel
pool.
This proposed amendment results in a net safety benefit because
a larger capacity cask is being used to move spent fuel out of the
spent fuel pool that was previously evaluated (68 fuel assemblies
versus 18 fuel assemblies), while maintaining the same maximum
analyzed cask weight described in the USAR. This yields fewer casks
to be loaded, fewer heavy load lifts, and, as a result, fewer
opportunities for events such as load drops. Because the maximum
weight of the loaded spent fuel transfer cask is the same as that
assumed for the shipping cask and for which the FBCHC was designed,
all design safety margins for use of the FBCHC remain unchanged. The
rated capacity of the FBCHC is approximately 16 times that of
components lifted to place the spent fuel canister lid, yielding
significant safety margins for that particular lift.
Based on the above review, it is concluded that: (1) the
proposed amendment does not constitute a significant hazards
consideration as defined by 10 CFR 50.92; and (2) there is
reasonable assurance that the health and safety of the public will
not be endangered by the proposed amendment; and (3) this action
will not result in a condition which significantly alters the impact
of the station on the environment as described in the NRC Final
Environmental Impact Statement.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Allen G. Howe.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1
and 2, Ogle County, Illinois
Docket No. 50-237, Dresden Nuclear Power Station, Unit 2, Grundy
County, Illinois
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and
2, LaSalle County, Illinois
Date of amendment request: February 25, 2005.
Description of amendment request: The proposed change would delete
the applicable sections of the Facility Operating Licenses (FOLs); NPF-
72, NPF-77, NPF-37, NPF-66, DPR-19, NPF-11, and NPF-18, respectively;
which require Exelon Generation Company, LLC, to report violations of
the requirements contained in Section 2.C of the Braidwood Station,
Units 1 and 2, and Byron Station, Units 1 and 2 FOLs; Section 2.C of
the Dresden Nuclear Power Station, Unit 2, renewed FOL; and Sections
2.C and 2.E of the LaSalle County Station, Units 1 and 2, FOLs. The
proposed change will reduce unnecessary regulatory burden and will
allow Exelon to take full advantage of the revisions to Title 10, Code
of Federal Regulations (10 CFR), Section 50.72, ``Immediate
notification requirements for operating nuclear power reactors,'' and
10 CFR 50.73, ``Licensee event report system.''
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 involves an administrative change only. The
proposed change does not involve the modification of any plant
equipment or affect plant operation. The proposed change will have
no impact on any safety related structures, systems or components.
The reporting requirement section of the FOL is not required because
the requirements are either adequately addressed by 10 CFR 50.72 and
10 CFR 50.73, or other regulatory requirements, or are not required
based on the nature of the Condition.
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
[[Page 21457]]
accident from any accident previously evaluated?
Response: No.
The proposed change has no impact on the design, function or
operation of any plant structure, system or component. The proposed
change is administrative in nature and does not affect plant
equipment or accident analyses. The reporting requirement section of
the FOL is not required because the requirements are either
adequately addressed by 10 CFR 50.72 and 10 CFR 50.73, or other
regulatory requirements, or are not required based on the nature of
the Condition.
Therefore, the proposed change 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 proposed change is administrative in nature, does not negate
any existing requirement, and does not adversely affect existing
plant safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there is no change being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by deletion of the
reporting requirement that is adequately addressed elsewhere.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Unit Nos. 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: February 25, 2005.
Description of amendment request: The proposed change would delete
the applicable sections of the Limerick Generating Station, Units 1 and
2, Facility Operating Licenses (FOLs), NPF-39 and NPF-85, which require
Exelon Generation Company, LLC, (Exelon), to report violations of the
requirements contained in Section 2.C of these licenses. The proposed
change will reduce unnecessary regulatory burden and will allow AmerGen
to take full advantage of the revisions to Title 10, Code of Federal
Regulations (10 CFR), Section 50.72, ``Immediate notification
requirements for operating nuclear power reactors,'' and 10 CFR 50.73,
``Licensee event report system.''
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 involves an administrative change only. The
proposed change does not involve the modification of any plant
equipment or affect plant operation. The proposed change will have
no impact on any safety related structures, systems or components.
The reporting requirement section of the FOL is not required because
the requirements are either adequately addressed by 10 CFR 50.72 and
10 CFR 50.73, or other regulatory requirements, or are not required
based on the nature of the Condition.
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 has no impact on the design, function or
operation of any plant structure, system or component. The proposed
change is administrative in nature and does not affect plant
equipment or accident analyses. The reporting requirement section of
the FOL is not required because the requirements are either
adequately addressed by 10 CFR 50.72 and 10 CFR 50.73, or other
regulatory requirements, or are not required based on the nature of
the Condition.
Therefore, the proposed change 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 proposed change is administrative in nature, does not negate
any existing requirement, and does not adversely affect existing
plant safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there is no change being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by deletion of the
reporting requirement that is adequately addressed elsewhere.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Darrell J. Roberts.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: February 11, 2005.
Description of amendment request: The proposed changes would modify
the BVPS-1 and 2 Technical Specifications (TSs) to implement the
relaxed axial offset control (RAOC) and FQ surveillance
methodologies. These methodologies are used to reduce operator action
required to maintain conformance with power distribution control TSs,
and increase the ability to return to power after a plant trip while
still maintaining margin to safety limits under all operating
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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed changes will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes do not initiate an accident. Evaluations
and analyses of accidents, which are potentially affected by the
parameters and assumptions, associated with the RAOC and
FQ(Z) methodologies have shown that all design standards
and applicable safety criteria will continue to be met. The
consideration of these changes does not result in a situation where
the design, material, or construction standards that were applicable
prior to the change are altered. Therefore, the proposed changes
will not result in any additional challenges to plant equipment that
could increase the probability of any previously evaluated accident.
The proposed changes associated with the RAOC and
FQ(Z) methodologies do not affect plant systems such that
their function in the control of radiological consequences is
adversely affected. The actual plant configuration, performance of
systems, or initiating event mechanisms are not being
[[Page 21458]]
changed as a result of the proposed changes. The design standards
and applicable safety criteria limits will continue to be met,
therefore, fission barrier integrity is not challenged. The proposed
changes associated with the RAOC and FQ(Z) methodologies
have been shown not to adversely affect the plant response to
postulated accident scenarios. The proposed changes will therefore
not affect the mitigation of the radiological consequences of any
accident described in the Updated Final Safety Analysis Report
(UFSAR).
Therefore the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed changes do not challenge the performance or integrity
of any safety-related system. The possibility for a new or different
type of accident from any accident previously evaluated is not
created since the proposed change does not result in a change to the
design basis of any plant structure, system or component. Evaluation
of the effects of the proposed changes has shown that all design
standards and applicable safety criteria continue to be met.
Equipment important to safety will continue to operate as
designed and component integrity will not be challenged. The
proposed changes do not result in any event previously deemed
incredible being made credible. The proposed changes will not result
in conditions that are more adverse and will not result in any
increase in the challenges to safety systems.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed changes will not involve a
significant reduction in a margin of safety. The proposed changes
will assure continued compliance within the acceptance limits
previously reviewed and approved by the NRC for RAOC and
FQ(Z) methodologies. All of the appropriate acceptance
criteria for the various analyses and evaluations will continue to
be met.
The impact associated with the implementation of RAOC on peak
cladding temperature (PCT) has been evaluated for the planned
extended power uprate. This evaluation has determined that
implementation of RAOC at the extended power uprate power level will
not result in a significant reduction in a margin of safety for
either unit.
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: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: February 17, 2005.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.7.7.1, ``Control Room Emergency
Habitability Systems'' (BVPS-1), and TS 3.7.7, ``Control Room Emergency
Air Cleanup and Pressurization System'' (BVPS-2), by dividing each
specification into two specifications, addressing control room
emergency ventilation and control room air cooling functions
separately. Other minor changes are proposed to improve consistency
with the Standard TSs and consistency between BVPS-1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No.
The proposed changes do not adversely affect accident initiators
or precursors or alter the design assumptions, conditions or
configuration of the facility. The proposed changes do not alter or
prevent the ability of structures, systems, or components to perform
their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits. The proposed
change revises the TSs for the control room ventilation systems
which are mitigating systems designed to minimize inleakage, to
filter the control room atmosphere and to provide heat removal for
the control room envelope. These functions maintain the control room
temperature within design limits and protect the control room
personnel following accidents previously analyzed. The proposed
changes do not alter or reduce the capability of the affected
systems to maintain the control room temperature and protect the
control room personnel consistent with the assumptions of the
applicable safety analyses. Therefore, the probability of any
accident previously evaluated is not significantly increased. The
proposed change continues to assure [that] adequate system and
component testing is performed to verify the operability of the
control room habitability systems to ensure mitigation features are
capable of performing the assumed functions. Therefore, the
consequences of any accident previously evaluated are not
significantly increased.
Therefore, it is concluded that the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No.
The proposed changes will not adversely impact the accident
analysis. The changes will not alter the requirements of the control
room ventilation systems or their functions during accident
conditions. No new or different accidents result from the
application of the revised TS requirements. The changes do not
involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The changes
do not alter assumptions made in the safety analyses. The proposed
changes are consistent with the safety analyses assumptions and
current plant operating practices.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not affected by these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis for an unacceptable period of time without compensatory
measures. The proposed changes do 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: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
[[Page 21459]]
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: February 28, 2005.
Description of amendment request: The proposed amendments would
allow the use of the Small Break Loss of Coolant Accident (SBLOCA)
methodology described in Westinghouse WCAP 10054-P-A Addendum 2
Revision 1, ``Addendum to the Westinghouse Small Break emergency core
cooling system (ECCS) Evaluation Model Using the NOTRUMP Code: Safety
Injection into the Broken Loop and COSI Condensation Model'' dated July
1997. This revised methodology determines the core response following a
SBLOCA event and will be used to assure compliance with the post Loss
of Coolant Accident (LOCA) acceptance criteria specified in 10 CFR
50.46.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment will change the Prairie Island Nuclear
Generating Plant licensing basis by allowing the use of the approved
NOTRUMP SBLOCA Evaluation Model described in Westinghouse WCAP
10054-P-A Addendum 2 Revision 1, ``Addendum to the Westinghouse
Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety
Injection into the Broken Loop and COSI Condensation Model''.
The methodology used to perform small break loss of coolant
accident (SBLOCA) analyses is not an accident initiator, thus
changing the methodology does not increase the probability of an
accident.
The fuel heat-up results generated by the proposed methodology
will be utilized to demonstrate that the loss of coolant accident
(LOCA) criteria for design basis for fission product barriers as
described in 10 CFR Part 50.46 are not exceeded. The proposed
methodology does not alter the nuclear reactor core, reactor coolant
system, or equipment used directly in mitigation of a Small Break
LOCA, thus radioactive releases due to a SBLOCA accident are not
affected by the proposed change in analysis methodology. Therefore,
this change does not increase the consequences of an accident
previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment will change the Prairie Island Nuclear
Generating Plant licensing basis by allowing the use of the approved
NOTRUMP SBLOCA Evaluation Model described in Westinghouse WCAP
10054-P-A Addendum 2 Revision 1, ``Addendum to the Westinghouse
Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety
Injection into the Broken Loop and COSI Condensation Model''.
The analysis of a SBLOCA accident using the proposed methodology
does not alter the nuclear reactor core, reactor coolant system, or
equipment used directly in mitigation of a Small Break LOCA.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment will change the Prairie Island Nuclear
Generating Plant licensing basis by allowing the use of the approved
NOTRUMP SBLOCA Evaluation Model described in Westinghouse WCAP
10054-P-A Addendum 2 Revision 1, ``Addendum to the Westinghouse
Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety
Injection into the Broken Loop and COSI Condensation Model''.
The methodology in the proposed licensing basis change has
previously been reviewed and approved by the Nuclear Regulatory
Commission as a conservative methodology. The Prairie Island
configuration is representative of the modeling used in the
methodology. Therefore, the proposed licensing basis change will
result in a conservative calculation of fuel conditions following a
SBLOCA event. This will ensure that there is no reduction in the
margin of safety for Prairie Island SBLOCA analyses that utilize
this methodology.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: March 31, 2005.
Description of amendment request: The proposed amendment will
increase the licensed power level to 1522 megawatts thermal (MWt) or
1.50 percent greater than the current power level of 1500 MWt. The
requested increase in licensed rated power is the result of a
measurement uncertainty recapture (MUR) power uprate. The information
provided in support of this request is based on the NRC's Regulatory
Issue Summary 2002-03, ``Guidance on the Content of Measurement
Uncertainty Recapture Power Uprate Applications,'' dated January 31,
2002.
On July 18, 2003, the licensee submitted, and the NRC subsequently
approved, an MUR power uprate amendment to increase the licensed power
level to 1524 MWt or 1.6 percent greater than the current level of 1500
MWt. Problems during implementation resulted in the submission of an
exigent license amendment request (LAR), which returned the licensed
power to its original level (1500 MWt). The current LAR references the
analysis from the July 18, 2003 submittal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Response: No.
There are no changes as a result of the MUR power uprate to the
design or operation of the plant that could affect system,
component, or accident functions. All systems and components
function as designed and the performance requirements have been
evaluated and found to be acceptable.
The reduction in power measurement uncertainty allows for safety
analyses to continue to be used without modification. This is
because those safety analyses were performed or evaluated at 102% of
1500 MWt (1530 MWt) or higher. Analyses at these power levels
support a core power level of 1522 MWt with a measurement
uncertainty of 0.5%. Radiological consequences of USAR [Updated
Safety Analysis Report] Chapter 14 accidents were assessed
previously using the alternate source term methodology (Reference
10.2 [Agencywide Documents Access Management System accession number
ML013410095]). These analyses were performed at 102% of 1500 MWt
(1530 MWt) and continue to be bounding. Updated Safety
[[Page 21460]]
Analysis Report (USAR) Chapter 14 analyses and accident analyses
continue to demonstrate compliance with the relevant accident
analyses' acceptance criteria. Therefore, there is no significant
increase in the consequences of any accident previously evaluated.
The primary loop components (reactor vessel, reactor internals,
control element drive mechanisms, loop piping and supports, reactor
coolant pumps, steam generators, and pressurizer) were evaluated at
an uprated core power level of 1524 MWt and continue to comply with
their applicable structural limits. These analyses also demonstrate
the components will continue to perform their intended design
functions. Changing the heatup and cooldown curves is based on
uprated fluence values. This does not have a significant effect on
the reactor vessel integrity. Thus, there is no significant increase
in the probability of a structural failure of the primary loop
components. The LBB [leak before break] analysis conclusions remain
valid and the breaks previously exempted from structural
consideration remain unchanged.
All of the NSSS [nuclear steam system supplier] systems will
continue to perform their intended design functions during normal
and accident conditions. The auxiliary systems and components
continue to comply with the applicable structural limits and will
continue to perform their intended functions. The NSSS/BOP [nuclear
steam system supplier/balance of plant] interface systems were
evaluated at 1522 MWt and will continue to perform their intended
design functions. Plant electrical equipment was also evaluated and
will continue to perform their intended functions. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Response: No.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of the proposed change. All
systems, structures, and components previously required for the
mitigation of an event remain capable of fulfilling their intended
design function at the uprated power level. The proposed change has
no adverse effects on any safety related systems or component and
does not challenge the performance or integrity of any safety
related system. 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 change does not involve a significant reduction
in a margin of safety.
Response: No.
Operation at 1522 MWt core power does not involve a significant
reduction in the margin of safety. The current accident analyses
have been previously performed with a 2% power measurement
uncertainty or at uprated core powers that exceed the MUR uprated
core power. System and component analyses have been completed at the
MUR uprated core power conditions. Analyses of the primary fission
product barriers at uprated core powers have concluded that all
relevant design basis criteria remain satisfied in regard to
integrity and compliance with the regulatory acceptance criteria. As
appropriate, all evaluations have been both reviewed and approved by
the NRC, or are currently under review (the proposed Pressure-
Temperature Limits Report). Therefore, the proposed change does not
involve a significant reduction in 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,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: December 28, 2004.
Description of amendment requests: The proposed amendments would
relocate reactor coolant system related cycle-specific parameters from
the Technical Specifications to the Core Operating Limits Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes are programmatic and administrative in
nature, which do not physically alter safety related systems, nor
affect the way in which safety related systems perform their
functions. More specific requirements regarding the safety limits
(i.e., departure from nucleate boiling ratio limit and peak fuel
centerline temperature limit) are being imposed in Technical
Specification (TS) 2.1.1, ``Reactor Core SLs [Safety Limits],''
which replace the reactor core safety limits figure and are
consistent with the values stated in the Final Safety Analysis
Report Update (FSARU). The proposed changes remove cycle-specific
parameters from TS 3.4.1 and relocate them to the Core Operating
Limits Report (COLR), which do not change the plant design or affect
system operating parameters. In addition, the minimum limit for
reactor coolant system (RCS) total flow rate is being retained in TS
3.4.1 to assure that a lower flow rate than reviewed by the NRC will
not be used. The proposed changes do not, by themselves, alter any
of the parameters. The removal of the cycle-specific parameters from
the TS does not eliminate existing requirements to comply with the
parameters.
The proposed changes to TS 5.6.5b to reference only the topical
report number and title for three of the topical reports do not
alter the use of the analytical methods used to determine core
operating limits that have been reviewed and approved by the NRC.
This method of referencing topical reports would allow the use of
current topical reports to support limits in the COLR without having
to submit a request for an amendment to the operating license.
Implementation of revisions to these topical reports would still be
reviewed in accordance with 10 CFR 50.59 and, where required,
receive NRC review and approval.
Although the relocation of the cycle-specific parameters to the
COLR would allow revision of the affected parameters without prior
NRC approval, there is no significant effect on the probability or
consequences of an accident previously evaluated. Future changes to
the COLR parameters could result in event consequences which are
either slightly less or slightly more severe than the consequences
for the same event using the present parameters. The differences
would not be significant and would be bounded by the existing
requirement of TS 5.6.5c to meet the applicable limits of the safety
analyses.
The cycle-specific parameters being transferred from the TS to
the COLR will continue to be controlled under existing programs and
procedures. The FSARU accident analyses will continue to be examined
with respect to changes in the cycle-dependent parameters obtained
using NRC reviewed and approved reload design methodologies,
ensuring that the transient evaluation of new reload designs are
bounded by previously accepted analyses. This examination will
continue to be performed pursuant to 10 CFR 50.59 requirements,
ensuring that future reload designs will not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Additionally, the proposed changes do not
allow for an increase in plant power levels, do not increase the
production, nor alter the flow path or method of disposal of
radioactive waste or byproducts. Therefore, the proposed changes do
not change the type or increase the amount of any effluents released
offsite.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes that retain the minimum limit for RCS total
flow rate in the TS, and that relocate certain cycle-specific
parameters from the TS to the COLR, thus removing the requirement
for prior NRC approval of revisions to those parameters, do not
involve a physical change to the plant. No new equipment is being
introduced, and
[[Page 21461]]
installed equipment is not being operated in a new or different
manner. There are no changes being made to the parameters within
which the plant is operated, other than their relocation to the
COLR. There are no set points affected by the proposed changes at
which protective or mitigative actions are initiated. The proposed
changes will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. No alteration in the procedures which ensure the plant
remains within analyzed limits is being proposed, and no change is
being made to the procedures relied upon to respond to an off-normal
event. As such, no new failure modes are being introduced.
The proposed changes to reference only the topical report number
and title do not alter the use of the analytical methods used to
determine core operating limits that have been reviewed and approved
by the NRC. This method of referencing topical reports would allow
the use of current topical reports to support limits in the COLR
without having to submit a request for an amendment to the operating
license. Implementation of revisions to topical reports would still
be reviewed in accordance with 10 CFR 50.59 and, where required,
receive NRC review and approval.
Relocation of cycle-specific parameters has no influence or
impact on, nor does it contribute in any way to the possibility of a
new or different kind of accident. The relocated cycle-specific
parameters will continue to be calculated using the NRC reviewed and
approved methodology. The proposed changes do not alter assumptions
made in the safety analysis, and operation within the core operating
limits will continue.
Therefore, the proposed changes do not create a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety is established through equipment design,
operating parameters, and the set points at which automatic actions
are initiated. The proposed changes do not physically alter safety-
related systems, nor do they affect the way in which safety-related
systems perform their functions. The set points at which protective
actions are initiated are not altered by the proposed changes.
Therefore, sufficient equipment remains available to actuate upon
demand for the purpose of mitigating an analyzed event. As the
proposed changes to relocate cycle-specific parameters to the COLR
will not affect plant design or system operating parameters, there
is no detrimental impact on any equipment design parameter, and the
plant will continue to operate within prescribed limits.
The development of cycle-specific parameters for future reload
designs will continue to conform to NRC reviewed and approved
methodologies, and will be performed pursuant to 10 CFR 50.59 to
assure that the plant operates within cycle-specific parameters.
The proposed changes to reference only the topical report number
and title do not alter the use of the analytical methods used to
determine core operating limits that have been reviewed and approved
by the NRC. This method of referencing topical reports would allow
the use of current NRC-approved topical reports to support limits in
the COLR without having to submit a request for an amendment to the
operating license. Implementation of revisions to topical reports
would still be reviewed in accordance with 10 CFR 50.59 and, where
required, receive NRC review and approval.
Therefore, the proposed changes do 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 requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Robert A. Gramm.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: December 31, 2004.
Description of amendment requests: The proposed amendments would
revise Technical Specification 3.4.10, ``Pressurizer Safety Valves'' to
add a separate Action and associated Completion Times for one or more
inoperable pressurizer safety valves for the condition where the valves
are inoperable solely due to loop seal temperatures being outside of
design limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This proposed change revises Technical Specification (TS)
3.4.10, ``Pressurizer Safety Valves,'' to add a separate Action and
associated Completion Times (CTs) for one or more inoperable
pressurizer safety valves (PSV) for the condition where the valves
are inoperable solely due to loop seal temperatures being outside of
design limits. Currently, when a PSV is in such a condition, it is
conservatively declared inoperable and TS 3.4.10 Condition A is
entered which has a CT of 15 minutes. A CT of 15 minutes normally
provides insufficient time for restoring a PSV loop seal temperature
to within limits. The new Action will provide CTs of 12 hours for
exceeding the high temperature limit and 24 hours (MODES 1 and 2) or
72 hours (MODES 3 and 4) for exceeding the low temperature limit. In
addition, two new PSV loop seal temperature surveillance
requirements are proposed to assist in assuring PSV operability.
Loop seals are provided in the PSV inlet piping to maintain PSV
body temperature within vendor recommended limits. This prevents PSV
seat leakage that can result from spring relaxation with increased
temperature. However, the water in the loop seals must be maintained
at or above a minimum temperature to allow it to flash to steam when
a PSV lifts. Because of the low density and low mass flow rate, PSV
steam relief imposes minimal loading on the discharge piping
ensuring acceptable pipe stresses. However, if cooler water is
maintained in the loop seals, it may not flash completely, and a
water and steam mixture could be discharged when a PSV lifts.
Because of the higher density and higher mass flow rate, PSV relief
of water and steam could impose increased loading and could result
in unacceptably high pipe stresses on the discharge piping which
could render the PSVs inoperable and/or damage the discharge piping.
The concern with the PSV opening during liquid relief conditions
or with the loop seal temperature outside design limits, is the
ability to ensure the valve reseats properly and no leakage occurs
after the valve closes. However, even under liquid relief
conditions, PSVs are still capable of providing their required
relief capacity.
Failure of the PSV to reseat following discharge would result in
an unisolable reactor coolant system leak. The consequences of such
a leak are bounded by existing Final Safety Analysis Report Update
(FSARU) accident analyses. Probabilistic risk assessment methods and
a deterministic analysis have been utilized to determine there is no
significant increase in core damage frequency or large early release
frequency.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Failure of one or more PSVs to reseat following discharge would
result in an unisolable reactor coolant system leak. The
consequences of such a leak are bounded by existing FSARU accident
analyses and no new failure modes are introduced.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is based upon both a deterministic
evaluation and a risk-informed assessment.
The deterministic evaluation concluded that even with the loop
seal temperature outside of design limits, causing one or more PSVs
to be declared inoperable, the PSVs
[[Page 21462]]
would still lift on demand to perform their safety function. Failure
of one or more PSVs to reseat following discharge, resulting in an
unisolable reactor coolant system leak, is an event bounded by
existing FSARU accident analyses.
The risk assessment performed to support this license amendment
request concluded that the increase in plant risk is small and
consistent with the NRC's Safety Goal Policy Statement, ``Use of
Probabilistic Risk Assessment Methods in Nuclear Activities: Final
Policy Statement,'' Federal Register, Volume 60, p. 42622, August
16, 1995 and guidance contained in of Regulatory Guides (RG) 1.174,
``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing
Basis,'' dated July 1998 and RG 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decisionmaking: Technical Specifications,''
dated August 1998.
Together, the deterministic evaluation and the risk-informed
assessment provide high assurance that the PSVs will meet their
design requirements.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Robert A. Gramm.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: March 11, 2005.
Description of amendment requests: The proposed amendment would
modify Technical Specification (TS) 5.5.9, ``Steam Generator (SG) Tube
Surveillance Program,'' and 5.6.10, ``Steam Generator (SG) Tube
Inspection Report,'' to allow the use of the SG tube W star (W*)
alternate repair criteria (ARC) on a permanent basis. The W* ARC allows
axial primary water stress corrosion cracking indications in the
Westinghouse explosive tube expansion (WEXTEX) region to remain in
service if the indication is located below the bottom of the WEXTEX
transition. In addition, TS 5.6.10.d for NRC notification requirements
of the voltage-based ARC would be revised.
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.
Of the various accidents previously evaluated, the permanent use
of the steam generator (SG) tube W star (W*) alternate repair
criteria (ARC) only affects the steam generator tube rupture (SGTR)
accident evaluation and the postulated main steam line break (MSLB)
accident evaluation. Loss-of-coolant accident (LOCA) conditions
cause a compressive axial load to act on the tube. Therefore, since
the LOCA tends to force the tube into the tubesheet rather than pull
it out, it is not a factor in this evaluation.
For the SGTR accident, the required structural margins of the SG
tubes will be maintained by the presence of the tubesheet. Tube
rupture is precluded for cracks in the Westinghouse explosive tube
expansion (WEXTEX) region due to the constraint provided by the
tubesheet. Therefore, Regulatory Guide (RG) 1.121, ``Bases for
Plugging Degraded PWR Steam Generator Tubes,'' margins against burst
are maintained for both normal and postulated accident conditions.
WCAP-14797-P, Revision 2, defines a length, W*, of degradation-
free expanded tubing that provides the necessary resistance to tube
pullout due to the pressure-induced forces (with applicable safety
factors applied). The W* length supplies the necessary resistive
force to preclude pullout loads under both normal operating and
accident conditions. The contact pressure results from the WEXTEX
expansion process, thermal expansion mismatch between the tube and
tubesheet and from the differential pressure between the primary and
secondary side as offset at higher tubesheet elevations by bow of
the tubesheet. The proposed changes do not affect other systems,
structures, components, or operational features. Therefore, the
proposed change results in no significant increase in the
probability of the occurrence of an SGTR or MSLB accident.
The consequences of an SGTR accident are affected by the
primary-to-secondary leakage flow during the accident. Primary-to-
secondary leakage flow through a postulated broken tube is not
affected by the proposed changes since the tubesheet enhances the
tube integrity in the region of the WEXTEX expansion by precluding
tube deformation beyond its initial expanded outside diameter. The
resistance to both tube rupture and collapse is strengthened by the
tubesheet in that region. At normal operating pressures, leakage
from primary water stress corrosion cracking in the W* length is
limited by both the tube-to-tubesheet crevice and the limited crack
opening permitted by the tubesheet constraint. No leakage has been
observed in any in situ test of W* indications to date.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region.
MSLB leakage is limited by leakage flow restrictions resulting
from the crack and tubesheet that provide a restricted leakage path
and also limit the degree of crack face opening compared to free
span indications. The total leakage, that is, the combined leakage
for all such tubes, plus the combined leakage developed by any other
ARC and non-ARC degradation, is limited to less than the maximum
allowable MSLB accident dose analysis leak rate limit, such that
offsite dose is maintained less than the guideline value in Title 10
to the Code of Federal Regulations (10 CFR) Part 100 and control
room dose is maintained less than the value in General Design
Criterion (GDC) 19 of Appendix A to 10 CFR Part 50. In addition, the
editorial changes made to Technical Specifications 5.5.9 and 5.6.10
have no impact on the MSLB leakage [and the SGTR].
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different accident from any accident previously evaluated?
Response: No.
The proposed changes do not introduce any changes or mechanisms
that create the possibility of a new or different kind of accident.
Tube bundle integrity is expected to be maintained for all plant
conditions upon continued implementation of the W* ARC.
Axial indications left in service shall have the upper crack tip
below the top of the tubesheet (TTS) by at least the value of the
nondestructive examination (NDE) uncertainty and crack growth
allowance, such that at the end of the subsequent operating cycle
the entire crack remains below the tubesheet secondary face, thereby
minimizing the potential for free span cracking and demonstrating
that an acceptable level of risk is maintained for tubes returned to
service under W* ARC. This repair criterion is in addition to
ensuring that the upper crack tip is located below the bottom of the
WEXTEX transition by at least the NDE measurement uncertainty.
Condition monitoring will verify that all tube cracks returned to
service under W* ARC remain below the TTS, including an allowance
for NDE uncertainty.
These changes do not introduce any new equipment or any change
to existing equipment. No new effects on existing equipment are
created nor are any new malfunctions introduced.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes maintain the required structural margins of
the SG tubes for both normal and accident conditions. RG 1.121 is
used as the basis in the development of the W* ARC for determining
that SG tube integrity considerations are maintained within
acceptable limits. RG 1.121 describes a method acceptable to the NRC
staff for meeting General Design Criteria 14, 15, 31,
[[Page 21463]]
and 32 by reducing the probability and consequences of an SGTR. RG
1.121 concludes that by determining the limiting safe conditions of
tube wall degradation beyond which tubes with unacceptable cracking,
as established by inservice inspection, should be removed from
service or repaired, the probability and consequences of a SGTR are
reduced. This RG uses safety factors on loads for tube-burst that
are consistent with the requirements of Section III of the ASME
Code.
For primarily axially oriented cracking located within the
tubesheet, tubeburst is precluded due to the presence of the
tubesheet. WCAP-14797-P, Revision 2, defines a length, W*, of
degradation free expanded tubing that provides the necessary
resistance to tube pullout due to the pressure induced forces (with
applicable safety factors applied). Application of the W* ARC will
preclude unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining MSLB leakage due to
indications within the tubesheet region provides for large margins
between calculated and actual leakage values. In addition, the total
leakage, including leakage due to use of other ARC, is maintained
below the maximum allowable MSLB accident dose analysis leak rate
limit, such that offsite dose is maintained less than the guideline
value in 10 CFR Part 100 and control room dose is maintained less
than the value in GDC 19. In addition, the editorial changes made to
Technical Specifications 5.5.9 and 5.6.10 have no impact on the
determination of MSLB leakage [and the SGTR].
Plugging of the SG tubes reduces the reactor coolant flow margin
for core cooling. Continued implementation of W* ARC will result in
maintaining the margin of flow that may have otherwise been reduced
by tube plugging.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above evaluation, PG&E [Pacific Gas and Electric
Company] concludes that the proposed change presents no significant
hazards consideration under the standards set forth in 10 CFR
50.92(c), and accordingly, a finding of ``no significant hazards
consideration'' is justified.
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: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Robert Gramm.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: November 9, 2004.
Description of amendment request: The proposed amendments would
change the SSES 1 and 2 Technical Specifications (TSs) 3.8.4, ``DC
Sources-Operating,'' 3.8.5, ``DC Sources-Shutdown,'' 3.8.6, ``Battery
Cell Parameters,'' and add a new TS Section, 5.5.13, ``Battery
Monitoring and Maintenance Program.'' These changes are consistent with
Technical Specifications Change Traveler (TSTF) 360, Revision 1 to
request new actions with increased completion times for an inoperable
battery chargers and alternate battery charger testing criteria for
limiting condition for operation (LCO) 3.8.4 and LCO 3.8.5.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed changes restructure the Technical
Specifications (TSs) for the DC Electrical Power Systems. The
proposed changes add actions to specifically address battery charger
inoperability. This change will rely upon the capability of
providing the battery charger function by an alternate means (e.g.,
a 125 volts direct current (VDC) portable battery charger or a 250
VDC portable battery charger) to justify the proposed Completion
Times. The DC electrical power systems, including associated battery
chargers, are not initiators to any accident sequence analyzed in
the Final Safety Analysis Report (FSAR). Operation in accordance
with the proposed TS ensures that the DC electrical power systems
are capable of performing functions as described in the FSAR.
Therefore the mitigative functions supported by the DC Power Systems
will continue to provide the protection assumed by the analysis.
The relocation of preventive maintenance surveillance, and
certain operating limits and actions to a newly-created, licensee-
controlled TS 5.5.13, ``Battery Monitoring and Maintenance
Program,'' will not challenge the ability of the DC electrical power
systems to perform their design functions. The maintenance and
monitoring required by current TS, which are based on industry
standards, will continue to be performed. In addition, the DC Power
Systems are within the scope of 10 CFR 50.65, ``Requirements for
Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants,'' which will ensure the control of maintenance activities
associated with the DC electrical power systems. The integrity of
fission product barriers, plant configuration, and operating
procedures as described in the FSAR will not be affected by the
proposed changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed changes involve restructuring the TS for the DC
electrical power systems. These changes will rely upon the
capability of providing the battery charger function by an alternate
means to justify the proposed completion times when a normal battery
charger is inoperable. The DC electrical power systems, which
include the associated battery chargers, are not initiators to any
accident sequence analyzed in the FSAR. Rather, the DC electrical
power systems are used to supply equipment used to mitigate an
accident. These mitigative functions, supported by the DC electrical
power systems are not affected by these changes and they will
continue to provide the protection assumed by the safety analysis
described in the FSAR. There are no new types of failures or new or
different kinds of accidents or transients that could be created by
these changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The margin of safety is established through equipment
design, operating parameters, and the set points at which automatic
actions are initiated. The proposed changes will not adversely
affect operation of plant equipment. These changes will not result
in a change to the set points at which protective actions are
initiated. Sufficient DC electrical system capacity is ensured to
support operation of mitigation equipment. The changes associated
with the new Battery Maintenance and Monitoring Program will ensure
that the station batteries are maintained in a highly reliable
state. The use of spare battery chargers will increase the
reliability of the DC electrical systems during periods of normal
battery charger inoperability. The equipment fed by the DC
electrical sources will continue to provide adequate power to safety
related loads in accordance with analysis assumptions. 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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
[[Page 21464]]
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph M.
Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: January 19, 2005.
Description of amendment request: The proposed amendments would
revise the Updated Final Safety Analysis Report to allow the use of
fire rated electrical cable for fire areas 2-013 and 2-042 in lieu of a
one hour rated electrical cable raceway fire barrier enclosure as
described by Title 10 of the Code of Federal Regulations (10 CFR) Part
50, Appendix R, Section III.G.2 for protection of safe shutdown
circuits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change 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) from performing their intended function to
mitigate the consequences of an initiating event within the assumed
acceptance limits. This is a revision to the FSAR to use [mineral
insulated] MI cable in fire areas 2-013 and 2-042. The MI cable has
been tested to applicable requirements and the implementation design
reflects the test results. Therefore, the probability of any
accident previously evaluated is not increased. Equipment required
to mitigate an accident remain capable of performing the assumed
function. Therefore, the consequences of any accident previously
evaluated are not increased.
Therefore, it is concluded that this change does not
significantly increase the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change will not alter the requirements or function
for systems required during accident conditions. No new or different
accidents result from implementing MI cable for fire areas 2-013 and
2-042. The MI cable has been tested to applicable requirements, and
the implementation design reflects the test results. The use of MI
cable is not a significant change in the methods governing normal
plant operation. The proposed change is consistent with the safety
analysis assumptions and current plant operating practice.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated is not created.
3. Does the proposed change involve a significant reduction in a
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 safety analysis acceptance criteria
are not affected by this change. The proposed change will not result
in plant operation in a configuration outside the design basis for
an unacceptable period of time without mitigating actions. The
proposed change does not affect systems that respond to safely
shutdown the plant and to maintain the plant in a safe shutdown
condition.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: John A. Nakoski.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: March 24, 2005.
Brief description of amendments: These proposed changes would
revise Technical Specification (TS) 3.3.1 entitled ``Reactor Trip
System Instrumentation'' (RTS) and TS 3.3.2 entitled ``Engineered
Safety Feature Actuation System Instrumentation'' (ESFAS) Required
Action Notes to reflect the wording in Standard Technical
Specifications (STS) for plants with bypass capability per TS Task
Force Traveler 418, Revision 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
[Westinghouse Topical Report] WCAP-14333 provided the technical
justification for relaxing various RTS and ESFAS Instrumentation
bypass test times, Completion Times, and Surveillance Frequencies
located in TS 3.3.1 and 3.3.2. As such, the proposed changes do not
represent a significant hazards consideration or present a reduction
in the margin of safety.
The protection system performance will remain within the bounds
of the previously performed accident analyses since no hardware
changes are proposed. The same Reactor Trip System (RTS)
Instrumentation and Engineered Safety Feature Actuation (ESFAS)
Instrumentation will continue to be used and remain unchanged. The
protection systems will continue to function in a manner consistent
with the plant design basis. These changes to the TS do not result
in a condition where the design, material, and construction
standards, which were applicable prior to these changes, are
altered.
The proposed changes will not modify any system interface. The
proposed changes will not affect the probability of any event
initiators. There will be no degradation in the performance of or an
increase in the number of challenges imposed on safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance. The proposed changes will not alter any
assumptions or change any mitigation actions in the radiological
consequence evaluations in the FSAR [final safety analysis report].
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configurations of the facility or change the manner in which the
plant is operated and maintained. The proposed changes do not alter
or prevent the ability of structures, systems, and components (SSCs)
from performing their intended function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes will not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
The proposed changes are consistent with safety analysis assumptions
and resultant consequences.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
There are no hardware changes nor is there any change in the
method by which any safety-related plant system performs its safety
function. The proposed changes will not affect the normal method of
plant operation. No performance requirements will be affected or
eliminated. The proposed changes will not result in physical
alteration to any plant system nor will there be any change in the
method by which any safety-related plant system performs its safety
function.
There will be no setpoint changes or changes to accident
analysis assumptions. No new accident scenarios, transient
precursors, failure mechanisms, or limiting single failures are
introduced as a result of these changes. There will be no adverse
effect or
[[Page 21465]]
challenges imposed on any safety-related system as a result of these
changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). There will be no effect on the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions. The radiological dose consequence acceptance criteria
listed in the Standard Review Plan will continue to be met.
Redundant RTS and ESFAS trains are maintained and diversity,
with regard to the signals that provide reactor trip and engineered
safety features actuation, is also maintained. All signals are
credited as primary or secondary and all operator actions credited
in the accident analyses will remain the same. The proposed changes
will not result in plant operation in a configuration outside the
design basis.
Therefore, the proposed changes do not involve a 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Allen G. Howe.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: March 1, 2005.
Description of amendment request: The proposed changes to the
Technical Specifications (TS) would revise the frequency for the Trip
Actuating Device Operational Test of the P-4 Interlock Function and add
Mode 4 to the Applicability for TS 3.3.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do changes involve a significant increase in the probability
or consequences of an accident previously evaluated?
The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the UFSAR [Updated Final Safety Analysis Report]. These interlocks
and the associated testing do not directly initiate an accident. The
consequences of accidents previously evaluated in the UFSAR are not
adversely affected by these proposed changes because the changes are
made to accurately reflect the design of the ESFAS [Engineered
Safety Features Actuation System] system. Therefore, the proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do changes create the possibility of a new or different kind
of accident from any accident previously evaluated?
The proposed changes do not create the possibility of a new or
different kind of accident from any accident already evaluated in
the UFSAR. No new accident scenarios, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes. The proposed changes do not challenge the performance or
integrity of any safety-related systems. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Do changes involve a significant reduction in the margin of
safety?
The proposed changes do not involve a significant reduction in a
margin of safety. The proposed changes are made to accurately
reflect the design of the ESFAS system. The nominal actuation set
points specified by the Technical Specifications and the safety
analysis limits assumed in the transient and accident analysis are
unchanged. Therefore, the proposed changes will not significantly
reduce the margin of safety as defined in the Technical
Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: March 8, 2005.
Description of amendment request: The proposed changes would revise
the auxiliary feedwater (AFW) operability requirements and add an AFW
allowed outage time and required actions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revision to the AFW pump and flowpath requirements,
as well as the revision of AFW surveillances, does not increase the
probability of accidents previously evaluated since the AFW System
is not required to operate until after the occurrence of the
previously evaluated accidents. The change does not impact any of
the initiators of the accidents. The proposed change does not
involve a significant increase in the consequences of an accident
previously evaluated because the AFW System will continue to perform
its intended safety function for these accidents. The operation of
the AFW System with the revised required action statements and added
surveillances continues to meet the applicable design criteria.
2. Create the possibility of a new or different type of accident
from any accident previously identified.
The safety function of the AFW System continues to be the same
and is met using the same equipment. The change does not involve any
plant modifications and does not revise the design of the plant or
the AFW System. Operation of the AFW System with the revised
required action statements and revised surveillances continues to
meet the applicable design criteria and is consistent with the Surry
accident analyses. Therefore, the proposed change does not introduce
any new failures that could create the possibility of a new or
different kind of accident from any accident previously identified.
3. Involve a significant reduction in a margin of safety.
The revised requirements for the AFW pumps and flowpaths, as
well as the revision of AFW surveillances, continue to assure that
the margins of safety assumed in the accidents and transients that
rely upon operation of the AFW System are maintained. The proposed
required action statements appropriately place the plant in a safe
condition for the circumstances being addressed. Therefore, this
proposed revision does not affect 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: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
[[Page 21466]]
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: March 17, 2005.
Description of amendment request: The proposed change would
incorporate a license condition that would permit irradiation of the
fuel assemblies to a lead rod average burnup of 62,000 MWD/MTU.
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 probability of occurrence or the consequences of an
accident previously evaluated is not significantly increased.
For most of the accidents analyzed in the UFSAR [Updated Final
Safety Analysis Report] (e.g., LOCA [loss-of-coolant accident],
Steam Line Break, etc.) the fuel design has no impact on the
likelihood of initiation of an accident. Fuel performance is
evaluated as a consequence of the accident. The only accident where
the fuel design may have an impact on the likelihood of a Chapter 14
accident is the Fuel Handling Accident discussed in Chapter 14.4.1
of the Surry UFSAR. The activity being evaluated is a slight
increase in the lead rod average burnup limit for the fuel
assemblies. No change in fuel design or fuel enrichment will be
required to increase the lead rod average burnup. The fuel rods at
the extended lead rod average burnup will continue to meet the
design limits with respect to fuel rod growth, clad fatigue, rod
internal pressure and corrosion. Thus, there will be no impact on
the capability to engage the fuel assemblies with the handling
tools. Therefore, it is concluded that the change will not result in
more than a minimal increase in the frequency of occurrence of any
accident previously evaluated in the UFSAR. The impact of extending
the lead rod average burnup to 62,000 MWD/MTU from 60,000 MWD/MTU on
the Core Kinetics Parameter, Core Thermal-Hydraulics/DNBR [Departure
from Nucleate Boiling Ratio], Specific Accident Considerations, and
Radiological Consequences was considered. Based on the evaluation of
these considerations, it is concluded that increasing the lead rod
average burnup limit to 62,000 MWD/MTU will not result in a
significant increase in the consequences of the accidents previously
evaluated in the Surry UFSAR.
2. The possibility for a new or different type of accident from
any accident previously evaluated is not created.
The fuel is the only component affected by the change in the
burnup limit. The change does not affect the thermal hydraulic
response to any transient or accident. The fuel rod design criteria
[will] continue to be met at the higher burnup limit. Thus, the
change does not create the possibility of an accident of a different
type.
3. The margin of safety as defined in the Bases to the Surry
Technical Specifications is not significantly reduced.
The operation of the Surry cores with a limited number of fuel
assemblies with some fuel rods irradiated to a lead rod average
burnup of 62,000 MWD/MTU will not change the performance
requirements of any system or component such that any design
criteria will be exceeded. The normal limits on core operation
defined in the Surry Technical Specifications will remain applicable
for the irradiation of the fuel to a lead rod average burnup of
62,000 MWD/MTU. Therefore, the margin of safety as defined in Bases
to the Surry Technical Specifications is not significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: July 6, 2004, as supplemented by
letters dated September 21, and December 23, 2004.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to allow a one-time change in the Appendix J, Type
A, Containment Integrated Leak Rate Test from the required 10 years to
15 years.
Date of issuance: April 6, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 285.
Facility Operating License No. DPR-65: The amendment revised the
TSs.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5237). The September 21 and December 23, 2004, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the
application beyond the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 6, 2005.
No significant hazards consideration comments received: No.
[[Page 21467]]
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: April 6, 2004, as supplemented
by letter dated August 5, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications to allow a diesel generator battery to remain
operable with no more than one cell less than 1.36 Volts DC on float
charge.
Date of issuance: March 29, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 221 and 216.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 14, 2004 (69
FR 55469). The supplement dated August 5, 2004 provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 29, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 28, 2004.
Brief description of amendments: The amendments eliminate the
technical specification requirements to submit monthly operating
reports and annual occupational radiation exposure reports.
Date of issuance: March 31, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 222 and 217.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 23, 2004 (69
FR 68182).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: June 3, 2003, as supplemented
by letters dated July 29 and December 7, 2004, and January 18, 2005.
Brief description of amendments: The amendments revise TS 3.6.14 to
allow a pressurizer enclosure hatch between the upper and lower
containment volumes to be open for up to 6 hours to facilitate
inspections of components such as the power operated relief valve block
valves.
Date of issuance: April 5, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 228/210.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43383). The supplemental letters dated July 29 and December 7, 2004,
and January 18, 2005, provided clarifying information that did not
change the initial proposed no significant hazards consideration
determinations.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 5, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: September 20, 2004.
Brief description of amendments: The amendments deleted the
Technical Specifications associated with hydrogen recombiners and
hydrogen monitors.
Date of issuance: April 4, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 227 and 209.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5239)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: September 28, 2004.
Brief description of amendments: The amendments eliminate the
technical specification requirements to submit monthly operating
reports and annual occupational radiation exposure reports.
Date of issuance: March 31, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 226 and 208.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 23, 2004 (69
FR 68182).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: October 16, 2003, as
supplemented by letters dated May 11, 2004, and January 10, 2005.
Brief description of amendments: The amendments revised the
Technical Specification (TS) 3.4.9 and the associated Bases to change
the minimum pressurizer heater capacity from 126 kW to 400 kW to
correct a non-conservative TS associated with a pressurizer design-
basis deficiency.
Date of Issuance: March 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 343, 345, & 344.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 20, 2004 (69 FR
2740).
The supplements dated May 11, 2004, and January 10, 2005, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register on January 20, 2004
(69 FR 2740).
[[Page 21468]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 28, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: September 20, 2004.
Brief description of amendments: The amendments delete the
Technical Specifications associated with hydrogen monitors.
Date of Issuance: April 4, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days after completion of the Spring 2005 refueling outage for
Unit 1.
Amendment Nos.: 344, 346 & 345.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5239).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 2005.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: August 5, 2004.
Brief description of amendment: This amendment revises Technical
Specification Section 5.5.12, ``Primary Containment Integrity,'' to
allow a one-time extension of its Appendix J, Type A, Containment
Integrated Leak Rate Test interval from the current 10-year interval to
a proposed 15-year interval.
Date of issuance: April 12, 2005.
Effective date: April 12, 2005, and shall be implemented within 30
days.
Amendment No.: 191.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53102).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: June 24, 2004.
Brief description of amendment: The amendment modifies Technical
Specification (TS) requirements to adopt the provisions of Industry/TS
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: April 6, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 226.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62474).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 6, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: December 30, 2004.
Brief description of amendment: The amendment changes the frequency
for Technical Specification surveillance requirement (SR) 3.1.4.2,
which verifies each tested control rod scram time is within limits with
reactor steam dome pressure >= 800 psig. Specifically, the SR frequency
increases from 120 days to 200 days of cumulative operation in MODE 1
(power operation).
Date of issuance: April 5, 2005.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 283.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5241).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 5, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: September 2, 2004.
Brief description of amendment: The amendment revised Technical
Specification (TS) 4.5.B.2.2 to change the surveillance requirement
frequency for air testing the drywell and suppression pool spray
headers and nozzles from ``once per 5 years'' to ``following
maintenance that could result in nozzle blockage.''
Date of issuance: April 12, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 214.
Facility Operating License No. DPR-35: The amendment revised the
TSs.
Date of initial notice in Federal Register: December 21, 2004 (69
FR 76490).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: April 13, 2004.
Brief description of amendments: The amendments eliminate the
requirements in Technical Specifications (TSs) associated with hydrogen
recombiners, and hydrogen and oxygen monitors.
Date of issuance: April 13, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 173 and 135.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the TSs.
Date of initial notice in Federal Register: June 8, 2004 (69 FR
32073).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 13, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: April 30, 2004.
Brief description of amendments: The amendments modify technical
specification (TS) requirements to adopt the provisions of Industry/TS
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: April 11, 2005.
Effective date: As of the date of issuance, to be implemented
within 180 days.
Amendment Nos.: 252 and 255.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the Technical Specifications.
[[Page 21469]]
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60681).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 11, 2005.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: August 31, 2004.
Brief description of amendment: The amendment revised Technical
Specification 3.4.1, ``Recirculation Loops Operating,'' associated with
single recirculation loop operation by incorporating limits for the
linear heat generation rate fuel thermal limit into the limiting
condition for operation.
Date of issuance: March 31, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 134.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 2005 (70 FR
401).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 2005.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: September 21, 2004.
Brief description of amendment: The amendment deletes the Technical
Specifications associated with hydrogen monitors.
Date of issuance: April 5, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 216.
Facility Operating License No. DPR-72: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5245).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 5, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: July 6, 2004, as supplemented
January 27, 2005.
Brief description of amendment: The amendment relocates the
calibration requirement of Table TS 4.1-1, Item 22, ``Accumulator Level
and Pressure,'' and the surveillance requirements of Table TS 4.1-1,
Item 25, ``Portable Radiation Survey Instruments,'' from the Technical
Specifications to licensee-controlled documents.
Date of issuance: April 6, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 182.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53112).
The supplement dated January 27, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the Nuclear
Regulatory Commission staff's original proposed no significant hazards
consideration. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 6, 2005.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: July 23, 2004, as supplemented
January 6, 2005.
Brief description of amendments: The amendments modified the
Technical Specification (TS) definition OPERABLE with respect to
requirements for availability of normal and emergency power.
Additionally, required actions for shutdown power TSs were modified.
Date of issuance: April 1, 2005.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment Nos.: 264 and 246.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9983).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 1, 2005.
No significant hazards consideration comments received: Comments
received were addressed in the Safety Evaluation dated April 1, 2005.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: July 26, 2004, as supplemented
on March 7, 2005.
Brief description of amendment: The amendment revised the Technical
Specifications by eliminating the requirements to provide the NRC
monthly operating reports and annual occupational radiation exposure
reports.
Date of issuance: April 13, 2005.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 89.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications and/or License.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60685). The supplemental letter provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 13, 2005.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: December 10, 2004.
Brief description of amendments: These amendments delete the
Technical Specifications associated with hydrogen monitors.
Date of issuance: March 29, 2005.
Effective date: March 29, 2005, to be implemented within 60 days of
issuance.
Amendment Nos.: Unit 2--194; Unit 3--185.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2896).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 29, 2005.
No significant hazards consideration comments received: No.
[[Page 21470]]
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Dates of application for amendments: February 26 and April 28,
2008, as supplemented by letters dated July 8 and October 20, 2004.
Brief description of amendments: The amendments revised Technical
Specification (TS) Section 5.6.6, Reactor Coolant System (RCS) Pressure
Temperature Limits Report (PTLR), to facilitate future licensee-
controlled changes to the PTLR. The changes include a revised PTLR that
provides new heatup and cooldown limits and Cold Overpressure
Protection System (COPS) set points, and to recalculate the minimum
size of the pressurizer power operated relief valve orifice of the RCS
vent. In addition, the changes relocate the COPS arming temperature to
the PTLR, and lower the COPS arming temperature from 350 [deg]F to 220
[deg]F. The licensee also included TS bases changes to support the
changes to the TSs.
Date of issuance: March 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 136 (Unit 1) and 115 (Unit 2).
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19575) and April 22, 2004 (69 FR 34707).
The supplements dated July 8 and October 20, 2004, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 28, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: October 14, 2004.
Brief description of amendments: The amendments eliminate the
technical specification requirements to submit monthly operating
reports and annual occupational radiation exposure reports.
Date of issuance: April 5, 2005.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 300 and 289.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5250).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 5, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: November 8, 2004.
Brief description of amendment: The amendment eliminates the
requirements in Technical Specifications to submit monthly operating
reports and annual occupational radiation exposure reports.
Date of issuance: March 21, 2005.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 57.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2902).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 21, 2005.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: September 8, 2004.
Brief description of amendment: These amendments delete the
Technical Specifications associated with hydrogen recombiners and
hydrogen monitors.
Date of issuance: March 22, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 238 and 219.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2902).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 22, 2005.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: June 23, 2004.
Brief Description of amendments: These amendments revise the
Technical Specifications Section 3.16, ``Emergency Power System,''
requirements for verifying the operability of the remaining emergency
diesel generator (EDG) when either unit's dedicated EDG or the shared
backup EDG is inoperable.
Date of issuance: April 5, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 241 and 240.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: August 19, 2004 (69 FR
51490).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 5, 2005.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: December 21, 2004.
Brief Description of amendments: These amendments revise the
Technical Specifications by eliminating the requirements to submit
monthly operating reports and occupational radiation exposure reports.
Date of issuance: March 22, 2005.
Effective date: March 22, 2005.
Amendment Nos.: 240 and 239.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2903).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 22, 2005.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 18th day of April 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 05-8166 Filed 4-25-05; 8:45 am]
BILLING CODE 7590-01-P