Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 52412-52426 [E8-20567]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 175 (Tuesday, September 9, 2008)]
[Notices]
[Pages 52412-52426]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-20567]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 14, 2008, to August 27, 2008. The 
last

[[Page 52413]]

biweekly notice was published on August 26, 2008 (73 FR 50356).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. Requests for a hearing and a 
petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR Part 2. Interested person(s) should consult a current copy of 
10 CFR 2.309, which is available at the Commission's PDR, located at 
One White Flint North, Public File Area 01F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Access and Management System's 
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web 
site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.

[[Page 52414]]

    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM is free and is 
available at https://www.nrc.gov/site-help/e-submittals/install-
viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-
submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First-class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
participants are requested not to include copyrighted materials in 
their submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
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 pdr@nrc.gov.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin

    Date of amendment request: August 14, 2008.
    Description of amendment request: The proposed amendment would 
modify Specification 4.4.f.1, ``Containment Isolation Device 
Verification,'' of the Technical Specifications (TS) to require 
verification that the 36-inch containment purge and vent isolation 
valves are sealed closed when the reactor is at greater than Cold 
Shutdown conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The Design Bases Accidents (DBA) that result in a release of 
radioactive material within containment are a steam line break, 
rupture of a rod cluster control assembly, and loss-of-coolant 
accident (LOCA). In the analyses for each of these accidents, it is 
assumed that containment isolation valves are either closed or 
function to close within the required isolation time following 
accident initiation. This ensures that potential leakage paths to 
the environment

[[Page 52415]]

through containment isolation valves (including containment purge 
and vent isolation valves) are minimized. The safety analyses assume 
that the containment purge and vent isolation valves are closed at 
accident initiation.
    The safety function of the containment purge and vent isolation 
valves is to support the Containment Isolation system by confining 
fission products within the Primary Containment system boundary 
during a DBA. The proposed amendment would require verification that 
the containment purge and vent isolation valves are sealed closed 
when the reactor is at greater than Cold Shutdown conditions. This 
requirement ensures the valves are in their required DBA post-
accident position when the reactor is at greater than Cold Shutdown 
conditions.
    Verifying the containment purge and vent isolation valves are 
sealed closed at 31-day intervals does not add, delete, or modify 
any KPS system, structure, or component (SSC). Verifying that the 
containment purge and vent isolation valves are sealed closed when 
the reactor is at greater than Cold Shutdown conditions has no 
adverse effect on the ability of the plant to mitigate the effects 
of DBAs. The subject surveillance requirement constitutes a 
verification of isolation valve position and has no effect on 
equipment. Verification of valve closure only ensures the previous 
assumptions made in evaluating the consequences of DBAs remain 
valid. Therefore, there is no increase in the probability of an 
accident by performing the surveillance in additional modes of plant 
operation.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Verifying the containment purge and vent isolation valves are 
sealed closed when the reactor is at greater than Cold Shutdown 
conditions at 31-day intervals ensures these valves are in their 
required DBA post-accident position when the design function is 
required. The proposed amendment does not change the manner in which 
these valves are operated when the reactor is at or below Cold 
Shutdown or their design function. The proposed amendment does not 
create any new failure mechanisms or malfunctions for plant 
equipment or the nuclear fuel.
    In addition, the containment purge and vent isolation valves are 
not accident initiators. Their function is only for mitigation of 
accidents.
    Therefore, the proposed amendment 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.
    Verifying the containment purge and vent isolation valves are 
sealed closed when the reactor is at greater than Cold Shutdown 
conditions at 31-day intervals ensures these valves are in their 
required DBA post-accident position when the design function is 
required. The proposed amendment does not change the manner in which 
these valves are operated when the reactor is at or below Cold 
Shutdown condition.
    The proposed amendment would align the KPS TS with applicable 
NRC requirements stated in NUREG-0800 [``Standard Review Plan,''], 
Section 6.2.4 and NUREG-0737 [``Clarification of Three Mile Island 
Action Plan Requirements,''], Item II.E.4.2. The proposed amendment 
does not result in altering or exceeding a design basis or safety 
limit for the plant. The safety analysis of record, including 
evaluations of the radiological consequences of design basis 
accidents, will remain applicable and unchanged.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Branch Chief: Lois James.

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 amendment request: August 1, 2008.
    Description of amendment request: The proposed amendments would 
authorize changes to the Updated Final Safety Analysis Report (UFSAR) 
to account for small areas of carbon steel (CS) and low alloy steel 
that may be exposed to the reactor coolant system (RCS).
    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?
    No.
    The Pressurizer vent nozzle and thermowell, as components of the 
RCS, must maintain system pressure boundary. RCS design pressure is 
2500 psig and design temperature is 670 [deg]F. The vent nozzle and 
thermowell replacements are designed for the RCS pressure and 
temperature. As described above, the material of the new Pressurizer 
vent nozzle and thermowell is an improvement in the PWSCC [primary 
water stress corrosion cracking] resistance of those components as 
compared to the original components. The design of the new 
Pressurizer vent nozzle and thermowell exposes small areas of the 
Pressurizer shell carbon steel to a stagnant reactor coolant 
environment. However, the corrosion of the Pressurizer shell is 
considered negligible. Therefore, the replacement of the Pressurizer 
vent nozzle and thermowell do not more than minimally increase the 
likelihood of occurrence of a malfunction. Corrosion evaluations 
performed show that all applicable ASME Code requirements are met.
    It is concluded that the consequences of a Pressurizer vent 
nozzle or Pressurizer thermowell failure resulting in a LOCA [loss-
of-coolant accident] are bounded by existing analysis. Therefore, 
there is no increase in the probability or consequences of an 
accident.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No.
    The only credible accident involving the failure of these 
components is bounded by existing LOCA analyses. There are no new 
accidents that need to be postulated due to the replacement of the 
Pressurizer vent nozzle and Pressurizer thermowell. Therefore, this 
proposed activity will not create the possibility of a new or 
different kind of accident from any kind of accident previously 
evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No.
    The mitigation technique selected for the Pressurizer vent 
nozzle and the Pressurizer thermowell exposes a small area of CS to 
the RCS environment. As required by the ASME Code, Section III, a 
supporting corrosion evaluation was developed within each of the two 
component designs. The technical package for the replacement of the 
Pressurizer vent nozzle and the Pressurizer thermowell utilized 
calculations to support the evaluation of the acceptability of this 
repair/replacement activity. The corrosion evaluation for the 
Pressurizer vent resulted in a conservative general stagnant 
corrosion rate of 0.0018 inches per year and the corrosion 
evaluation for the Pressurizer thermowell resulted in a conservative 
general corrosion rate of 0.00142 inches per year. The critical 
corrosion distance is the radius from the exposed CS surface to the 
edge of the weld pad. This distance is at least 1.1 inches for both 
the vent and thermowell designs. With this distance, a corrosion 
rate of less than 2 mils per year is not significant when compared 
to the 60 year component design life, which begins at the time of 
installation.
    The original Pressurizer was designed to meet Section III of the 
ASME Code, and the Pressurizer, as modified, meets Section III of 
the ASME code. Although this change does expose small areas of CS in 
the Pressurizer, the 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

[[Page 52416]]

review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Melanie C. Wong.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New 
York

    Date of amendment request: July 8, 2008.
    Description of amendment request: The proposed amendment to Indian 
Point Nuclear Generating Units Nos. 2 and 3 (IP2 and IP3) would require 
the licensee to submit information and analyses associated with 
extending the Reactor Vessel (RV) Inservice Inspection (ISI) Interval 
from 10 to 20 years for specific pressure retaining welds in the RV.
    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 will revise the license to 
require the submission of information and analyses to the NRC 
following completion of each ASME [American Society of Mechanical 
Engineers] [C]ode, Section XI, Category B-A and B-D Reactor Vessel 
weld inspection. The extension of the ISI from 10 to 20 years is 
being evaluated as part of the relief request independent from the 
license change. Submission of the information and analyses can have 
no effect on the consequences of an accident or the probability of 
an accident because the submission of information is not related to 
the operation of the plant or any equipment, the programs and 
procedures used to operate the plant, or the evaluation of 
accidents. The submittal of information and analyses provides the 
opportunity for the NRC to independently assess the information and 
analyses.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change will only affect the 
requirement to submit information and analyses when specified 
inspections are performed. There are no changes to plant equipment, 
operating characteristics or conditions, programs, and procedures or 
training. Therefore, there are no potential new system interactions 
or failures that could create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed change will revise the license to 
require the submission of information and analyses to the NRC 
following completion of each ASME [C]ode, Section XI, Category B-A 
and B-D Reactor Vessel weld inspection which does not affect any 
Limiting Conditions for Operation used to establish the margin of 
safety. The requirement to submit information and analyses is an 
administrative tool to assure the NRC has the ability to 
independently review information developed by the [l]icensee. The 
proposed change does not involve a significant reduction in [a] 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: July 9, 2008.
    Description of amendment request: The proposed amendment will 
revise the test acceptance criteria specified in the Technical 
Specification Surveillance Requirement (SR) 3.8.1.10 for the Diesel 
Generator (DG) endurance test. The load ranges and power factors 
specified for the test will be changed for consistency with the 
associated safety analyses.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change revises the acceptance criteria to be 
applied to an existing surveillance test of the facility emergency 
diesel generators (DGs). Performing a surveillance test is not an 
accident initiator and does not increase the probability of an 
accident occurring. The proposed new acceptance criteria will assure 
that the DGs are capable of carrying the peak electrical loading 
assumed in the various existing safety analyses which take credit 
for the operation of the DGs. Establishing acceptance criteria that 
bound existing analyses validates the related assumption used in 
those analyses regarding the capability of equipment to mitigate 
accident conditions. 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?
    No. The proposed change revises the test acceptance criteria for 
a specific performance test conducted on the existing DGs. The 
proposed change does not involve installation of new equipment or 
modification of existing equipment, so no new equipment failure 
modes are introduced. The proposed revision to the DG surveillance 
test acceptance criteria also is not a change to the way that the 
equipment or facility is operated and no new accident initiators are 
created. Therefore the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The conduct of performance tests on safety-related plant 
equipment is a means of assuring that the equipment is capable of 
maintaining the margin of safety established in the safety analyses 
for the facility. The proposed change in the DG technical 
specification surveillance test acceptance criteria is consistent 
with values assumed in existing safety analyses is consistent with 
the design rating of the DGs. Therefore the propose change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: May 5, 2008.
    Description of amendment request: The proposed amendment would 
correct an error in Section A.1 of the renewed operating license and 
remove several outdated license conditions relating to surveillance 
requirements. Specifically, it would remove the words ``filed by 
Entergy Nuclear Palisades, LLC (ENP) and Entergy Nuclear

[[Page 52417]]

Operations, Inc. (ENO)'' in Section A.1, spell-out acronyms used in 
Section 1.F, and delete license conditions 2.C.(4) and 2.C.(5), and 
delete Table 2.C.(5).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment deletes incorrect or outdated 
information from the renewed facility operating license. The 
proposed amendment does not involve operation of the required 
structures, systems or components (SSCs) in a manner or 
configuration different from those previously recognized or 
evaluated.
    Modification of renewed facility operating license sections 1.A 
and 1.F and deletion of license conditions 2.C.(4), 2.C.(5), and 
Table 2.C.(5) is administrative and has no impact on plant operation 
or equipment.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously 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 license amendment does not involve a physical 
alteration of any SSC or change the way any SSC is operated. The 
proposed license amendment does not involve operation of any 
required SSCs in a manner or configuration different from those 
previously recognized or evaluated.
    Modification of renewed facility operating license sections 1.A 
and 1.17 and deletion of license conditions 2.C.(4), 2.C.(5), and 
Table 2.C.(5) is administrative and has no impact on plant operation 
or equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Modification of renewed facility operating license sections 1.A 
and 1.F and deletion of license conditions 2.C.(4), 2.C.(5), and 
Table 2.C.(5) is administrative and has no impact on plant operation 
or equipment or on any margin of safety.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Lois M. James.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: May 5, 2008
    Description of amendment request: The proposed amendment would 
revise renewed facility operating license DPR-20 to remove license 
condition 2F. This license condition describes reporting requirements 
for exceeding the facility steady-state reactor core power level 
described in condition 2.C.(1). The proposed change is consistent with 
the Nuclear Regulatory Commission (NRC)-approved change notice 
published in the Federal Register on November 4, 2005, announcing the 
availability of this improvement through the consolidated line item 
improvement process. The Federal Register Notice included a model 
safety evaluation and model no significant hazards consideration (NSHC) 
determination, relating to the elimination of the license condition 
involving reporting of violations of other requirements (typically in 
License Conditions 2.C) in the operating license of some commercial 
nuclear power plants. The licensee affirmed the applicability of the 
model NSHC determination in its application dated May 5, 2008.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and therefore does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Lois M. James.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood; Station, Units 1 and 2, Will County, Illinois; Docket Nos. 
STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle 
County, Illinois

    Date of amendment request: July 29, 2008.
    Description of amendment request: The proposed amendments would 
remove time, cycle, or modification-related items from the operating 
licenses (OLs) and technical specifications (TSs) at both stations. 
Additionally, the proposed amendments would correct typographical 
errors introduced into the TSs at both stations in previous amendments. 
The time, cycle, or modification-related items have been implemented or 
superseded, are no longer applicable, and no longer need to be 
maintained in their associated OLs or TSs.
    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 initial conditions and methodologies used in the accident 
analyses remain unchanged. The proposed changes do not change or 
alter the design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, accident 
analyses results are not impacted.
    All changes proposed by EGC in this amendment request are 
administrative in nature, and are removing one-time requirements 
that have been satisfied or items that are no longer applicable. 
There are no physical changes to the facilities, nor any changes to 
the station operating procedures,

[[Page 52418]]

limiting conditions for operation, or limiting safety system 
settings.
    Based on the above discussion, the proposed changes do 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.
    None of the proposed changes affect the design or operation of 
any system, structure, or component in the plant. The safety 
functions of the related structures, systems, or components are not 
changed in any manner, nor is the reliability of any structure, 
system, or component reduced by the revised surveillance or testing 
requirements. The changes do not affect the manner by which the 
facility is operated and do not change any facility design feature, 
structure, system, or component. No new or different type of 
equipment will be installed. Since there is no change to the 
facility or operating procedures, and the safety functions and 
reliability of structures, systems, or components are not affected, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Based on this evaluation, 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to the Facility Operating Licenses and TS 
are administrative in nature and have no impact on the margin of 
safety of any of the TS. There is no impact on safety limits or 
limiting safety system settings. The changes do not affect any plant 
safety parameters or setpoints. The Operating License Conditions 
have been satisfied as required. There are no changes to the 
conditions themselves.
    Based on this evaluation, the proposed change does not involve a 
significant reduction in a margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

Florida Power and Light Company, et al. , Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: January 23, 2008.
    Description of amendment request: Replace the current Technical 
Specification pressure/temperature (P/T) limit curves with new P/T 
limit curves applicable to 55 effective full-power years (EFPY). The 
low-temperature overpressure protection (LTOP) requirements, which are 
based on the P/T limits, will also be applicable to 55 EFPY.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes have been determined in accordance with the 
methodologies set forth in the regulations to provide an adequate 
margin of safety to ensure that the reactor vessel will withstand 
the effects of normal startup and shutdown cyclic loads due to 
system temperature and pressure changes as well as the loads 
associated with reactor trips. The regulations of 10 CFR Part 50 
Appendix A, Design Criterion 14 and Design Criterion 31 remains 
satisfied. The pressure-temperature (P/T) limit curves in the 
Technical Specifications are conservatively generated in accordance 
with the fracture toughness requirements of the ASME [American 
Society of Mechanical Engineers] Code Section XI, Appendix G. The 
margins of safety against fracture provided by the P/T limits using 
the requirements of 10 CFR 50 Appendix G are equivalent to those 
recommended in ASME Section XI, Appendix G. The Adjusted Reference 
Temperature (ART) values are based on the guidance of RG [Regulatory 
Guide] 1.99 [Reference 4].
    The proposed changes will not result in physical changes to 
structures, systems or components SSCs or to event initiators or 
precursors. Changing the heatup and cooldown curves and the pressure 
relief setpoints to reflect 55 EFPY does not affect the ability to 
control the RCS [reactor coolant system] at low temperatures such 
that the integrity of the reactor coolant pressure boundary would 
not be compromised by violating the P/T limits.
    The proposed changes will not impact assumptions and conditions 
previously used in the radiological consequence evaluations nor 
affect mitigation of these consequences due to an accident described 
in the UFSAR [Updated Final Safety Analysis Report]. Also, the 
proposed changes will not impact a plant system such that previously 
analyzed SSCs might be more likely to fail. The initiating 
conditions and assumptions for accidents described in the UFSAR 
remain as analyzed.
    Thus, based on the above, reasonable assurance is provided that 
the proposed amendment does not significantly increase the 
probability or consequences of accidents previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The requirements for P/T limit curves and LTOP have been in 
place since the beginning of plant operation. The revised curves are 
based on a later edition of Section XI of the ASME Code that 
incorporates current industry standards for P/T curves. The revised 
curves also are based on reactor vessel irradiation damage 
predictions using RG 1.99 methodology. No new failure modes are 
identified nor are any SSCs required to be operated outside of their 
design bases. Consequently, the proposed amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed P/T curves continue to maintain the safety margins 
of 10 CFR 50 Appendix G by defining the limits of operation which 
prevent nonductile failure of the reactor pressure vessel. Analyses 
have demonstrated that the fracture toughness requirements are 
satisfied and that conservative operating restrictions are 
maintained for the purpose of low temperature overpressure 
protection. The P/T limit curves provide assurance that the RCS 
pressure boundary will behave in a ductile manner and that the 
probability of a rapidly propagating fracture is minimized. 
Therefore, operation in accordance with the proposed amendment would 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Thomas H. Boyce.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant (MNGP), Wright County, Minnesota

    Date of amendment request: April 4, 2008.
    Description of amendment request: The licensee proposed to change 
the Technical Specifications (TS) to revise requirements for 
unavailable barriers by adding Limiting Condition for Operation (LCO) 
3.0.9. This LCO would establish conditions under which systems would 
remain operable when required physical barriers are not capable of 
providing their related support function. This proposed amendment is 
consistent with the NRC's approved Technical Specification Task

[[Page 52419]]

Force (TSTF) Improved Standard Technical Specifications Change 
Traveler, TSTF-427, Revision 2. A notice of availability of this TS 
improvement was published in the Federal Register on October 3, 2006 
(71 FR 58444) as part of NRC's Consolidated Line Item Improvement 
Process (CLIIP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided an 
analysis of the issue of no significant hazards consideration by citing 
the proposed NSHC determination published by the NRC staff in the 
Federal Register referenced above. That proposed NSHC is reproduced 
below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier if risk is assessed and managed. 
The postulated initiating events which may require a functional 
barrier are limited to those with low frequencies of occurrence, and 
the overall TS system safety function would still be available for 
the majority of anticipated challenges. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on the 
allowance provided by proposed LCO 3.0.9 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.9. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change. The addition of a 
requirement to assess and manage the risk introduced by this change 
will further minimize possible concerns. Therefore, this change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable barrier, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed. The postulated initiating 
events which may require a functional barrier are limited to those 
with low frequencies of occurrence, and the overall TS system safety 
function would still be available for the majority of anticipated 
challenges. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in RG [Regulatory 
Guide] 1.177. A bounding risk assessment was performed to justify 
the proposed TS changes. This application of LCO 3.0.9 is predicated 
upon the licensee's performance of a risk assessment and the 
management of plant risk. The net change to the margin of safety is 
insignificant as indicated by the anticipated low levels of 
associated risk (ICCDP [incremental conditional core damage 
probability] and ICLERP [incremental conditional large early release 
probability] ) as shown in Table 1 of Section 3.1.1 in the Safety 
Evaluation. Therefore, this change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the analysis cited by the licensee, and 
has found that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the proposed 
amendment involves no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: Lois M. James.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant (MNGP), Wright County, Minnesota

    Date of amendment request: April 22, 2008.
    Description of amendment request: The licensee proposed to change 
the Technical Specifications (TS) to (1) revise the surveillance 
requirement frequency in Specification 3.1.3, ``Control Rod 
Operability,'' to require control rod notch testing to be performed at 
a 31-day frequency for both partially and fully withdrawn control rods; 
and (2) revise Example 1.4-3 in Section 1.4, ``Frequency,'' to clarify 
the applicability of the 1.25 surveillance test interval extension. 
These proposed changes are consistent with the NRC's approved Technical 
Specification Task Force (TSTF) Improved Standard Technical 
Specifications (STS) Change Traveler, TSTF-475, Revision 1. A notice of 
availability of this TS improvement was published in the Federal 
Register on November 13, 2007 (72 FR 63935), as part of the NRC's 
Consolidated Line Item Improvement Process (CLIIP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 FR 50.91(a), the licensee provided an 
analysis of the issue of no significant hazards consideration by citing 
the proposed NSHC determination published by the NRC staff in the 
Federal Register notice referenced above. That proposed NSHC is 
reproduced below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM [Source Range 
Monitoring] Insert Control Rod Action.'' TSTF-475, Revision 1, 
modifies NUREG-1433 (BWR [Boiling Water Reactor]/4) and NUREG-1434 
(BWR/6) STS. The changes (1) revise TS testing frequency for 
surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod 
OPERABILITY,'', and (2) revise Example 1.4-3 in Section 1.4 
``Frequency'' to clarify the applicability of the 1.25 surveillance 
test interval extension. The consequences of an accident after 
adopting TSTF-475, Revision 1 are no different than the consequences 
of an accident prior to adoption. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Accident Previously 
Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change will not introduce new failure modes or effects and 
will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously analyzed. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    TSTF-475, Revision 1 will: (1) Revise the TS SR 3.1.3.2 
frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' and (2) revise 
Example 1.4-3 in Section 1.4 ``Frequency'' to clarify the 
applicability of the 1.25 surveillance test interval extension. The 
GE Nuclear Energy Report, ``CRD Notching Surveillance Testing for 
Limerick Generating Station,'' dated November 2006, concludes that 
extending the control rod notch test interval from weekly to monthly 
is not expected to impact the reliability of the scram system and 
that the analysis supports the decision to change the surveillance 
frequency. Therefore, the proposed changes in TSTF-475, Revision 1

[[Page 52420]]

are acceptable and do not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the analysis cited by the licensee, and 
has found that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the proposed 
amendment involves no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: Lois M. James.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: June 26, 2008.
    Description of amendment request: The proposed amendments would 
amend the Facility Operating Licenses by revising the licensing basis 
loss of coolant accident and main steam line break accident 
radiological dose consequences for Prairie Island Nuclear Generating 
Plant, Units 1 and 2, as currently described in the Updated Safety 
Analysis Report Section 14.5 and Section 14.9. This proposed amendment 
also proposes concomitant amendments to Appendix A of the Facility 
Operating Licenses, Technical Specifications (TS) 3.3.5, ``Containment 
Ventilation Isolation Instrumentation'', 3.4.17, ``RCS [Reactor Coolant 
System] Specific Activity'', and 3.6.3, ``Containment Isolation 
Valves'', which are necessary to implement the proposed revised 
analyses.
    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.
    This license amendment request proposes implementing revised 
loss of coolant accident and main steam line break accident dose 
consequence analyses to address modeling nonconservatisms and update 
the analyses for new fuel types and provide margin for power 
uncertainty. These analyses assumed that the containment inservice 
purge system penetrations are isolated, thus this license amendment 
request proposes Technical Specification revisions which will 
require these penetrations to be blind flanged during plant 
operations; these changes allow the Technical Specification 
requirements for containment ventilation isolation instrumentation 
to be removed. This license amendment request also proposes 
associated more restrictive limits in the Technical Specification 
for reactor coolant system specific activity since the main steam 
line break accident analysis assumed lower limits.
    The accident radiological dose consequences analyses inputs, 
methodologies and outputs modified by this request are not accident 
initiators and do not affect the frequency of occurrence of 
previously analyzed transients. Likewise, the reactor coolant system 
specific activity limits are not accident initiators and do not 
affect the frequency of occurrence of previously analyzed 
transients.
    The containment inservice purge system is not an accident 
initiator and therefore removal of its Technical Specifications does 
not involve an increase in the probability of an accident. The 
Technical Specification changes proposed in this license amendment 
request require the containment inservice purge system to be blind 
flanged during Modes 1, 2, 3, and 4, therefore removal of the 
containment ventilation isolation instrumentation Technical 
Specifications and other Technical Specification system operating 
requirements does not involve an increase in the consequences of an 
accident previously evaluated.
    The loss of coolant accident and main steam line break accident 
radiological dose consequences analyses demonstrated the results are 
within the applicable regulatory limits and guidance using revised 
inputs, including the proposed lower Technical Specification reactor 
coolant system specific activity limits, and methodologies. Thus 
these changes do not involve a significant increase in the 
consequences of an accident.
    Therefore, the proposed changes do 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.
    This license amendment request proposes implementing revised 
loss of coolant accident and main steam line break accident dose 
consequence analyses to address modeling nonconservatisms and update 
the analyses for new fuel types and provide margin for power 
uncertainty. These analyses assumed that the containment inservice 
purge system penetrations are isolated, thus this license amendment 
request proposes Technical Specification revisions which will 
require these penetrations to be blind flanged during plant 
operations; these changes allow the Technical Specification 
requirements for containment ventilation isolation instrumentation 
to be removed. This license amendment request also proposes 
associated more restrictive limits in the Technical Specification 
for reactor coolant system specific activity since the main steam 
line break accident analysis assumed lower limits.
    This license amendment request does not involve physical changes 
to the plant structures, systems or components and there is no 
adverse impact on component or system interactions due to the 
proposed changes. The modes of operation of the plant remain 
unchanged and the design functions of the safety systems remain in 
compliance with the applicable safety analysis acceptance criteria. 
These changes do not create new failure modes or mechanisms and no 
new accident precursors are generated.
    When the containment inservice purge system is not being 
operated, current Technical Specifications require the system's 
penetrations to be blind flanged in Modes 1, 2, 3, and 4 to provide 
post-accident containment integrity. This license amendment proposes 
to require the system penetrations to be blind flanged at all times 
during these Modes and prevent operation of the system in these 
Modes. Since containment integrity is provided with the penetrations 
blind flanged and this change only extends the time during which the 
system is in this configuration, these changes do not create the 
possibility of a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?