[Federal Register Volume 73, Number 88 (Tuesday, May 6, 2008)]
[Notices]
[Pages 25034-25050]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-9679]
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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 10 to April 23, 2008. The last
biweekly notice was published on April 22, 2008 (73 FR 21567).
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
[[Page 25035]]
within 30 days after the date of publication of this notice will be
considered in making any final determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice,
person(s) 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
via electronic submission through the NRC E-Filing system 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 person(s) 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 hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve documents over the internet
or in some cases to mail copies on electronic storage media.
Participants may not submit paper copies of their filings unless they
seek a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or
[[Page 25036]]
representative) already holds an NRC-issued digital ID certificate).
Each petitioner/requestor will need to download the Workplace Forms
ViewerTM to access the Electronic Information Exchange
(EIE), a component of the E-Filing system. The Workplace Forms
ViewerTM is free and is available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying
for a digital ID certificate is available on NRC's public Web site at
http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737.
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to submit documents in paper format. Such
filings must be submitted 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; or (2) courier, express mail, or expedited
delivery service to the Office of the Secretary, Sixteenth Floor, One
White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852,
Attention: Rulemaking and Adjudications Staff. Participants filing a
document in this manner are responsible for serving the document on all
other participants. Filing is considered complete by first-class mail
as of the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings. With respect to copyrighted
works, except for limited excerpts that serve the purpose of the
adjudicatory filings and would constitute a Fair Use application,
participants are requested not to include copyrighted materials in
their submission.
For further details with respect to this amendment 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].
Arizona Public Service Company (APS), et al., Docket Nos. STN 50-528,
STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of amendment request: January 17, 2008, as supplemented
February 29, 2008.
Description of amendment request: The proposed amendments would
modify the Technical Specifications (TS) to establish more effective
and appropriate action, surveillance, and administrative requirements
related to ensuring the habitability of the control room envelope (CRE)
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task
Force (TSTF) Standard Technical Specification change traveler TSTF-448,
Revision 3, ``Control Room Habitability.'' Specifically, the proposed
amendments would modify TS 3.7.11, ``Control Room Essential Filtration
System (CREFS),'' and add new TS 5.5.17, ``Control Room Envelope
Habitability Program,'' to TS Administrative Controls Section 5.5,
``Programs and Manuals.''
The NRC staff issued a ``Notice of Availability of Technical
Specification Improvement to Modify Requirements Regarding Control Room
Envelope Habitability Using the Consolidated Line Item Improvement
Process,'' associated with TSTF-448, Revision 3, in the Federal
Register on January 17, 2007 (72 FR 2022). The notice included a model
safety evaluation, a model no significant hazards consideration (NSHC)
determination, and a model license amendment request. In its
application dated January 17, 2008, the licensee affirmed the
applicability of the model NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee is presented below:
Criterion 1--The Proposed Change[s] [Do] Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change[s] [do] not adversely affect accident
initiators or precursors nor alter the design assumptions, conditions,
or configuration of the facility. The proposed change[s] [do] not alter
or prevent the ability of structures, systems, and components (SSCs) to
perform their intended function to mitigate the consequences of an
initiating event within the assumed
[[Page 25037]]
acceptance limits. The proposed change[s] [revise] the TS for the CRE
[essential filtration] system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE [essential
filtration] system is the CRE boundary. The CRE [essential filtration]
system is not an initiator or precursor to any accident previously
evaluated. Therefore, the probability of any accident previously
evaluated is not increased. Performing tests to verify the operability
of the CRE boundary and implementing a program to assess and maintain
CRE habitability ensure that the CRE [essential filtration] system is
capable of adequately mitigating radiological consequences to CRE
occupants during accident conditions, and that the CRE [essential
filtration] system will perform as assumed in the consequence analyses
of design basis accidents. Thus, the consequences of any accident
previously evaluated are not increased. Therefore, the proposed
change[s] [do] not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change[s] [Do] Not Create the Possibility of
a New or Different Kind of Accident From any Accident Previously
Evaluated
The proposed change[s] [do] not impact the accident analysis. The
proposed change[s] [do] not alter the required mitigation capability of
the CRE [essential filtration] system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new surveillance
or following the new program. The proposed change[s] [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 proposed change[s] [do] not alter
any safety analysis assumptions and is consistent with current plant
operating practice. Therefore, [the] change[s] [do] not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3--The Proposed Change[s] [Do] Not Involve a Significant
Reduction in the Margin of Safety
The proposed change[s] [do] not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change[s] [do] not affect safety
analysis acceptance criteria. The proposed change[s] will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The proposed
change[s] [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[s] [do] not involve a significant
reduction in a margin of safety. Based upon the reasoning presented
above and the previous discussion of the amendment request, the
requested change does not involve a no-significant-hazards
consideration.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on that review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the request for amendments involves NSHC.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: January 22, 2008.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) 3.8.3 requirements related to
Diesel Fuel Oil, Lube Oil, and Starting Air by replacing the specific
fuel oil and lube oil storage values with the corresponding number of
days supply. The specific volumes would be relocated to a licensee-
controlled document (i.e., the TS Bases). It would also expand the
``clear and bright'' test in TS 5.5.10 by allowing a water and sediment
test to be performed to establish the acceptability of new fuel oil
prior to addition to the storage tanks.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the Diesel Fuel Oil, Lube Oil, and
Starting Air Specification relocates the volume of diesel fuel oil
and lube oil required to support 7 day operation of the onsite
diesel generators, and the volume equivalent to a 6 day supply, to
licensee control. The specific volume of fuel oil equivalent to a 7
and 6 day supply is calculated using the NRC approved methodology
described in Regulatory Guide 1.137, Revision 1, ``Fuel Oil Systems
for Standby Diesel Generators'' and ANSI/ANS [American National
Standards Institute/American Nuclear Society] 59.51-1997 (formerly
ANSI N195-1976), ``Fuel Oil Systems for Safety-Related Emergency
Diesel Generators.'' The specific volume of lube oil equivalent to a
7 and 6 day supply is based on the Emergency Diesel Generator (EDG)
manufacturer's consumption values for the run time of the EDG.
Because the requirements to maintain a 7 day supply of diesel fuel
oil and lube oil are not changed and are consistent with the
assumptions in the accident analyses, and the actions taken when the
volume of fuel oil and lube oil are less than a 6 day supply have
not changed, neither the probability nor the consequences of any
accident previously evaluated will be affected. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to the Diesel Fuel Oil Testing Program adds
an option to use already approved testing methodology. Since the
methodology is already discussed in ASTM D975 [``Standard
Specification for Diesel Fuel Oils''] as an acceptable standard to
determine water and sediment content, neither the probability nor
the consequences of any accident previously evaluated will be
affected. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the Diesel Fuel Oil, Lube Oil and
Starting Air Specification and Diesel Fuel Oil Testing Program do
not involve physical alterations of the plant (i.e., no new or
different type of equipment will be installed) or changes in the
methods governing normal plant operation. The changes do not alter
assumptions made in the safety analysis but ensure that the diesel
generator operates as assumed in the accident analysis. The proposed
changes are consistent with the safety analysis assumptions.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to the Diesel Fuel Oil, Lube Oil, and
Starting Air Specification relocates the volume of diesel fuel oil
and lube oil required to support 7 day operation of the onsite
diesel generators, and the volume equivalent to a 6 day supply, to
[[Page 25038]]
licensee control. As the bases for the existing limits on diesel
fuel oil and lube oil are not changed and the methods used to
determine these limits have been previously approved, no change is
made to the accident analysis assumptions and no margin of safety is
reduced as part of this change. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
The proposed change to the Diesel Fuel Oil Testing Program
provides an option to use a quantitative method of testing for
sediment and water content as an alternative to a qualitative
method. This option uses an already accepted method for assessing
fuel oil quality. Based on this, there are no alterations to any
assumptions used in the accident analysis and this change does not
reduce any margin of safety. Therefore, the proposed change does not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York
Date of amendment request: February 7, 2008.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) Surveillance Requirement (SR)
3.1.3.2 frequency in TS 3.1.3, ``Control Rod OPERABILITY'' from ``7
days after the control rod is withdrawn and THERMAL POWER is greater
than the [Low Power Setpoint] LPSP of [Rod Worth Minimizer] RWM'' to
``31 days after the control rod is withdrawn and THERMAL POWER is
greater than the LPSP of the RWM'' and revise Example 1.4-3 in Section
1.4 ``Frequency'' to clarify the applicability of the 1.25 surveillance
test interval extension. The proposed amendment does not adopt the
clarification of Source Range Monitor (SRM) TS action for inserting
control rods. This clarification was previously adopted during the
JAFNPP conversion to Improved Standard Technical Specifications, TS
Section 3.3.1.2, required Action E.2, ``Source Range Monitoring [SRM]
Instrumentation.''
Date of publication of individual notice in Federal Register: April
2, 2008 (73 FR 18008).
Expiration date of individual notice: May 2, 2008.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: March 13, 2008.
Description of amendment request: The licensee proposes to change
the Surveillance Requirement (SR) 3.6.5.8 to require verification that
the reactor building spray nozzles are unobstructed following
maintenance that could result in nozzle blockage in lieu of the current
SR of performing the test every 10 years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Reactor Building Spray System is not an initiator of any
analyzed event. The proposed change does not have a detrimental
impact on the integrity of any plan structure, system, or component
that may initiate an analyzed event. The proposed change will not
alter the operation or otherwise increase the failure probability of
any plant equipment that can initiate an analyzed accident. This
change does not affect the plant design. There is no increase in the
likelihood of formation of significant corrosion products. Due to
their location at the top of the containment, introduction of
foreign material into the spray headers is unlikely. Foreign
materials exclusion controls during and following maintenance
provides assurance that the nozzles remain unobstructed.
Consequently, there is no significant increase in the probability of
an accident previously evaluated.
The Reactor Building Spray system is designed to address the
consequences of a Loss of Coolant Accident (LOCA) or a Main
Steamline Break (MSLB) inside the reactor building. The Reactor
Building Spray system is capable of performing its function
effectively with the single failure of any active component in the
system, any of its subsystems, or any of its support systems.
Therefore, the consequences of an accident previously evaluated
are not significantly affected by the proposed change.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not physically alter the plant (no new
or different type of equipment will be installed) or change the
methods governing normal plant operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The system piping and nozzles are made if material that is not
susceptible to corrosion. Obstruction from sources external to the
system is highly unlikely due to the location high in the reactor
building and not being readily accessible. Strict controls are
established to ensure the foreign material is not introduced into
the Reactor Building Spray system during maintenance or repairs.
Maintenance activities that could introduce significant foreign
material into the system require subsequent system cleanliness
verification which would prevent nozzle blockage. The spray header
nozzles are expected to remain unblocked and available in the event
that the safety function is required. The capacity of the system
would remain unaffected.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Entergy Nuclear Operations, P.O. Box 31995, Jackson,
Mississippi 39286-1995.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: March 13, 2008.
Description of amendment request: The proposed changes would
replace the current Technical Specification (TS) 3.4.12, ``RCS [Reactor
Coolant System] Specific Activity'' limit on reactor coolant system
(RCS) gross specific activity with a new limit on RCS noble gas
specific activity. The noble gas specific activity limit would be based
on a new dose equivalent Xe-133 (DEX) definition that would replace the
current E Bar average disintegration energy definition. In addition,
the current dose equivalent I-131 (DEI) definition would be revised to
allow the use of additional thyroid dose conversion factors (DCFs).
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
[[Page 25039]]
consequences of an accident previously evaluated?
Response: No.
Reactor coolant specific activity is not an initiator for any
accident previously evaluated. The Completion Time when primary
coolant gross activity is not within limit is not an initiator for
any accident previously evaluated. The current variable limit on
primary coolant iodine concentration is not an initiator to any
accident previously evaluated. As a result, the proposed change does
not significantly increase the probability of an accident. The
proposed change will limit primary coolant noble gases to
concentrations consistent with the accident analyses. The proposed
change to the Completion Time has no impact on the consequences of
any design basis accident since the consequences of an accident
during the extended Completion Time are the same as the consequences
of an accident during the current Completion Time. As a result, the
consequences of any accident previously evaluated are not
significantly increased.
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 in specific activity limits does not alter
any physical part of the plant nor does it affect any plant
operating parameter. The change does not create the potential for a
new or different kind of accident from any previously calculated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the limits on noble gas
radioactivity in the primary coolant. The proposed change is
consistent with the assumptions in the safety analyses and will
ensure the monitored values protect the initial assumptions in the
safety analyses.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Entergy Nuclear Operations, P.O. Box 31995, Jackson,
Mississippi 39286-1995.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: March 13, 2008.
Description of amendment request: The proposed changes would
replace the current TS 3.4.8, ``Reactor Coolant System Specific
Activity'' limit on reactor coolant system (RCS) gross specific
activity with a new limit on RCS noble gas specific activity. The noble
gas specific activity limit would be based on a new dose equivalent Xe-
133 (DEX) definition that would replace the current E Bar average
disintegration energy definition. In addition, the current dose
equivalent I-131 (DEI) definition would be revised to allow the use of
additional thyroid dose conversion factors (DCFs).
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.
Reactor coolant specific activity is not an initiator for any
accident previously evaluated. The Completion Time when primary
coolant gross activity is not within limit is not an initiator for
any accident previously evaluated. The current variable limit on
primary coolant iodine concentration is not an initiator to any
accident previously evaluated. As a result, the proposed change does
not significantly increase the probability of an accident. The
proposed change will limit primary coolant noble gases to
concentrations consistent with the accident analyses. The proposed
change to the Completion Time has no impact on the consequences of
any design basis accident since the consequences of an accident
during the extended Completion Time are the same as the consequences
of an accident during the current Completion Time. As a result, the
consequences of any accident previously evaluated are not
significantly increased.
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 in specific activity limits does not alter
any physical part of the plant nor does it affect any plant
operating parameter. The change does not create the potential for a
new or different kind of accident from any previously calculated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the limits on noble gas
radioactivity in the primary coolant. The proposed change is
consistent with the assumptions in the safety analyses and will
ensure the monitored values protect the initial assumptions in the
safety analyses.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Entergy Nuclear Operations, P. O. Box 31995, Jackson,
Mississippi 39286-1995.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: March 13, 2008.
Description of amendment request: The proposed change will relocate
Technical Specification (TS) 3.4.7, ``Reactor Coolant System
Chemistry,'' to the Technical Requirements Manual (TRM).
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 acts to relocate current Reactor Coolant
System (RCS) chemistry limits and monitoring requirements from the
TSs to the TRM. Monitoring and maintaining RCS chemistry minimizes
the potential for corrosion of RCS piping and components. Corrosion
effects are considered a long-term impact on RCS structural
integrity. Because RCS chemistry will continue to be monitored and
controlled, relocating the current TS requirements to the TRM will
not present an adverse impact to the RCS and, subsequently, will not
impact the probability or consequences of an accident previously
evaluated. Furthermore, once relocated to the TRM, changes to RCS
chemistry limits or monitoring requirements will be controlled in
accordance with 10 CFR 50.59.
Therefore, the proposed change does not involve a significant
increase in the
[[Page 25040]]
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not result in any plant modifications
or changes in the way the plant is operated. The proposed change
only acts to relocate current RCS chemistry limits and monitoring
requirements from the TSs to the TRM.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will maintain limits on RCS chemistry
parameters and will continue to provide associated monitoring
requirements. Once relocated to the TRM, changes to RCS chemistry
limits or monitoring requirements will be controlled in accordance
with 10 CFR 50.59. In addition, the RCS chemistry limits are not a
structure, system, or component which operating experience or
probabilistic risk assessment has shown to be significant to public
health and safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Entergy Nuclear Operations, P.O. Box 31995, Jackson,
Mississippi 39286-1995.
NRC Branch Chief: Thomas G. Hiltz.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: December 12, 2007.
Description of amendment request: The proposed changes are
administrative in nature and provide editorial changes to the technical
specifications (TSs). The proposed changes involve: (1) Correcting the
index; (2) removing cycle specific requirements or notes that have
since expired and are no longer applicable; (3) deleting references to
previously deleted requirements; (4) changing references to the
location of previously relocated information; and (5) other editorial
corrections. These proposed changes correct minor inconsistencies that
have been introduced over time as a result of previous changes to the
TSs or involve changes that are solely editorial in nature.
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 changes are administrative in nature and do not
impact the physical configuration or function of plant structures,
systems, or components (SSCs) or the manner in which SSCs are
operated, maintained, modified, tested, or inspected. The proposed
changes do not impact the initiators or assumptions of analyzed
events, nor do they impact mitigation of accidents or transient
events.
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 changes are administrative in nature and do not
alter plant configuration, require that new plant equipment be
installed, alter assumptions made about accidents previously
evaluated, or impact the function of plant SSCs or the manner in
which SSCs are operated, maintained, modified, tested, or inspected.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes are administrative in nature and do not
involve any physical changes to plant SSCs or the manner in which
SSCs are operated, maintained, modified, tested, or inspected. The
proposed changes do not involve a change to any safety limits,
limiting safety system settings, limiting conditions of operation,
or design parameters for any SSC. The proposed changes do not impact
any safety analysis assumptions and do not involve a change in
initial conditions, system response times, or other parameters
affecting an accident analysis. 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: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS),
Units 2 and 3, York and Lancaster Counties, Pennsylvania
Date of amendment request: July 13, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specifications to support application of
Alternative Source Term (AST) methodology at PBAPS Units 2 and 3. The
fission product release from the reactor core into containment is
referred to as the ``source term,'' and is characterized by the
composition and magnitude of the radioactive material, the chemical and
physical properties of the material, and the timing of the release from
the reactor core as discussed in Technical Information Document (TID)
14844, ``Calculation of Distance Factors for Power and Test Reactor
Sites.'' Since the publication of TID 14844, advances have been made in
understanding the composition and magnitude, chemical form, and timing
of fission product releases from severe nuclear power plant accidents.
In light of these insights, NUREG-1465, ``Accident Source Terms for
Light-Water Nuclear Power Plants,'' was published in 1995 with revised
ASTs for use in the licensing of future light-water reactors.
The Nuclear Regulatory Commission (NRC), in Title 10 of the Code of
Federal Regulations, Section 50.67 (10 CFR 50.67), ``Accident source
term,'' subsequently allowed the use of the ASTs described in NUREG-
1465 at operating plants. This request to apply the AST methodology is
made in accordance with 10 CFR 50.67, with the exception that TID 14844
will continue to be used as the radiation dose basis for equipment
qualification at PBAPS Units 2 and 3. Application of the AST
methodology at PBAPS Units 2 and 3 requires that radiation dose limits
specified in 10 CFR 50.67 are adhered to for the exclusion area
boundary, the low population zone outer boundary, and the facility
control room.
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
[[Page 25041]]
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The implementation of alternative source term (AST) assumptions
has been evaluated in revisions to the analyses of the following
limiting design basis accidents (DBAs) at Peach Bottom Atomic Power
Station (PBAPS):
Loss-of-Coolant Accident,
Fuel Handling Accident,
Control Rod Drop Accident, and
Main Steam Line Break Accident.
Based upon the results of these analyses, it has been
demonstrated that, with the requested changes, the dose consequences
of these limiting events are within the regulatory guidance provided
by the NRC for use with the AST. This guidance is presented in 10
CFR 50.67 and associated Regulatory Guide 1.183, and Standard Review
Plan Section 15.0.1. The Alternative Source Term is an input to
calculations used to evaluate the consequences of an accident, and
does not by itself affect the plant response, or the actual pathway
of the radiation released from the fuel. It does, however, better
represent the physical characteristics of the release, so that
appropriate mitigation techniques may be applied. Therefore, the
consequences of an accident previously evaluated are not
significantly increased.
The equipment affected by the proposed changes is mitigative in
nature, and relied upon after an accident has been initiated.
Application of the Alternative Source Term (AST) does not involve
any physical changes to the plant design. While the operation of
various systems do change as a result of these proposed changes,
these systems are not accident initiators. Application of the AST is
not an initiator of a design basis accident. The proposed changes to
the Technical Specifications (TS), while they revise certain
performance requirements, do not involve any physical modifications
to the plant. As a result, the proposed changes do not affect any of
the parameters or conditions that could contribute to the initiation
of any accidents. As such, removal of operability requirements
during the specified conditions will not significantly increase the
probability of occurrence for an accident previously analyzed. Since
design basis accident initiators are not being altered by adoption
of the Alternative Source Term analyses, the probability of an
accident previously evaluated is not affected.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment does not involve a physical alteration of
the plant (no new or different type of equipment will be installed
and there are no physical modifications to existing equipment
associated with the proposed changes). Similarly, it does not
physically change any structures, systems or components involved in
the mitigation of any accidents; thus, no new initiators or
precursors of a new or different kind of accident are created. New
equipment or personnel failure modes that might initiate a new type
of accident are not created as a result of the proposed amendment.
As such, the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Safety margins and analytical conservatisms have been evaluated
and have been found acceptable. The analyzed events have been
carefully selected and margin has been retained to ensure that the
analyses adequately bound postulated event scenarios. The dose
consequences due to design basis accidents comply with the
requirements of 10 CFR 50.67 and the guidance of Regulatory Guide
1.183. The proposed amendment is associated with the implementation
of a new licensing basis for PBAPS Design Basis Accidents (DBAs).
Approval of the change from the original source term to a new source
term taken from Regulatory Guide 1.183 is being requested. The
results of the accident analyses, revised in support of the proposed
license amendment, are subject to revised acceptance criteria. The
analyses have been performed using conservative methodologies, as
specified in Regulatory Guide 1.183. Safety margins have been
evaluated and analytical conservatism has been utilized to ensure
that the analyses adequately bound the postulated limiting event
scenario. The dose consequences of these DBAs remain within the
acceptance criteria presented in 10 CFR 50.67, ``Accident Source
Term'', and Regulatory Guide 1.183.
The proposed changes continue to ensure that the doses at the
exclusion area boundary (EAB) and low population zone boundary
(LPZ), as well as the Control Room, are within corresponding
regulatory limits.
Therefore, operation of PBAPS in accordance with the proposed
changes will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. J. Bradley Fewell, Associate General
Counsel, Exelon Generation Company LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
FPL Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: March 31, 2008.
Description of amendment request: FPL Energy Point Beach, LLC,
requests adoption of an approved change to the Standard Technical
Specifications (STS) for pressurized-water reactor (PWR) plants (NUREG-
1430, NUREG-1431, & NUREG-1432) and plant-specific technical
specifications (TS), to replace the current limits on primary coolant
gross specific activity with limits on primary coolant noble gas
activity. The noble gas activity would be based on dose equivalent
Xenon-133 and would take into account only the noble gas activity in
the primary coolant. In addition, the current dose equivalent I-131
definition would be revised to allow the use of additional thyroid dose
conversion factors. The changes are consistent with Nuclear Regulatory
Commission (NRC)-approved Industry/Technical Specification Task Force
(TSTF) Standard Technical Specification Change Traveler, TSTF-490,
Revision 0.
Basis for proposed no-significant-hazards-consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated Reactor coolant specific activity is not an initiator for any
accident previously evaluated. The Completion Time when primary coolant
gross activity is not within limit is not an initiator for any accident
previously evaluated. The current variable limit on primary coolant
iodine concentration is not an initiator to any accident previously
evaluated. As a result, the proposed change does not significantly
increase the probability of an accident. The proposed change will limit
primary coolant noble gases to concentrations consistent with the
accident analyses. The proposed change to the Completion Time has no
impact on the consequences of any design basis accident since the
consequences of an accident during the extended Completion Time are the
same as the consequences of an accident during the Completion Time. As
a result, the consequences of any accident previously evaluated are not
significantly increased.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident from any Accident Previously
Evaluated.
The proposed change in specific activity limits does not alter any
physical part of the plant nor does it affect any plant operating
parameter.
[[Page 25042]]
The change does not create the potential for a new or different kind of
accident from any previously calculated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The proposed change revises the limits on noble gas radioactivity
in the primary coolant. The proposed change is consistent with the
assumptions in the safety analyses and will ensure the monitored values
protect the initial assumptions in the safety analyses. Based upon the
reasoning presented above, the requested change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the 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: Antonio Fernandez, Esquire, Senior Attorney,
FPL Energy Point Beach, LLC, P. O. Box 14000, Juno Beach, FL 33408-
0420.
NRC Branch Chief: Lois M. James.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: March 31, 2008.
Description of amendment request: The licensee proposed to increase
the current maximum power level authorized by Section 2.C(1) of the
renewed facility operating license from 1,775 megawatts thermal (Mwt)
to 1,870 Mwt, an approximately five percent increase from the current
licensed thermal power. The current maximum power level of 1,775 Mwt
was approved in 1998, an increase of 6.3 percent from the original
licensed thermal power of 1670 Mwt. Thus, when approved, the licensee's
proposed amendment would take the maximum power level to about 12
percent above the original license thermal power. The licensee's
application addresses in details each of the following major technical
areas: Extended power uprate, containment analysis methods change,
increase in credit for containment overpressure for low head emergency
core cooling system (ECCS) pumps, and reactor internal pressure
differentials (RIPDs) for the steam dryer.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Part 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration (NSHC).
The licensee's NSHC analysis, addressing each technical area listed
above, is reproduced below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Extended Power Uprate
Response: No.
The probability (frequency of occurrence) of [d]esign [b]asis
[a]ccidents occurring is not affected by the increased power level,
because Monticello Nuclear Generating Plant (MNGP) continues to
comply with the regulatory and design basis criteria established for
plant equipment. A probabilistic risk assessment demonstrates that
the calculated core damage frequencies do not significantly change
due to [e]xtended [p]ower [u]prate (EPU). Scram setpoints (equipment
settings that initiate automatic plant shutdowns) are established
such that there is no significant increase in scram frequency due to
EPU. No new challenges to safety-related equipment result from EPU.
The changes in consequences of postulated accidents, which would
occur from 102 percent of the EPU [rated thermal power] RTP compared
to those previously evaluated, are acceptable. The results of EPU
accident evaluations do not exceed the NRC[-] approved acceptance
limits. The spectrum of postulated accidents and transients has been
investigated, and are shown to meet the plant's currently licensed
regulatory criteria. In the area of fuel and core design, for
example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and
other applicable Specified Acceptable Fuel Design Limits (SAFDL) are
still met. Continued compliance with the SLMCPR and other SAFDLs
will be confirmed on a cycle[-]specific basis consistent with the
criteria accepted by the NRC.
Challenges to the [r]eactor [c]oolant [p]ressure [b]oundary were
evaluated at EPU conditions (pressure, temperature, flow, and
radiation) and were found to meet their acceptance criteria for
allowable stresses and overpressure margin. Challenges to the
containment have been evaluated, and the containment and its
associated cooling systems continue to meet the current licensing
basis. The increase in the calculated post[-] LOCA suppression pool
temperature above the currently assumed peak temperature was
evaluated and determined to be acceptable. Radiological release
events (accidents) have been evaluated, and have been shown to meet
the guidelines of 10 CFR 50.67.
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks, variable RHR [residual heat
removal] heat exchanger capability K-value, and mechanistic heat and
mass transfer from the suppression pool surface to the wetwell
airspace after 30 seconds for the long[-]term design[-] basis [-
]accident loss of coolant accident (DBA-LOCA) containment analysis
are not relevant to accident initiation, but rather, pertain to the
method used to accurately evaluate postulated accidents. The use of
these elements does not, in any way, alter existing fission product
boundaries, and provides a conservative prediction of the
containment response to DBA-LOCAs. Therefore, the containment
analysis method change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Increase in Credit for Containment Overpressure for Low Head
Emergency Core Cooling System (ECCS) Pumps
Response: No.
These changes update parameters used in the MNGP safety analyses
and expand the range and scope of the analyses. This will result in
a more realistic analysis of available containment overpressure
under design [-]basis accident conditions. The updated analyses
affect only the evaluation of previously reviewed accidents. No
plant structure, system, or component (SSC) is physically affected
by the updated and expanded analyses. No method of operation of any
plant SSC is affected. Therefore, there is no significant increase
in the probability or consequence of a previously evaluated
accident.
Reactor Internal Pressure Differentials (RIPDs) for the Steam Dryer
Response: No.
The revised steam dryer RIPDs are used in evaluating loads in
reactor vessel internals for various conditions (i.e., during
normal, upset and faulted conditions). The values more accurately
represent the actual plant configuration. No plant structure,
system, or component (SSC) is physically affected by the updated and
expanded analyses. No method of operation of any plant SSC is
affected. Therefore, there is no significant increase in the
probability or consequence of a previously evaluated accident.
The analyses supporting the above evaluations were performed at
the EPU power level of 2,004 Mwt, which bounds this license
amendment request to operate at 1,870 Mwt. 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?
Extended Power Uprate
Response: No.
Equipment that could be affected by EPU has been evaluated. No
new operating mode, safety-related equipment lineup, accident
scenario, or equipment failure mode was identified. The full
spectrum of accident considerations has been evaluated and no new or
different kind of accident has been identified. EPU uses developed
technology and applies it within capabilities of existing or
modified plant safety[-]related equipment in accordance with the
regulatory criteria (including NRC[-]approved codes, standards and
methods). No new accidents or event precursors have been identified.
The MNGP TS require revision to implement EPU. The revisions
have been assessed and it was determined that the proposed change
will not introduce a different accident than that previously
evaluated. Therefore, the proposed changes
[[Page 25043]]
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks, variable RHR heat exchanger
capability K-value, and mechanistic heat and transfer from the
suppression pool surface to the wetwell airspace after 30 seconds
for the long term DBA-LOCA containment analysis are not relevant to
accident initiation, but pertain to the method used to evaluate
currently postulated accidents. The use of these analytical tools
does not involve any physical changes to plant structures or
systems, and does not create a new initiating event for the spectrum
of events currently postulated. Further, they do not result in the
need to postulate any new accident scenarios. Therefore, the
containment analysis method change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Increase in Credit for Containment Overpressure for Low Head ECCS
Pumps
Response: No.
The proposed change involves the updating and expansion in scope
of the existing design bases analysis with respect to the available
containment overpressure. No new failure mode or mechanisms have
been created for any plant SSC important to safety nor has any new
limiting single failure been identified as a result of the proposed
analytical changes. Therefore, the change to containment
overpressure credited for low pressure ECCS pumps does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Reactor Internal Pressure Differentials for the Steam Dryer
Response: No.
The revised steam dryer RIPDs are used in evaluating loads in
reactor vessel internals for various conditions (i.e., during
normal, upset and faulted conditions). The steam dryer RIPDs are not
relevant to accident initiation, but only pertain to the method used
to evaluate reactor vessel internals loads. The revised steam dryer
RIPD values more accurately represent the actual plant
configuration. Therefore, the change to steam dryer RIPDs does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
The analyses supporting the above evaluations were performed at
the EPU power level of 2,004 Mwt, which bounds this license
amendment request to operate at 1,870 Mwt. 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?
Extended Power Uprate
Response: No.
The EPU affects only design and operational margins. Challenges
to the fuel, reactor coolant pressure boundary, and containment were
evaluated for EPU conditions. Fuel integrity is maintained by
meeting existing design and regulatory limits. The calculated loads
on affected structures, systems and components, including the
reactor coolant pressure boundary, will remain within their design
allowables for design[-]basis event categories. No NRC acceptance
criterion is exceeded. Because the MNGP configuration and responses
to transients and postulated accidents do not result in exceeding
the presently approved NRC acceptance limits, the proposed changes
do not involve a significant reduction in a margin of safety.
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks, variable RHR heat exchanger
capability K-value, and mechanistic heat and mass transfer from the
suppression pool surface to the wetwell airspace after 30 seconds
for the long[-]term DBA-LOCA containment analysis are realistic
phenomena and provide a conservative prediction of the plant
response to DBA-LOCAs. The increase in pressure and temperature are
relatively small and are within design limits. Therefore, the
containment analysis methods change does not involve a significant
reduction in the margin of safety.
Increase in Credit for Containment Overpressure for Low Head ECCS
Pumps
Response: No.
The proposed changes revise containment response analytical
methods and scope for containment pressure to assist in ECCS pump
net positive suction head (NPSH). The changes are still based on
conservative but more realistic analysis of available containment
overpressure determined using analysis methods that minimize
containment pressure and maximize suppression pool temperature.
These changes do not constitute a significant reduction in the
margin of safety.
Reactor Internal Pressure Differentials for the Steam Dryer
Response: No.
The revised steam dryer RIPDs are used in evaluating loads in
reactor vessel internals for various conditions (i.e., during
normal, upset and faulted conditions). The revised steam dryer RIPD
values more accurately represent the actual plant configuration. The
changes are still conservative but more accurately represent the
MNGP configuration. These changes do not constitute a significant
reduction in the margin of safety.
The analyses supporting the above evaluations were performed at
the EPU power level of 2,004 Mwt, which bounds this license
amendment request to operate at 1,870 Mwt. 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
the NRC staff's own analysis above, it appears that the three standards
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the proposed amendment involves no significant hazards
consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: April 3, 2008.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) requirements related to control
room envelope (CRE) habitability in TS Section 3.7.4, ``Control Room
Emergency Filtration (CREF) System,'' and Section 5.5, ``Programs and
Manuals.'' The proposed changes are consistent with Technical
Specification Task Force (TSTF) Standard Technical Specifications (STS)
change TSTF-448, Revision 3.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Part 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration (NSHC) by
referencing the NRC staff's model NSHC analysis published on January
17, 2007 (72 FR 2022). The NRC staff's model NSHC analysis is
reproduced below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of
[[Page 25044]]
design basis accidents. Thus, the consequences of any accident
previously evaluated are not increased. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident From any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's referenced analysis, and
has found that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the proposed
amendment involves no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: March 28, 2008.
Description of amendment request: The amendments would revise PPL
Susquehanna, LLC, Units 1 and 2 (PPL) Technical Specifications (TSs)
3.8.4, ``DC Sources--Operating,'' to establish two new Conditions, A
and B the associated Required Actions with their completion times, and
also, make some editorial and administrative 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?
No. The proposed changes revise the Technical Specifications
(TS) for the DC Electrical Power Systems and propose new Actions
with increased completion times for an inoperable battery charger.
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 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 only involve revising the TS for the DC
electrical power systems. 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 setpoints 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 setpoints at which protective actions are
initiated. Sufficient DC electrical system capacity is ensured to
support operation of mitigation equipment. The equipment fed by the
DC electrical sources will continue to provide adequate power to
safety related loads in accordance with the safety analysis.
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 Branch Chief: Mark G. Kowal.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: March 28, 2008.
Description of amendment request: The amendments would revise PPL
Susquehanna, LLC, Units 1 and 2 (PPL) Technical Specifications (TSs) TS
3.6.4.1 ``Secondary Containment,'' and TS 3.6.4.3 ``Standby Gas
Treatment System,'' as follows:
(1) To add a new Required Action option for TS 3.6.4.1 Condition A,
to allow additional time to restore secondary containment to OPERABLE
when the inoperability is not caused by a loss of secondary containment
integrity,
(2) To add a new Actions note TS 3.6.4.1, to allow opening of
secondary containment heating ventilation and air conditioning duct
access doors and opening of a secondary containment equipment ingress/
egress door (102 door) under administrative controls provided no
movement of irradiated fuel assemblies in the secondary containment,
CORE ALTERATIONS, or operations with a potential for draining the
reactor vessel (OPDRVs) are in progress,
(3) To modify the existing note to Surveillance Requirement (SR)
3.6.4.1.3 and add a second note to this same SR, to expand upon the
existing SR exception note by adding other types of door access
openings that occur for entry and exit of people or equipment, and
(4) The administrative change to remove a one-time allowance in TS
3.6.4.1 and TS 3.6.4.3 ``Standby Gas Treatment System [SGTS],'' that
extended the allowable Completion Time for Secondary Containment
[[Page 25045]]
inoperable and two SGTS subsystems inoperable in MODE 1, 2, or 3. This
allowance was previously incorporated into both Unit 1 and Unit 2 TSs
to facilitate Reactor Recirculating Fan Damper Motor work.
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.
These changes do not involve any physical change to structures,
systems, or components (SSCs) and do not alter the method of
operation of any SSCs. The current assumptions in the safety
analysis regarding accident initiators and mitigation of accidents
are unaffected by these changes. No SSC failure modes or mechanisms
are being introduced, and the likelihood of previously analyzed
failures remains unchanged.
Operation in accordance with the proposed new Required Action
option for TS 3.6.4.1 Condition A and the Notes that are being
modified and added in both the Unit 1 and Unit 2 Technical
Specifications ensures that the secondary containment remains
capable of performing its function. The Required Action change,
which will permit up to 72 hours to restore secondary containment
vacuum, only provides this additional time when it can be shown that
the vacuum loss has not been caused through compromise of the
secondary containment boundary.
The proposed Note modifications and additions addressing
secondary containment access door and duct access door openings will
provide relief from TS requirements that must currently be
implemented in response to various routine plant activities. These
activities can be managed through administrative controls that will
ensure doors can be closed quickly (within 30 minutes) to re-
establish secondary containment before the early in-vessel release
phase begins (Regulatory Guide 1.183).
These changes do not, therefore, result in an 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 changes do not involve a physical alteration of any
plant equipment. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints, at which protective or mitigative actions are
initiated, affected by this change. This change does not alter the
manner in which equipment operation is initiated, nor will the
function demands on credited equipment be changed. No alterations in
the procedures that ensure the plant remains within analyzed limits
are being proposed, and no changes are being made to the procedures
relied upon to respond to an off-normal event as described in the
FSAR [final safety analysis report]. As such, no new failure modes
are being introduced. The change does not alter assumptions made in
the safety analysis and licensing basis.
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 margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes are acceptable because the
Completion Time for the new Required Action to verify secondary
containment boundary integrity within 4 hours has been established
to be consistent with the current completion time of Condition A. A
failure or inability to complete this verification will result in
the implementation of LCO [limiting condition for operation] 3.6.4.1
requirements in the same timeframe that currently exists. Upon
successful completion of this verification, however, the proposed
change will provide 72 hours to restore secondary containment to an
operable status through vacuum restoration. When in this condition,
the secondary containment and SGTS are capable of performing their
design basis function.
The Note modifications and additions to TS 3.6.4.1 are also
acceptable because the revised Notes provide allowances and
exemptions to Technical Specification entry for routine plant
activities that can be administratively controlled and quickly
restored.
The plant response to analyzed events is not affected by these
changes and will, continue to provide the margin of safety assumed
by the safety analysis.
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Mark G. Kowal.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment request: November 30, 2007.
Description of amendment request: The proposed Technical
Specification changes will provide operational flexibility supported by
direct current (DC) electrical subsystem design upgrades that are in
progress. These upgrades will provide increased capacity batteries,
additional battery chargers, and the means to cross-connect DC
subsystems while meeting all design battery loading requirements. With
these modifications in place, it will be feasible to perform routine
surveillances as well as battery replacements online.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed changes to Technical Specifications (TS) 3.8.4 and
3.8.6 would allow extension of the Completion Time (CT) for
inoperable Direct Current (DC) distribution subsystems to manually
cross-connect DC distribution buses of the same safety train of the
operating unit for 21 days (30 days for upgrade to 1800 amp-hour
rated batteries). Currently the CT only allows for 2 hours to
ascertain the source of the problem before a controlled shutdown is
initiated. Loss of a DC subsystem is not an initiator of an event.
However, complete loss of a Train A (subsystems A and C) or Train B
(subsystems B and D) DC system would initiate a plant transient/
plant trip.
Operation of a DC Train in cross-connected configuration does
not affect the quality of DC control and motive power to any system.
Therefore, allowing the cross-connect of DC distribution systems
does not significantly increase the probability of an accident
previously evaluated in Chapter 15 of the Updated Final Safety
Analysis Report (UFSAR).
The above conclusion is supported by Probabilistic Risk
Assessment (PRA) evaluation which encompasses all accidents,
including UFSAR Chapter 15.
New TS Surveillance Requirement (SR) 3.8.4.4 is added to allow
the application of the modified performance discharge testing on
batteries rated at 1800 amp-hour using a frequency of 30 months. The
application of the modified performance test is the preferred choice
at SONGS for Class 1 E 1800 amp-hour rated batteries. Therefore,
only the modified performance discharge test will be used which uses
the combined duty cycle of the cross-connected subsystems AC or B-D.
Battery life expectancy is optimized by using a 30-month modified
performance test (service and performance test combined). The more
rigorous modified performance discharge test will be applied in
intervals of 30 months over the entire battery life. Using the same
test method and test frequency throughout the battery life ensures
that best
[[Page 25046]]
trending results are achieved. The test frequency of 30 months will
better correspond with scheduling of the more rigorous 60-month
interval battery performance of modified performance discharge
tests. Based on operating experience, the interval of 30 months is
not expected to affect SONGS' capability to detect battery health
and capacity.
The relocation of preventive maintenance surveillances and
certain operating limits and actions to the Licensee Controlled
Specifications and new Battery Monitoring and Maintenance Program
will not challenge the ability of the DC electrical power system to
perform its design function. Appropriate monitoring and maintenance
consistent with industry standards will continue to be performed. In
addition, the DC electrical power system is 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
system. Enhancements from TSTF-360, Rev. 1 and IEEE 450-2002 have
been incorporated into TSs 3.8.4, 3.8.5, and 3.8.6. These changes do
not impact the probability or consequences of an accident previously
evaluated.
Further, changes are made of an editorial nature or provide
clarification regarding electrical `Trains' and `Subsystems' by
using a more conventional terminology. TSs affected by editorial
changes include 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, 3.8.9, and
3.8.10. The changes being proposed in the TS do not affect
assumptions contained in other safety analyses or the physical
design of the plant, nor do they affect other Technical
Specifications that preserve safety analysis assumptions.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously analyzed.
2. Will operation of the facility in accordance with this
proposed change create the possibility of new or different kind of
accident from any accident previously evaluated?
Response: No.
The proposed changes involve restructuring the TS for the DC
electrical power system. The DC electrical power system, including
associated battery chargers, is not an initiator to any accident
sequence analyzed in the UFSAR. Rather, the DC electrical power
system is used to supply equipment used to mitigate an accident.
The proposed change modifies TSs and surveillances for batteries
and chargers to meet the improvements of TSTF-360, Rev. 1 and IEEE
450-2002 whose intent is to maintain the same equipment capability
as previously assumed in Southern California Edison's (SCE's)
commitment to IEEE 450-1980.
The proposed change will allow the cross-tie of DC subsystems
and allow extension of the CT for an inoperable subsystem to 21 days
(30 days for upgrade to 1800 amp-hour rated batteries). Failure of
the cross-tied DC buses and/or associated battery(ies) is bounded by
existing evaluations for the failure of an entire electrical train.
Swing battery chargers are added to increase the overall DC
system reliability. Administrative and mechanical controls are in
place to ensure the design and operation of the DC systems continue
to meet the UFSAR design basis.
Therefore, operation of the facility in accordance with this
proposed change will not create the possibility of new or different
kind of accident from any accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints 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 setpoints at which protective actions are initiated.
Sufficient DC capacity to support operation of mitigation equipment
is ensured. The changes associated with the new battery maintenance
and monitoring program will ensure that the station batteries are
maintained in a highly reliable manner. The equipment fed by the DC
electrical sources will continue to provide adequate power to safety
related loads in accordance with analysis assumptions.
Improvements in accordance with IEEE 450-2002 and TSTF-360, Rev.
1 maintain the same level of equipment performance stated in the
UFSAR and the current Technical Specifications.
The addition of swing battery chargers increases the overall DC
system reliability. Administrative and mechanical controls will be
in place to ensure that the design and operation of the DC systems
continue to meet the UFSAR design basis.
The addition of the DC cross-tie capability proposed for TS
3.8.4 has been evaluated, as described previously, using PRA and
determined to be of acceptable risk as long as the duration while
cross-tied is limited to 30 days. A new Condition has been included
as part of this proposed change to ensure that plant operation, with
DC buses cross-tied, will not exceed 21 days (30 days for upgrade to
1800 amp-hour rated batteries).
Revising the LCO statement to reflect the SONGS-specific design
terminology and renaming existing conditions to make the Condition
more consistent with the Standard Technical Specifications (STS) is
considered administrative.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: August 29, 2006, as supplemented
November 6, November 27, 2006, January 30, June 22, July 16, August 13,
October 18, December 11, 2007, January 24, February 4, February 25 (two
letters, nos. 1389 and 0175), February 27, and March 13, 2008.
Description of amendment request: The proposed amendments would
revise the licensing and design basis, including the Technical
Specifications, with a full scope implementation of an alternative
source term (AST). The licensee states that the AST analyses include
determination of the onsite radiological doses, specifically the main
control room, technical support center and off-site radiological doses
resulting from the loss-of-coolant, main steam line break, control rod
drop, and fuel-handling design-basis accident (DBA) analyses. The
licensee states that the analyses demonstrate that, using AST
methodologies, the post-accident onsite and offsite doses remain within
regulatory acceptance 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. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Adoption of the AST and those plant systems affected by
implementing AST do not initiate DBAs. The AST does not affect the
design or manner in which the facility is operated; rather, once the
occurrence of an accident has been postulated, the new accident
source term is an input to analyses that evaluate the radiological
consequences. The implementation of the AST and changed Technical
Specifications have been incorporated in the analyses for the
limiting DBAs at HNP. The structures, systems, and components
affected by the proposed change are mitigative in nature and relied
upon after an accident has been initiated. Based on the revised
analyses, the proposed changes to the Technical Specifications
(including revised leakage limits) impose certain performance
criteria which do not increase accident initiation probability. The
proposed changes do not involve a revision to the parameters
[[Page 25047]]
or conditions that could contribute to the initiation of a DBA
discussed in Chapter 15 of the Unit 2 Final Safety Analysis Report.
Therefore, the proposed change does not result in an increase in the
probability of an accident previously identified. Plant specific AST
radiological analyses have been performed and, based on the results
of these analyses, it has been demonstrated that the dose
consequences of the limiting events considered in the analyses are
within the regulatory guidance provided by the Nuclear Regulatory
Commission for use with the AST. This guidance is presented in
[Title 10 of the Code of Federal Regulations, Section 50.67] (10 CFR
50.67), [Accident Source Term] Regulatory Guide 1.183, [Alternative
Radiological Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Reactors (ML003716792)] and Standard Review Plan,
Section 15.0.1. Therefore, the proposed change does not result in a
significant increase in the consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Implementation of AST and associated changes does not alter or
involve any design basis accident initiators. These changes do not
affect the design function or mode of operations of systems,
structures, or components in the facility prior to a postulated
accident. Since systems, structures, and components are operated
essentially no differently after the AST implementation, no new
failure modes are created by this proposed change. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant decrease in
the margin of safety?
The changes proposed are associated with a revision to the
licensing basis for HNP. Approval of the licensing basis change from
the original source term to the AST is requested by this application
for a license amendment. The results of the accident analyses
revised in support of the proposed change are subject to the
acceptance criteria in 10 CFR 50.67. The analyzed events have been
carefully selected, and the analyses supporting these changes have
been performed using approved methodologies and conservative inputs
to ensure that analyzed events are bounding and safety margin has
been retained. The dose consequences of these limiting events are
within the acceptance criteria presented in 10 CFR 50.67, Regulatory
Guide 1.183, and Standard Review Plan 15.0.1. Therefore, because the
proposed changes continue to result in dose consequences within the
applicable regulatory limits, the changes are considered to not
result in 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Branch Chief: Melanie C. Wong.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: March 27, 2008.
Description of amendment request: The proposed amendment would
revise the allowable value for Function 3, ``Containment Purge Exhaust
Radiation Monitors,'' in Technical Specifications (TSs) Table 3.3.6-1,
``Containment Vent Isolation Instrumentation,'' of Limiting Conditions
for Operation 3.3.6, during Modes 1 through 4. The current allowable
value was found to be non-conservative for operating Modes 1 through 4
because the basis for the specified value inappropriately credited the
containment purge exhaust filters, which are only required during
movement of irradiated fuel assemblies within containment. The current
allowable value remains acceptable during movement of irradiated fuel
assemblies within containment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change is associated with radiation effluent
monitoring and isolation of Containment Purge exhaust flow in the
event of a design basis SBLOCA [small break loss of coolant
accident]. The change is not associated with equipment or processes
which can initiate a design basis accident. Consequently, this
change does not affect the probability of an accident previously
evaluated.
The revised purge exhaust monitor allowable value will ensure
the monitors isolate the purge exhaust and will limit the offsite
doses associated with a SBLOCA to well within the limits of 10 CFR
100. This change serves to ensure the consequences of an accident
previously evaluated remain bounded by the plant's current licensing
basis. Therefore, the consequences of accidents previously evaluated
are not increased by 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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change is associated with radiation effluent
monitoring and isolation of Containment Purge exhaust flow in the
event of a design basis SBLOCA. The change is not associated with
equipment or processes which can initiate a design basis accident.
The change does not introduce new accident initiators or physical
changes in plant equipment.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change involves a conservative change in the
Containment Purge exhaust radiation monitor allowable value in TS
Table 3.3.6-1. The new allowable value reflects a change in the
monitor analytical limit which does not assume credit for the
Containment Purge exhaust filters. The proposed allowable value will
ensure the monitors will isolate the purge exhaust as assumed in the
existing design basis SBLOCA analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
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
[[Page 25048]]
amendments satisfy the criteria for categorical exclusion in accordance
with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no
environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: April 12, 2007.
Brief description of amendment: The amendment modifies the TMI-1
technical specifications related to control room envelope habitability
consistent with Technical Specification Task Force (TSTF) Traveler
TSTF-448.
Date of issuance: April 16, 2008.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 264.
Facility Operating License No. DPR-50. Amendment revised the
license and the technical specifications.
Date of initial notice in Federal Register: June 5, 2007 (72 FR
31100). The supplements dated January 18, 2008, and March 14, 2008,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed and did not
change the NRC staff's original proposed no significant hazards
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 16, 2008.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: January 22, 2007, as
supplemented on June 21, July 18, July 31, and October 15, 2007, and
January 24, February 14, March 5, and March 21, 2008.
Brief description of amendments: Change the Technical
Specifications (TSs) to support the transition to AREVA fuel and core
design methodologies.
Date of issuance: March 27, 2008.
Effective date: Date of issuance, to be implemented on Unit 1 prior
to startup from the 2008 refueling outage, and to be implemented on
Unit 2 prior to startup from the 2009 refueling outage.
Amendment Nos.: 246 and 274.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the TSs.
Date of initial notice in Federal Register: December 4, 2007 (72 FR
68208). The supplements dated January 24, February 14, March 5, and
March 21, 2008, 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 27, 2008.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: November 7, 2007.
Brief description of amendment: The amendment deletes License
Condition 2.F, which requires reporting of violations of certain other
requirements contained in Section 2.C of the license.
Date of issuance: April 15, 2008.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 206.
Facility Operating License No. NPF-21: The amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: December 4, 2007 (72 FR
68211) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 15, 2008.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: May 22, 2007.
Brief description of amendment: The amendment incorporates
technical specification (TS) changes based on Nuclear Regulatory
Commission (NRC)-approved TS Task Force (TSTF)-497-A, ``Changes to
Reflect Revision of 10 CFR 50.55a,'' Revision 0, as modified by NRC-
approved TSTF-497, ``Limit Inservice Testing Program [Surveillance
Requirements] SR 3.0.2 Application to Frequencies of Two years or
Less.'' Specifically, the amendment revises Palisades Nuclear Plant TS
Section 5.5.7, ``Inservice Testing Program,'' to update references to
the American Society of Mechanical Engineers code and applicability of
the provisions of SR 3.0.2.
Date of issuance: April 15, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 232.
Renewed Facility Operating License No. DPR-20: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 28, 2007 (72 FR
49575). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 15, 2008.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: October 18, 2007.
Brief description of amendment: The amendment revised the Technical
Specifications to change requirements related to emergency diesel
generator (EDG) fuel oil tank volume, EDG fuel oil testing and reactor
building crane inspections.
Date of Issuance: April 17, 2008.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 231.
Facility Operating License No. DPR-28: Amendment revised the
License and Technical Specifications.
[[Page 25049]]
Date of initial notice in Federal Register: December 18, 2007 (72
FR 71711).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated April 17, 2008.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: April 24, 2007, as supplemented by
electronic mail dated February 12, 2008.
Brief description of amendment: The change adds Optimized ZIRLO as
an acceptable fuel rod cladding material in the Waterford Steam
Electric Station, Unit 3, Technical Specification (TS) 5.3.1, ``Fuel
Assemblies.'' TS 5.3.1 currently identifies, in part, Zircaloy or ZIRLO
\PM\ fuel rod cladding as the allowable fuel rod cladding material.
Date of issuance: April 16, 2008.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 215.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 22, 2007 (72 FR
28720). The supplemental electronic mail dated February 12, 2008,
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 16, 2008.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 2, 2007, as supplemented by
letters dated January 17, March 10, and electronic mail dated March 24,
2008. In addition, Entergy submitted for review and approval the
revised emergency core cooling system (ECCS) performance analysis by
letter dated August 9, 2007, as supplemented by letter dated January
21, 2008; and a supplement to the ECCS performance analysis by letter
dated October 4, 2007, as supplemented by letter dated March 4, 2008.
Brief description of amendment: The changes to the technical
specifications add new analytical methods and modify the containment
average air temperature and safety injection tank level to support the
implementation of Combustion Engineering 16 x 16 Next Generation Fuel
(NGF) as defined in Westinghouse Topical Report WCAP-16500-P beginning
in Cycle 16 commencing after the spring 2008 refueling outage.
Date of issuance: April 15, 2008.
Effective date: As of the date of issuance and shall be shall be
implemented prior to startup following the spring 2008 refueling
outage.
Amendment No.: 214.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 11, 2007 (72
FR 51858). The supplemental letters dated January 17, and March 10,
2008, and electronic mail dated March 24, 2008, for changes to the TSs;
the supplemental letter dated January 21, 2008, for review and approval
of the revised ECCS performance analysis; and the supplemental letter
dated March 4, 2008, for review and approval of the supplement to the
ECCS performance analysis, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 2008.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station (Braidwood), Units 1 and 2, Will County, Illinois
Date of application for amendment: February 25, 2008, as
supplemented by letters dated March 27, 2008, and April 9, 2008.
Brief description of amendment: The amendments revise Technical
Specification (TS) 5.5.9, ``Steam Generator (SG) Program,'' and TS
5.6.9, ``Steam Generator (SG) Tube Inspection Report.'' For TS 5.5.9,
the amendment replaces the existing alternate repair criteria in the
provisions for SG tube repair criteria, during Braidwood, Unit 2,
Refueling Outage 13 and the subsequent operating cycle. For TS 5.6.9,
three new reporting requirements are added to the existing seven
requirements for Braidwood Station (Braidwood), Unit 2. These changes
only affect Braidwood, Unit 2; however, this action is docketed for
Braidwood, Units 1 and 2, because the TS are common to both units.
Date of issuance: April 18, 2008.
Effective date: As of the date of issuance and shall be implemented
prior to the return to service from Braidwood, Unit 2, spring 2008
Refueling Outage 13.
Amendment Nos.: Unit 1-150; Unit 2-150.
Facility Operating License Nos. NPF-72 and NPF-77: The amendment
revised the TSs and License.
Date of initial notice in Federal Register: March 11, 2008 (73 FR
13029).
The March 27, 2008, and April 9, 2008, supplemental letters
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 18, 2008.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315, Donald C. Cook
Nuclear Plant, Units 1 and 2 (DCCNP-1 and DCCNP-2), Berrien County,
Michigan
Date of application for amendments: February 29, 2008.
Brief description of amendments: The amendments revised the
licensing basis of ice condenser ice fusion time, specifying conditions
under which plant operation may proceed in less than 5 weeks after ice
baskets have been reloaded.
Date of issuance: April 16, 2008.
Effective date: As of the date of issuance, and shall be
implemented prior to Unit 1 entering Mode 4 at the end of the 2008
refueling outage.
Amendment No.: 303 (for DCCNP-1) and 286 (for DCCNP-2).
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Renewed Operating Licenses.
Date of initial notice in Federal Register: March 12, 2008 (73 FR
13253)
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated April 16, 2008.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of application for amendment: July 26, 2007, as supplemented
by letters dated October 3 and December 21, 2007, and February 29,
2008.
[[Page 25050]]
Brief description of amendment: The proposed amendment would add a
new reference to Technical Specification 6.9.1.14.a, which lists
documents that have been approved by the U.S. Nuclear Regulatory
Commission for use in determining the core operating limits. The new
reference is the Areva NP, Inc., Topical Report EMF-2103P-A,
``Realistic Large Break LOCA [Loss-Of-Coolant Accident] Methodology for
Pressurized Water Reactors.''
Date of issuance: April 10, 2008.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No. 311.
Facility Operating License No. DPR-79: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: August 28, 2007 (72 FR
49583). The supplemental letters dated October 3 and December 21, 2007,
and February 29, 2008, provided additional information that clarified
the application, did not expand the scope of the application as
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 10, 2008.
No significant hazards consideration comments received: No.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona
Date of amendment request: April 10, 2008.
Brief Description of amendment request: The proposed amendment
would revise Technical Specification (TS) 3.5.5, Refueling Water Tank
(RWT) to increase the minimum required RWT level indications and the
corresponding borated water volumes in TS Figure 3.5.5-1, ``Minimum
Required RWT Volume,'' by 3 percent. This change will ensure that there
is adequate water volume available in the RWT to ensure that the
engineered safety feature pumps and the new containment recirculation
sump strainers will meet their design functions during loss-of-coolant
accidents.
Date of publication of individual notice in Federal Register: April
17, 2008 (73 FR 20961).
Expiration date of individual notice: May 1, 2008.
Dated at Rockville, Maryland, this 28th day of April, 2008.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E8-9679 Filed 5-5-08; 8:45 am]
BILLING CODE 7590-01-P