[Federal Register Volume 75, Number 182 (Tuesday, September 21, 2010)]
[Notices]
[Pages 57521-57532]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-23388]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0297]
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) is publishing this regular biweekly notice. The Act
requires the Commission publish notice of any amendments issued, or
proposed to be issued and grants the Commission the authority to issue
and make immediately effective any amendment to an operating license
upon a determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 26, 2010, to September 8, 2010. The
last biweekly notice was published on September 7, 2010 (75 FR 54390-
54400).
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 Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
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
[[Page 57522]]
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules,
Announcements and Directives Branch (RADB), TWB-05-B01M, 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 faxed to the RADB at 301-492-3446.
Documents may be examined, and/or copied for a fee, at the NRC's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
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 O1F21, 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 by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
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
requestor/petitioner 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 requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. 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 requestor/petitioner to relief. A requestor/
petitioner 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, 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.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory 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 an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the participant should
contact the Office of the Secretary by e-mail at
[email protected], or by telephone at (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
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
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. System requirements for accessing
the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at
http://www.nrc.gov/site-help/e-submittals.html. Participants may
attempt to use other software not listed on the Web site, but should
note that the NRC's E-Filing system does not support unlisted software,
and the NRC Meta System Help Desk will not be able to offer assistance
in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through EIE, users will be required to install a Web
browser plug-
[[Page 57523]]
in from the NRC Web site. Further information on the Web-based
submission form, including the installation of the Web browser plug-in,
is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant 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
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing 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
E-Filing 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 using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, 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. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
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, or the 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, unless an NRC regulation or other law requires submission of
such information. 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.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment
application, 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 O1F21, 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. Persons who do not have
access to ADAMS or who encounter problems in accessing the documents
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 20, 2010.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) Limiting Condition for Operation
(LCO) 3.7.1.2, ``Emergency Feedwater System,'' to clarify the
acceptability of transitioning from Mode 4 to Mode 3 with the turbine-
driven emergency feedwater (EFW) pump inoperable but available. This
proposal would grant an exception to TS LCO 3.0.4 and Surveillance
Requirement 4.0.4 allowing entry into operational Mode 3 with TS LCO
equipment, the turbine-driven EFW pump, associated with a shutdown
action inoperable.
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 addition of an exception to TS LCO 3.0.4 for entry
into Mode 3 during a plant startup for the turbine-driven EFW pump
for a plant condition when the turbine driven EFW pump would be
unable to complete its post maintenance activities (i.e. dynamic
final calibration of the governor valve speed control unit governor
control system) due to insufficient steam pressure in the steam
generator secondary side and then to complete the quarterly IST
[Inservice Testing] and 18 month EFAS [Engineered Safety Features
Actuation System] SR [Surveillance Requirement] within the allowance
of the delay of the respective SR is administrative in nature.
This change will clarify that the turbine-driven EFW pump is not
required to fully demonstrate operability (i.e. be inoperable
pending completion of the quarterly IST and 18 month EFAS SR) during
plant startup prior to entry into Mode 3 under the conditions and
for the period as provided in the quarterly IST and 18 month EFAS SR
as granted by the NRC [Nuclear Regulatory Commission] in Reference
7.1 [NRC letter to Waterford 3 dated October 4, 2001, Waterford
Steam Electric Station--Unit 3, Issuance of Amendment RE: Emergency
Feedwater System (TAC No MB2010), Agencywide Documents Access and
Management System (ADAMS) Accession No. ML012840538]. When the plant
enters Mode 3 during plant
[[Page 57524]]
startup, the turbine-driven EFW pump is available (i.e., there is a
reasonable expectation that once sufficient steam pressure is
available to the turbine-driven EFW pump turbine, it will be able to
successfully complete the quarterly IST and 18 month EFAS
surveillance requirements to fully demonstrate operability).
Prior to entry into Mode 2, surveillance requirement testing of
various combinations of EFW pumps and valves will ensure ALL
required EFW system flow paths and equipment (which includes the
turbine-driven EFW pump) are demonstrated operable before sufficient
core heat is generated that would require the operation of the EFW
System during a subsequent shutdown.
Since the two motor-driven EFW pumps are required to be operable
when entering Modes 3 from Mode 4, then for the worst case
postulated accident scenario during plant startup, with the turbine-
driven EFW pump considered inoperable but available (utilizing the
exception to TS LCO 3.0.4 as tied to the quarterly IST and 18 month
EFAS SR for fully demonstrating operability of the turbine-driven
EFW pump), the EFW System safety function of achieving shutdown
cooling entry conditions would be met.
This request is merely a clarification and does not present any
change to equipment operation, design or practices. The proposed
clarification is not an accident initiator and will not adversely
affect plant safety functions. The EFW System capability to provide
its specified function of being able to achieve shutdown cooling
entry conditions of the Reactor Coolant [S]ystem is unchanged by
this clarification.
Therefore, this 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 addition of an exception to TS LCO 3.0.4 for entry
into Mode 3 during a plant startup for the turbine-driven EFW pump
for a plant condition when the turbine-driven EFW pump would be
unable to complete its post maintenance activities (i.e. dynamic
final calibration of the governor valve speed control unit governor
control system) due to insufficient steam pressure in the steam
generator secondary side and then to complete the quarterly IST and
18 month EFAS SR within the allowance of the delay of the respective
SR is administrative in nature.
This change will clarify that the turbine-driven EFW pump is not
required to fully demonstrate operability (i.e. be inoperable
pending completion of the quarterly IST and 18 month EFAS SR) during
plant startup prior to entry into Mode 3 under the conditions and
for the period as provided in the quarterly IST and 18 month EFAS SR
as granted by the NRC in Reference 7.1. When the plant enters Mode 3
during plant startup, the turbine-driven EFW pump is available (i.e.
there is a reasonable expectation that once sufficient steam
pressure is available to the turbine-driven EFW pump turbine, it
will be able to successfully complete the quarterly IST and 18 month
EFAS surveillance requirements to fully demonstrate operability).
Prior to entry into Mode 2, surveillance requirement testing of
various combinations of EFW pumps and valves will ensure ALL
required EFW system flow paths and equipment (which includes the
turbine-driven EFW pump) are demonstrated operable before sufficient
core heat is generated that would require the operation of the EFW
System during a subsequent shutdown.
The addition of this exception to TS LCO 3.0.4 for the turbine-
driven EFW pump introduces no new mode of plant operation and does
not alter the EFW System functional capability. The scope of this
proposed change does not establish a potential new accident
precursor. This proposed change will not change the design,
configuration or method of operation of the EFW System. No new
possibility for an accident is introduced by the proposed
clarification.
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 addition of an exception to TS LCO 3.0.4 for entry
into Mode 3 during a plant startup for the turbine-driven EFW pump
for a plant condition when the turbine-driven EFW pump would be
unable to complete its post maintenance activities (i.e. dynamic
final calibration of the governor valve speed control unit governor
control system) due to insufficient steam pressure in the steam
generator secondary side and then to complete the quarterly IST and
18 month EFAS SR within the allowance of the delay of the respective
SR is administrative in nature.
This change will clarify that the turbine-driven EFW pump is not
required to fully demonstrate operability (i.e. be inoperable
pending completion of the quarterly IST and 18 month EFAS SR) during
plant startup when entering Mode 3 under the conditions and for the
period as provided in the quarterly IST and 18 month EFAS SR as
granted by the NRC in Reference 7.1. When the plant enters Mode 3
during plant startup, the turbine-driven EFW pump is available (i.e.
there is a reasonable expectation that once sufficient steam
pressure is available to the turbine-driven EFW pump turbine, it
will be able to successfully complete the quarterly IST and 18 month
EFAS surveillance requirements to fully demonstrate operability).
Prior to entry into Mode 2, surveillance requirement testing of
various combinations of EFW pumps and valves will ensure ALL
required EFW system flow paths and equipment (which includes the
turbine-driven EFW pump) are demonstrated operable before sufficient
core heat is generated that would require the operation of the EFW
System during a subsequent shutdown.
The proposed clarification does not adversely affect Emergency
Feedwater equipment operating practices. The EFW System has the same
capabilities as before to mitigate accidents. Surveillance
requirements are not reduced by the proposed change. The EFW System
capability to provide its specified function of being able to
achieve shutdown cooling entry conditions of the Reactor Coolant
System following a worst case postulated accident is unchanged by
this clarification.
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: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
NextEra Energy Point Beach, LLC (the licensee), Docket Nos. 50-266 and
50-301, Point Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two
Creeks, Manitowac County, Wisconsin
Date of amendment request: April 7, 2009, as supplemented by
letters dated June 17, September 11, November 20, November 30, and
December 8 of 2009; and February 11, February 25, April 22, April 30,
July 21, July 28, and August 2 of 2010.
Description of amendment request: The proposed amendment would
revise Reactor Protection System (RPS) and Engineered Safety Feature
Actuation System (ESFAS) instrumentation setpoints for the PBNP, Units
1 and 2. The revised Technical Specification (TS) allowable values are
specified in Tables 3.3.1-1 and 3.3.2-1 for RPS and ESFAS,
respectively. These changes were originally included as part of the
April 7, 2009, extended power uprate (EPU) license amendment request,
but subsequently divided into a separate licensing action for
independent technical review. The proposed changes include both EPU and
non-EPU related changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or
[[Page 57525]]
consequences of an accident previously evaluated?
Response: No.
The proposed changes to the TSs will ensure that the results of
previously evaluated accidents at the uprated conditions remain
within the acceptance criteria. The proposed RPS and ESFAS setpoint
changes provide appropriate values for operation at EPU conditions.
The revised TS allowable values have been calculated to account for
new EPU analytical limits, instrument uncertainties, and instrument
drift. The proposed RPS and ESFAS setpoint changes are considered in
the safety analysis for the affected RPS and ESFAS functions, and do
not significantly increase the probability or consequences of the
accidents previously evaluated and the setpoint changes considered
in the safety analysis continue to meet the applicable acceptance
criteria. The safety analyses for these accidents have been
performed at the EPU power level and demonstrated acceptable
results.
The proposed changes will ensure that the instruments actuate as
assumed to mitigate accidents previously evaluated. The proposed
changes will not significantly affect accident initiators or
precursors and will not alter or prevent the ability of systems,
structures, or components from performing the intended safety
function to meet the applicable acceptance limits for the accidents
and events.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The change does not involve a physical alteration of the plant
or change the methods governing normal plant operation. The change
does not alter assumptions made in the safety analyses, but ensures
that the instruments behave as assumed in the accident analysis. The
proposed change is consistent with the safety analysis assumptions.
The proposed RPS and ESFAS Limiting Safety System Setting (LSSS)
changes do not create the possibility of a new or different type of
accident due to operation at EPU conditions. The revised TS LSSS
values have been calculated to account for new EPU analytical limits
and known instrument uncertainties. The proposed RPS and ESFAS
setpoint changes are used in the safety analysis for the affected
RPS and ESFAS functions, and do not significantly affect these
accidents or the applicable acceptance criteria.
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 changes clarify the TS requirements for
instrumentation to ensure that the automatic protection action will
correct the abnormal situation before a safety limit is exceeded.
The proposed change also revises the TSs to enhance the controls
used to maintain the variables and systems within the prescribed
operating ranges, in order to ensure that automatic protection
actions occur to initiate the operation of systems and components
important to safety as assumed in the accident analysis. No change
is made to the accident analysis assumptions.
The proposed changes to the RPS and ESFAS setpoint TSs provide
adequate margin such that PBNP Units 1 and 2 can be operated in a
safe manner at EPU conditions. No new accident scenarios, failure
mechanisms, or single failures are introduced as a result of the
proposed changes. All systems, structures and components previously
assumed for the mitigation of an event remain capable of fulfilling
their intended function. The proposed changes will not have any
significant effect on the 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: William Blair, Senior Attorney, NextEra
Energy Point Beach, LLC, P. O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Robert J. Pascarelli.
NextEra Energy Point Beach, LLC (the licensee), Docket Nos. 50-266 and
50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks,
Manitowac County, Wisconsin
Date of amendment request: April 7, 2009, as supplemented by
letters dated June 17 (two letters), September 11, September 25,
October 9, November 20 (two letters), November 21 (two letters),
November 30, December 8, and December 16 of 2009; and January 7,
January 8, January 22, February 11, February 25, March 3, April 15,
April 22, July 8, July 28, August 2, August 9, and August 24 of 2010.
Description of amendment request: The proposed amendment would
change the auxiliary feedwater (AFW) system design and Technical
Specifications (TS) 3.7.5, ``Auxiliary Feedwater (AFW),'' and TS 3.7.6,
``Condensate Storage Tank (CST),'' resulting from (1) modifications to
the AFW system to support requirements for transients and other
accidents at extended power uprate (EPU) conditions; (2) installation
of main feedwater isolation valves to support accident mitigation by
ensuring that containment pressure does not exceed safety analysis
limits; (3) automatic AFW switchover from a CST suction source to a
safety-related Service Water (SW) source; and (4) setpoint changes
supporting the aforementioned physical modifications. These changes
were originally included as part of the April 7, 2009, EPU license
amendment request, but subsequently divided into a separate licensing
action for independent technical review. The upgrades and modifications
to the AFW system are being installed to provide additional capacity
and reliability for the system. Although the proposed changes are also
designed to support the requirements for transients and other accidents
at EPU conditions, the proposed changes for this amendment are being
evaluated using the current licensing basis.
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. The NRC staff performed its own analysis, 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 design functions of the AFW system will not be altered by
the proposed change. The AFW system will continue to perform its
original intended design function, mitigating the consequences of
accidents previously evaluated. The proposed changes will not
significantly affect accident initiators or precursors. No new
accident scenarios, failure mechanisms, or single failures are
introduced as a result of the proposed modifications.
Implementation of the new AFW system design and the proposed
changes to TS 3.7.5 was evaluated against the current analysis of
record for the current licensed power level at PBNP, Units 1 and 2.
The current analyses remain applicable or are unaffected by
implementation of the new AFW system and associated TS changes, with
the exception of the steam line break containment response and steam
generator tube rupture (SGTR) radiological consequences. These two
accidents were reanalyzed with the current licensing basis for the
AFW modifications and the results were acceptable with the revised
minimum and maximum AFW flow rates and pump start timing.
Therefore, the consequences of accidents previously evaluated
for the current licensed power level are not significantly
increased.
A proposed change to TS 3.7.6 changes the surveillance
requirement (SR) for minimum CST water inventory to be maintained to
supply AFW pump suction in the event of a Station Blackout, when the
safety-related AFW suction source from the SW system is not
available. The proposed TS 3.7.6 SR increases the current minimum
required inventory to account for the increased flow rates from the
new AFW system design,
[[Page 57526]]
suction piping losses, instrument uncertainties, vortex prevention,
net positive suction head (NPSH) requirements, and the suction of
the AFW pumps under various combinations of CST and plant units in
operation. This change to the minimum required CST level inventory
will not increase the probability or consequences of previously
evaluated accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not introduce a new mode of plant
operation. The proposed changes involving the AFW system do not
significantly alter any design basis accident or event response. The
proposed changes will not significantly affect accident initiators
or precursors. The AFW system will continue to perform its design
function. No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of the proposed modifications.
All systems, structures, and components previously assumed for the
mitigation of an event remain capable of fulfilling their intended
design function. The new AFW system design and proposed changes to
TS 3.7.5 and the proposed increase in CST inventory in TS 3.7.6 do
not create the possibility of a new or different kind of accident or
event.
As previously discussed, implementation of the new AFW system
design and the proposed changes to TS 3.7.5 was evaluated against
the current analysis of record for the current licensed power level
at PBNP, Units 1 and 2. The current analyses remain applicable or
are unaffected by implementation of the new AFW system and
associated TS changes, with the exception of the steam line break
containment response and steam generator tube rupture (SGTR)
radiological consequences. These two accidents were reanalyzed with
the current licensing basis for the AFW modifications and the
results are acceptable with the revised minimum and maximum AFW flow
rates and pump start timing. The AFW system design change, the
changes to TS .3.7.5, and the increase in required CST inventory
established in TS 3.7.6, are not significant accident initiators or
precursor and will not create the possibility of a new or different
kind of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The upgrade to the AFW system is being made to support
requirements for transients and other accidents at EPU conditions.
This modification to the AFW system will provide additional capacity
and reliability for the system. As such, the proposed amendment does
not involve a significant reduction in safety.
The analyses and evaluations of the Nuclear Steam Supply System
(NSSS) and Balance of Plant (BOP) systems based on completion of the
required modifications, confirm that the systems and components will
function as designed and demonstrate that the NSSS and BOP systems
and components meet all applicable design and licensing requirements
at the uprated power level.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
Based on the above review, it appears that the three standards of
10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: William Blair, Senior Attorney, NextEra
Energy Point Beach, LLC,.P. O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Robert J. Pascarelli.
NextEra Energy Point Beach, LLC (the licensee), Docket Nos. 50-266 and
50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks,
Manitowac County, Wisconsin
Date of amendment request: June 1, 2010, as supplemented by letter
dated July 9, 2010.
Description of amendment request: The proposed amendment consists
of revising the current license basis regarding a postulated reactor
vessel head (RVH) drop event to conform to the NRC-endorsed guidance of
Nuclear Energy Institute (NEI) 08-05, ``Industry Initiative on Control
of Heavy Loads,'' Revision 0. The proposed change to the license basis
will revise Chapter 14.3.6, ``Reactor Vessel Head Drop Event,'' of the
Final Safety Analysis Report. The current license basis assumes failure
of the reactor coolant system (RCS) boundary caused by the predicted
maximum downward displacement of the reactor vessel which would sever
all 36 bottom-mounted instrument (BMI) conduit tubes. The new analysis
demonstrates that a postulated RVH drop would not result in a loss of
RCS inventory caused by an RCS boundary failure, since the BMI conduits
would remain intact.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment is limited in scope to a postulated RVH
drop and the administrative controls in place, which limit the
height of the RVH lift, ensuring an actual drop is bounded by the
analyses of record.
Incorporation of the analysis performed in accordance with NRC-
approved guidance, which demonstrates bottom-mounted instrumentation
(BMI) conduits will not sever following a postulated RVH drop, does
not increase the probability or consequences of a previously
evaluated accident. The evaluation, in fact, demonstrates that if
the postulated RVH drop occurred, the consequences would be
significantly less than are now assumed because the ability to
maintain a coolable geometry in the core has not been compromised.
In accordance with NRC-endorsed methodology contained in NEI 08-05,
which states, ``Previous evaluations have indicated that the
consequences of impacts between the upper vessel internals and the
fuel were not significant with respect to public health and
safety,'' a revised radiological analysis was not performed.
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 amendment is limited in scope to a postulated RVH
drop and the administrative controls in place, which limit the
height of the reactor RVH lift, ensuring an actual drop is bounded
by the analysis of record.
Incorporation of the analysis performed in accordance with NRC-
approved guidance, which demonstrates BMI conduits will not sever
following a postulated RVH drop, does not create the possibility of
a new or different kind of accident from any accident previously
evaluated. The proposed amendment does not: (1) Operate equipment in
alignments or in a manner different form that previously evaluated
in the FSAR; (2) install, remove or modify equipment important to
safety; or (3) introduce new failure modes or effects for any
existing system, structure or component.
Therefore, the proposed change does not create the possibility
of a new or different kind of any accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment is limited in scope to a postulated RVH
drop and the administrative controls in place, which limit the
height of the reactor RVH lift, ensuring an actual drop is bounded
by the analysis of record.
Incorporation of the analysis performed in accordance with NRC-
approved guidance, which demonstrates BMI conduits will not sever
following a postulated RVH drop, does not involve a significant
reduction in the margin of safety. The evaluation, in fact,
demonstrates that if the postulated RVH drop occurred, the
consequences would be significantly less than are now assumed
because the ability to maintain a coolable geometry in the core has
not been compromised.
[[Page 57527]]
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: William Blair, Senior Attorney, NextEra
Energy Point Beach, LLC, P. O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Robert J. Pascarelli.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: January 21, 2010.
Description of amendment request: The licensee proposed to amend
the MNGP Technical Specifications to allow operation in the Maximum
Extended Load Line Limit Analysis Plus (MELLLA+) expanded domain. The
licensee stated that the Nuclear Regulatory Commission (NRC) had
previously approved various aspects of the MELLLA+ methodology, but
that the current application is the first plant-specific use of such
methodology. The amendment would include changes to the Technical
Specifications to: (1) Prohibit the use of the MELLLA+ expanded
operating domain when in single loop operation; (2) change the
allowable value for Average Power Range Monitor (APRM)-Simulated
Thermal Power--High; (3) eliminate an unnecessary surveillance
requirement; (4) require certain content in the Core Operating Limits
Report. Approval of this amendment would allow the licensee to
implement operational changes to provide increased operational
flexibility for power maneuvering, to compensate for fuel depletion,
and to maintain efficient power distribution in the reactor core
without the need for more frequent rod pattern changes. MELLLA+ would
increase the operating range to the Extended Power Uprate rated thermal
power at 80 percent flow; thus creating a 20 percent flow-control
window. By operating in the MELLLA+ domain, a significantly lower
number of control rod movements will be required than in the present
operating domain. This would represent a significant improvement in
operating flexibility. It also provides safer operation, because
reducing the number of control rod manipulations would minimize the
likelihood of fuel failures, and reduce the likelihood of accidents
initiated by reactor maneuvers.
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 is reproduced below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability (frequency of occurrence) of [d]esign [b]asis
[a]ccidents occurring is not affected by the MELLLA+ operating
domain, because MNGP continues to comply with the regulatory and
design basis criteria established for plant equipment. Further, a
probabilistic risk assessment demonstrates that the calculated core
damage frequencies do not significantly change due to the MELLLA+.
There is no change in consequences of postulated accidents, when
operating in the MELLLA+ operating domain compared to the operating
domain previously evaluated. The results of accident evaluations
remain within the NRC[-]approved acceptance limits.
The spectrum of postulated transients has been investigated and
is 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) is still met. Continued
compliance with the SLMCPR 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 for the MELLLA+ operating domain conditions (pressure,
temperature, flow, and radiation) and were found to meet their
acceptance criteria for allowable stresses and overpressure margin.
Challenges to the containment were evaluated and the containment
and its associated cooling systems continue to meet the current
licensing basis. The calculated post[-]LOCA [loss-of-coolant
accident] suppression pool temperature remains acceptable.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Equipment that could be affected by the MELLLA+ operating domain
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. The
MELLLA+ operating domain uses developed technology and applies it
within the capabilities of existing plant safety-related equipment
in accordance with the regulatory criteria (including NRC approved
codes, standards and methods). No new accident or event precursor
has been identified.
The-MNGP TS require revision to implement the MELLLA+ operating
domain. 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 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 MELLLA+ operating domain affects only design and operational
margins. Challenges to the fuel, reactor coolant pressure boundary,
and containment were evaluated for the MELLLA+ operating domain
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.
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
proposed amendment involves no significant hazards consideration.
Attorney for the licensee: Peter M. Glass, Assistant General
Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN
55401.
NRC Branch Chief: Robert J. Pascarelli.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of amendment request: June 14, 2010.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to allow the use of a dedicated on-
line core power distribution monitoring system (PDMS) to enhance
surveillance of core thermal limits. The PDMS to be used at Prairie
Island Nuclear Generating Plant, Units 1 and 2, is the Westinghouse
proprietary core analysis system called the Best Estimate Analyzer for
Core Operations--Nuclear (BEACON\TM\).
[[Page 57528]]
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 PDMS performs continuous core power distribution monitoring
with data input from existing plant instrumentation. The system
passively supports Technical Specification (TS) surveillances which
ensure that core power distribution is within the same limits that
are currently prescribed. Further, the proposed TS Actions are
comparable to existing operator actions such that no new plant
configurations are prompted by the proposed change. The system's
physical interface with plant equipment is limited to an electronic
link from a new workstation to the plant process computer. The
system is passive in that it provides no control or alarm functions,
and does not promote any new plant configuration which would affect
the initiation, probability, or consequences of a previously-
evaluated accident. Continuous on-line core monitoring through the
use of PDMS provides significantly more information about the power
distributions present in the core than is currently available. This
system performance may result in an earlier determination of an
adverse core condition and more time for operator action, thus
reducing the probability of an accident occurrence and reduced
consequences should a previously-evaluated accident occur.
By virtue of its inherently passive surveillance function and
limited interface with plant systems, structures, or components, the
proposed changes will not result in any additional challenges to
plant equipment that could increase the probability or occurrence of
any previously-evaluated accident. Further, the proposed changes
will ensure conformance to the same core power distribution limits
that form the basis for initial conditions of previously evaluated
accidents. Thereby, the proposed changes will not affect the
consequences of any previously-evaluated accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence 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 system's physical interface with plant equipment is limited
to an electronic link from a new workstation to the plant process
computer. The system is passive in that it provides no control or
alarm functions, and the proposed changes (including operator
actions prescribed by the proposed TS) do not promote any new plant
configuration which would create the possibility for an accident of
a new or different type.
The NRC previously evaluated the effects of using the PDMS to
monitor core power distribution parameters and determined that all
design standards and applicable safety criteria limits are met. The
Technical Specifications will continue to require operation within
the required core operating limits, and appropriate actions will
continue to be taken when or if limits are exceeded. Thus, the
reactor core will continue to be operated within its reference
bounds of design such that an accident of a new or different type is
not credible.
The proposed change, therefore, does 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.
No margin of safety is adversely affected by the implementation
of the PDMS. The margins of safety provided by current TS
requirements and limits remain unchanged, as the TS will continue to
require operation within the core limits that are based on NRC-
approved reload design methodologies. The proposed change does not
result in changes to the core operating limits. Appropriate measures
exist to control the values of these cycle-specific limits, and
appropriate actions will continue to be specified and taken when
limits are violated. Such actions remain unchanged.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Robert J. Pascarelli.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 18, 2010.
Description of amendment request: The proposed amendments would
reduce system/equipment diversity in isolation of low-pressure residual
heat removal (RHR) system from high-pressure reactor coolant system
(RCS). The change will allow similarly qualified pressure transmitters
to be used in more than one RHR train as necessary regardless of
manufacturer of the transmitters.
The valves separating the RHR from the RCS are to have independent
and diverse interlocks to prevent both from opening unless the RCS
pressure is below that of the RHR in compliance with the Nuclear
Regulatory Commission's Technical Position ICSB-3, ``Isolation of Low
Pressure Systems from the High Pressure Reactor Coolant System.''
Consequently, the change would result in more than minimal increase in
the likelihood of a malfunction of systems, structures, or components
important to safety as previously evaluated in the plants' Updated
Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability consequences of an accident previously evaluated?
Response: No.
The proposed change revising the justification for diversity
associated with the RHR isolation valves will not cause an accident
to occur and will not result in any change in the operation of the
associated accident mitigation equipment. The proposed changes will
not revise the operability requirements (e.g., leakage limits) for
the RHR system. The design-basis accidents will remain the same
postulated events described in the STP Unit 1 and Unit 2 Updated
Final Safety Analysis Report[,] and the consequences of the design-
basis accidents will remain the same.
Therefore, the proposed changes will not increase the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes will not alter the plant configuration or
require any unusual operator actions. The proposed changes will not
alter the way any structure, system, or component functions, and
will not significantly alter the manner in which the plant is
operated. The response of the plant and the operators following an
accident will not be different. In addition, the proposed changes do
not introduce any new failure modes. In the event the RHR system is
overpressurized by the RCS, all leakages originating from RHR
components will be detected by the Reactor Coolant Pressure Boundary
Leakage Detection System as discussed in the STP UFSAR [Updated
Final Safety Analysis Report].
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
analyzed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to revise the rationale for diversity
associated with RHR system isolation valve operation will not
[[Page 57529]]
cause an accident to occur and will not result in any change in the
operation of the associated accident mitigation equipment. The
operability requirements for the isolation valves have not been
changed, and the RHR system will continue to function as assumed in
the safety analysis. In addition, the proposed changes will not
adversely affect equipment design or operation, and there are no
changes being made to required safety limits or safety system
settings that would adversely affect plant safety.
Therefore, the proposed changes will not result in a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 18, 2010.
Description of amendment request: The proposed amendments would
revise the Technical Specification (TS) 6.8.3.l, ``Containment Post-
Tensioning System Surveillance Program.'' TS 6.8.3.l states that the
containment post-tensioning system surveillance program shall be in
accordance with American Society of Mechanical Engineers (ASME) Code,
Section XI, Subsection IML, 1992 Edition with 1992 Addenda, as
supplemented by 10 CFR 50.55a(b)(2)(viii). The current inspection
interval of South Texas Project (STP), Units 1 and 2 ends in September
2010. The proposed amendments will provide for updating the
surveillance program consistent with the updated edition of the ASME
Code, Section XI as required by 10 CFR 50.55a.
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 Technical Specification change removes the specific
edition of the ASME [C]ode to be applied. Inspection practices will
continue to be consistent with the approved ASME [C]ode edition. The
proposed change is consistent with NUREG-1481 [guidance].
Therefore, the proposed changes will not increase the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes will not alter the plant configuration (no
new or different type of equipment will be installed) or require any
unusual operator actions. The proposed changes will not alter the
way any structure, system, or component functions, and will not
significantly alter the manner in which the plant is operated. The
response of the plant and the operators following an accident will
not be different. In addition, the proposed change does not
introduce any new failure modes.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
analyzed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed Technical Specification change removes the specific
edition of the ASME [C]ode to be applied. Inspection practices will
continue to be consistent with the approved ASME [C]ode edition. The
change is consistent with NUREG-1481 guidance.
Therefore, the proposed changes will not result in a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: June 28, 2010.
Description of amendment requests: The proposed amendments request
correction of an oversight in previous amendments (Amendment No. 185 to
Facility Operating License No. NPF-76 and Amendment No. 172 to Facility
Operating License No. NPF-80) that revised the Technical Specifications
(TSs) regarding control room envelope (CRE) habitability in accordance
with TS Task Force (TSTF) Traveler No. 448, Revision 3. In its
application for those previous amendments, STP Nuclear Operating
Company (STPNOC) did not specify what shutdown actions would be taken
if required actions for an inoperable CRE boundary were not met. This
was inconsistent with TSTF-448. The proposed amendments would correct
this oversight. STPNOC also requested to add a note to the required
actions for inoperable CRE boundary to clarify that the boundary is not
a required system, subsystem, train, component, or device that depends
on a diesel generator as a source of emergency power. This change would
clarify the application of TS action 3.8.1.1, ``AC Sources, DC Sources,
and Other Power Distribution,'' when the CRE is inoperable.
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 add the shutdown actions to TS ACTION
3.7.7.d is consistent with Nuclear Regulatory Commission (NRC)
noticed Industry/Technical Specification Task Force (TSTF) Standard
Technical Specification (STS) change TSTF-448 Revision 3, which has
been approved by an NRC safety evaluation.
The proposed change to add a note to the required action for an
inoperable control room envelope boundary does not change the design
function of the Control Room Makeup and Cleanup Filtration Systems
or the design function of the A.C. Sources, D.C. Sources, and Onsite
Power Systems or how these systems operate. The change only
clarifies that the Control Room Envelope boundary is not a required
system, subsystem, train, component, or device that depends on a
diesel generator as a source of emergency power.
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 to add the shutdown actions to TS ACTION
3.7.7.d is consistent with Nuclear Regulatory Commission (NRC)
noticed Industry/Technical Specification Task Force (TSTF) Standard
Technical Specification (STS) change TSTF-448 Revision 3, which has
been approved by an NRC safety evaluation.
The proposed change to add a note to the required action for an
inoperable control room envelope boundary does not change the design
of the Control Room Makeup and
[[Page 57530]]
Cleanup Filtration Systems or the design function of the A.C.
Sources, D.C. Sources, and Onsite Power Systems. The change only
clarifies that the Control Room Envelope boundary is not a required
system, subsystem, train, component, or device that depends on a
diesel generator as a source of emergency power.
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 to a
margin of safety?
Response: No.
The proposed change to add the shutdown actions to TS ACTION
3.7.7.d is consistent with Nuclear Regulatory Commission (NRC)
noticed Industry/Technical Specification Task Force (TSTF) Standard
Technical Specification (STS) change TSTF-448 Revision 3, which has
been approved by an NRC safety evaluation.
The proposed change to add a note to the required action for an
inoperable control room envelope boundary does not change any safety
margins associated with operation of the Control Room Makeup and
Cleanup Filtration Systems or any safety margins associated with the
A.C. Sources, D.C. Sources, and Onsite Power Systems. The change
only clarifies that the Control Room Envelope boundary is not a
required system, subsystem, train, component, or device that depends
on a diesel generator as a source of emergency power.
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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action, see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management System (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].
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application of amendments: August 31, 2009, as supplemented
April 14, 2010.
Brief description of amendments: The amendments revised the
Technical Specifications to allow one of the two required 230 kV
switchyard 125 Vdc power sources (batteries) to be inoperable for up to
10 days for the purpose of replacing an entire battery bank and
performing the required testing.
Date of Issuance: August 30, 2010.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 370, 372, 371.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 9, 2010 (75 FR
10828).
The supplement dated April 14, 2010, 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 August 30, 2010.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana
Parish, Louisiana
Date of amendment request: August 10, 2009, as supplemented by
letters dated December 8, 2009, and April 22, June 16, and August 17,
2010, and by emails dated June 29, July 12, and July 28, 2010.
Brief description of amendment: The amendment revised the TSs for
the RBS to support operation with 24-month fuel cycles. By letter dated
June 16, 2010, Entergy withdrew its proposed changes to TS 3.3.8
regarding the change to the degraded voltage instrumentation allowable
values as indicated on Table 3.3.8.1-1 and to extend the Surveillance
Requirement (SR) 3.3.8.1.3 and SR 3.3.8.1.4 from 18 to 24 months. By
letter dated August 17, 2010, Entergy withdrew the request for not
revising SR 3.3.8.1.4 and requested that this SR be extended as
originally requested.
Date of issuance: August 31, 2010.
Effective date: As of the date of issuance and shall be implemented
180 days from the date of issuance.
Amendment No.: 168.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 20, 2009 (74 FR
53776).
The supplements dated December 8, 2009, April 22, June 16, and
August 17, 2010, and emails dated June 29, July 12, and July 28, 2010,
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 August 31, 2010.
[[Page 57531]]
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: August 25, 2009 supplemented by
letter dated May 3, 2010.
Brief description of amendment: The amendment modifies technical
specification 5.5.14, ``Containment Leakage Rate Testing Program,'' to
allow a one-time extension to the 10-year frequency for the next 10 CFR
Part 50 Appendix J, Option B, Type A, containment integrity leakage
test (ILRT) or Type A test at Palisades Nuclear Plant. This amendment
permits the existing ILRT frequency to be extended from 10 years (120
months) to approximately 11.25 years (135 months). This amendment also
prevents the necessity of performing a Type A test six months prior to
the 10th anniversary of the completion of the last Type A test, which
was completed on May 3, 2001.
Date of issuance: August 23, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 240.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 20, 2009 (74 FR
53777).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated August 23, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: October 23, 2008, as
supplemented by letters dated September 28, and November 18, 2009,
March 29, and August 3, 2010.
Brief description of amendments: The amendments revise the
Technical Specifications to support the application of alternative
source term methodology with respect to the loss-of-coolant accident
and the fuel-handling accident.
Date of issuance: September 6, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 197, 184.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: April 7, 2009 (74 FR
15771).
The September 28, and November 18, 2009, March 29, and August 3,
2010 supplements contained clarifying information and did not change
the NRC staff's initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 6, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of application for amendments: August 31, 2009.
Brief description of amendments: The amendments modify the PBAPS
Technical Specifications (TS) by relocating specific surveillance
frequencies to a licensee-controlled program with the implementation of
Nuclear Energy Institute (NEI) 04-10, ``Risk-Informed Technical
Specifications Initiative 5b, Risk-Informed Method for Control of
Surveillance Frequencies.'' Additionally, the change adds a new
program, the Surveillance Frequency Control Program, to TS Section 5,
Administrative Controls. The changes are based on NRC-approved Industry
Technical Specifications Task Force (TSTF) Traveler 425, Revision 3,
``Relocate Surveillance Frequencies to Licensee Control--Risk Informed
Technical Specification Task Force Initiative 5b,'' with optional
changes and variations as described in Attachment 1, Section 2.2 of the
licensee's submittal dated August 31, 2009.
Date of issuance: August 27, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 278 and 281.
Renewed Facility Operating License Nos. DPR-44 and DPR-56:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: May 5, 2010 (75 FR
23815).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 27, 2010.
No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
No. 1, Rockingham County, New Hampshire
Date of amendment request: March 16, 2010, as supplemented on July
9, 2010.
Description of amendment request: This amendment revises the
Seabrook Technical Specifications requirement that the Operations
Manager shall have held a senior reactor operator license for the
Seabrook Station prior to assuming the Operations Manager position.
Specifically, the proposed change now requires the Operations Manager
to meet one of the following: (1) Hold a senior operator license; (2)
have held a senior operator license for a similar unit; or (3) have
been certified for equivalent senior operator knowledge.
Date of issuance: September 2, 2010.
Effective date: As of its date of issuance and shall be implemented
within 30 days.
Amendment No.: 124.
Facility Operating License No. NPF-86: The amendment revised the TS
and the License.
Date of initial notice in Federal Register: May 4, 2010 (75 FR
23816).
The supplemental letter dated July 9, 2010, 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 September 2, 2010.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey
Date of application for amendment: March 29, 2010, as supplemented
on June 25, and August 18, 2010.
Brief description of amendments: The amendment revises the
Technical Specifications (TSs) to allow a one-time replacement of the
2C 125-volt direct current battery while Salem Unit No. 2 is at power.
Date of issuance: September 1, 2010.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 280.
[[Page 57532]]
Facility Operating License No. DPR-75: The amendment revised the
TSs and the License.
Date of initial notice in Federal Register: June 1, 2010 (75 FR
30446).
The letters dated June 25, and August 18, 2010, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the application beyond
the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 1, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 10th day of September 2010.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2010-23388 Filed 9-20-10; 8:45 am]
BILLING CODE 7590-01-P