Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 57521-57532 [2010-23388]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 182 (Tuesday, September 21, 2010)]
[Notices]
[Pages 57521-57532]
From the Federal Register Online via the Government Printing 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, https://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 
hearing.docket@nrc.gov, 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 https://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 
https://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 https://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 https://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 https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, 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 
https://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, https://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 pdr.resource@nrc.gov.
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, L