Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 25034-25050 [E8-9679]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 88 (Tuesday, May 6, 2008)]
[Notices]
[Pages 25034-25050]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-9679]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 10 to April 23, 2008. The last 
biweekly notice was published on April 22, 2008 (73 FR 21567).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received

[[Page 25035]]

within 30 days after the date of publication of this notice will be 
considered in making any final determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. Requests for a hearing and a 
petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR Part 2. Interested person(s) should consult a current copy of 
10 CFR 2.309, which is available at the Commission's PDR, located at 
One White Flint North, Public File Area 01F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Access and Management System's 
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web 
site, 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.
    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

[[Page 25036]]

representative) already holds an NRC-issued digital ID certificate). 
Each petitioner/requestor will need to download the Workplace Forms 
ViewerTM to access the Electronic Information Exchange 
(EIE), a component of the E-Filing system. The Workplace Forms 
ViewerTM is free and is available at 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.

Arizona Public Service Company (APS), et al., Docket Nos. STN 50-528, 
STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of amendment request: January 17, 2008, as supplemented 
February 29, 2008.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications (TS) to establish more effective 
and appropriate action, surveillance, and administrative requirements 
related to ensuring the habitability of the control room envelope (CRE) 
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task 
Force (TSTF) Standard Technical Specification change traveler TSTF-448, 
Revision 3, ``Control Room Habitability.'' Specifically, the proposed 
amendments would modify TS 3.7.11, ``Control Room Essential Filtration 
System (CREFS),'' and add new TS 5.5.17, ``Control Room Envelope 
Habitability Program,'' to TS Administrative Controls Section 5.5, 
``Programs and Manuals.''
    The NRC staff issued a ``Notice of Availability of Technical 
Specification Improvement to Modify Requirements Regarding Control Room 
Envelope Habitability Using the Consolidated Line Item Improvement 
Process,'' associated with TSTF-448, Revision 3, in the Federal 
Register on January 17, 2007 (72 FR 2022). The notice included a model 
safety evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request. In its 
application dated January 17, 2008, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:
Criterion 1--The Proposed Change[s] [Do] Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change[s] [do] not adversely affect accident 
initiators or precursors nor alter the design assumptions, conditions, 
or configuration of the facility. The proposed change[s] [do] not alter 
or prevent the ability of structures, systems, and components (SSCs) to 
perform their intended function to mitigate the consequences of an 
initiating event within the assumed

[[Page 25037]]

acceptance limits. The proposed change[s] [revise] the TS for the CRE 
[essential filtration] system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE [essential 
filtration] system is the CRE boundary. The CRE [essential filtration] 
system is not an initiator or precursor to any accident previously 
evaluated. Therefore, the probability of any accident previously 
evaluated is not increased. Performing tests to verify the operability 
of the CRE boundary and implementing a program to assess and maintain 
CRE habitability ensure that the CRE [essential filtration] system is 
capable of adequately mitigating radiological consequences to CRE 
occupants during accident conditions, and that the CRE [essential 
filtration] system will perform as assumed in the consequence analyses 
of design basis accidents. Thus, the consequences of any accident 
previously evaluated are not increased. Therefore, the proposed 
change[s] [do] not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change[s] [Do] Not Create the Possibility of 
a New or Different Kind of Accident From any Accident Previously 
Evaluated
    The proposed change[s] [do] not impact the accident analysis. The 
proposed change[s] [do] not alter the required mitigation capability of 
the CRE [essential filtration] system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new surveillance 
or following the new program. The proposed change[s] [do] not involve a 
physical alteration of the plant (i.e., no new or different type of 
equipment will be installed) or a significant change in the methods 
governing normal plant operation. The proposed change[s] [do] not alter 
any safety analysis assumptions and is consistent with current plant 
operating practice. Therefore, [the] change[s] [do] not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
Criterion 3--The Proposed Change[s] [Do] Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change[s] [do] not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change[s] [do] not affect safety 
analysis acceptance criteria. The proposed change[s] will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The proposed 
change[s] [do] not adversely affect systems that respond to safely shut 
down the plant and to maintain the plant in a safe shutdown condition. 
Therefore, the proposed change[s] [do] not involve a significant 
reduction in a margin of safety. Based upon the reasoning presented 
above and the previous discussion of the amendment request, the 
requested change does not involve a no-significant-hazards 
consideration.
    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on that review, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the request for amendments involves NSHC.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: January 22, 2008.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) 3.8.3 requirements related to 
Diesel Fuel Oil, Lube Oil, and Starting Air by replacing the specific 
fuel oil and lube oil storage values with the corresponding number of 
days supply. The specific volumes would be relocated to a licensee-
controlled document (i.e., the TS Bases). It would also expand the 
``clear and bright'' test in TS 5.5.10 by allowing a water and sediment 
test to be performed to establish the acceptability of new fuel oil 
prior to addition to the storage tanks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the Diesel Fuel Oil, Lube Oil, and 
Starting Air Specification relocates the volume of diesel fuel oil 
and lube oil required to support 7 day operation of the onsite 
diesel generators, and the volume equivalent to a 6 day supply, to 
licensee control. The specific volume of fuel oil equivalent to a 7 
and 6 day supply is calculated using the NRC approved methodology 
described in Regulatory Guide 1.137, Revision 1, ``Fuel Oil Systems 
for Standby Diesel Generators'' and ANSI/ANS [American National 
Standards Institute/American Nuclear Society] 59.51-1997 (formerly 
ANSI N195-1976), ``Fuel Oil Systems for Safety-Related Emergency 
Diesel Generators.'' The specific volume of lube oil equivalent to a 
7 and 6 day supply is based on the Emergency Diesel Generator (EDG) 
manufacturer's consumption values for the run time of the EDG. 
Because the requirements to maintain a 7 day supply of diesel fuel 
oil and lube oil are not changed and are consistent with the 
assumptions in the accident analyses, and the actions taken when the 
volume of fuel oil and lube oil are less than a 6 day supply have 
not changed, neither the probability nor the consequences of any 
accident previously evaluated will be affected. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to the Diesel Fuel Oil Testing Program adds 
an option to use already approved testing methodology. Since the 
methodology is already discussed in ASTM D975 [``Standard 
Specification for Diesel Fuel Oils''] as an acceptable standard to 
determine water and sediment content, neither the probability nor 
the consequences of any accident previously evaluated will be 
affected. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the Diesel Fuel Oil, Lube Oil and 
Starting Air Specification and Diesel Fuel Oil Testing Program do 
not involve physical alterations of the plant (i.e., no new or 
different type of equipment will be installed) or changes in the 
methods governing normal plant operation. The changes do not alter 
assumptions made in the safety analysis but ensure that the diesel 
generator operates as assumed in the accident analysis. The proposed 
changes are consistent with the safety analysis assumptions. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the Diesel Fuel Oil, Lube Oil, and 
Starting Air Specification relocates the volume of diesel fuel oil 
and lube oil required to support 7 day operation of the onsite 
diesel generators, and the volume equivalent to a 6 day supply, to

[[Page 25038]]

licensee control. As the bases for the existing limits on diesel 
fuel oil and lube oil are not changed and the methods used to 
determine these limits have been previously approved, no change is 
made to the accident analysis assumptions and no margin of safety is 
reduced as part of this change. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.
    The proposed change to the Diesel Fuel Oil Testing Program 
provides an option to use a quantitative method of testing for 
sediment and water content as an alternative to a qualitative 
method. This option uses an already accepted method for assessing 
fuel oil quality. Based on this, there are no alterations to any 
assumptions used in the accident analysis and this change does not 
reduce any margin of safety. Therefore, the proposed change does not 
involve a significant reduction in the margin of safety.

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

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York

    Date of amendment request: February 7, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) Surveillance Requirement (SR) 
3.1.3.2 frequency in TS 3.1.3, ``Control Rod OPERABILITY'' from ``7 
days after the control rod is withdrawn and THERMAL POWER is greater 
than the [Low Power Setpoint] LPSP of [Rod Worth Minimizer] RWM'' to 
``31 days after the control rod is withdrawn and THERMAL POWER is 
greater than the LPSP of the RWM'' and revise Example 1.4-3 in Section 
1.4 ``Frequency'' to clarify the applicability of the 1.25 surveillance 
test interval extension. The proposed amendment does not adopt the 
clarification of Source Range Monitor (SRM) TS action for inserting 
control rods. This clarification was previously adopted during the 
JAFNPP conversion to Improved Standard Technical Specifications, TS 
Section 3.3.1.2, required Action E.2, ``Source Range Monitoring [SRM] 
Instrumentation.''
    Date of publication of individual notice in Federal Register: April 
2, 2008 (73 FR 18008).
    Expiration date of individual notice: May 2, 2008.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: March 13, 2008.
    Description of amendment request: The licensee proposes to change 
the Surveillance Requirement (SR) 3.6.5.8 to require verification that 
the reactor building spray nozzles are unobstructed following 
maintenance that could result in nozzle blockage in lieu of the current 
SR of performing the test every 10 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Reactor Building Spray System is not an initiator of any 
analyzed event. The proposed change does not have a detrimental 
impact on the integrity of any plan structure, system, or component 
that may initiate an analyzed event. The proposed change will not 
alter the operation or otherwise increase the failure probability of 
any plant equipment that can initiate an analyzed accident. This 
change does not affect the plant design. There is no increase in the 
likelihood of formation of significant corrosion products. Due to 
their location at the top of the containment, introduction of 
foreign material into the spray headers is unlikely. Foreign 
materials exclusion controls during and following maintenance 
provides assurance that the nozzles remain unobstructed. 
Consequently, there is no significant increase in the probability of 
an accident previously evaluated.
    The Reactor Building Spray system is designed to address the 
consequences of a Loss of Coolant Accident (LOCA) or a Main 
Steamline Break (MSLB) inside the reactor building. The Reactor 
Building Spray system is capable of performing its function 
effectively with the single failure of any active component in the 
system, any of its subsystems, or any of its support systems.
    Therefore, the consequences of an accident previously evaluated 
are not significantly affected by the proposed change.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will not physically alter the plant (no new 
or different type of equipment will be installed) or change the 
methods governing normal plant operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The system piping and nozzles are made if material that is not 
susceptible to corrosion. Obstruction from sources external to the 
system is highly unlikely due to the location high in the reactor 
building and not being readily accessible. Strict controls are 
established to ensure the foreign material is not introduced into 
the Reactor Building Spray system during maintenance or repairs. 
Maintenance activities that could introduce significant foreign 
material into the system require subsequent system cleanliness 
verification which would prevent nozzle blockage. The spray header 
nozzles are expected to remain unblocked and available in the event 
that the safety function is required. The capacity of the system 
would remain unaffected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Entergy Nuclear Operations, P.O. Box 31995, Jackson, 
Mississippi 39286-1995.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: March 13, 2008.
    Description of amendment request: The proposed changes would 
replace the current Technical Specification (TS) 3.4.12, ``RCS [Reactor 
Coolant System] Specific Activity'' limit on reactor coolant system 
(RCS) gross specific activity with a new limit on RCS noble gas 
specific activity. The noble gas specific activity limit would be based 
on a new dose equivalent Xe-133 (DEX) definition that would replace the 
current E Bar average disintegration energy definition. In addition, 
the current dose equivalent I-131 (DEI) definition would be revised to 
allow the use of additional thyroid dose conversion factors (DCFs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or

[[Page 25039]]

consequences of an accident previously evaluated?
    Response: No.
    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary 
coolant gross activity is not within limit is not an initiator for 
any accident previously evaluated. The current variable limit on 
primary coolant iodine concentration is not an initiator to any 
accident previously evaluated. As a result, the proposed change does 
not significantly increase the probability of an accident. The 
proposed change will limit primary coolant noble gases to 
concentrations consistent with the accident analyses. The proposed 
change to the Completion Time has no impact on the consequences of 
any design basis accident since the consequences of an accident 
during the extended Completion Time are the same as the consequences 
of an accident during the current Completion Time. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change in specific activity limits does not alter 
any physical part of the plant nor does it affect any plant 
operating parameter. The change does not create the potential for a 
new or different kind of accident from any previously calculated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the limits on noble gas 
radioactivity in the primary coolant. The proposed change is 
consistent with the assumptions in the safety analyses and will 
ensure the monitored values protect the initial assumptions in the 
safety analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Entergy Nuclear Operations, P.O. Box 31995, Jackson, 
Mississippi 39286-1995.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: March 13, 2008.
    Description of amendment request: The proposed changes would 
replace the current TS 3.4.8, ``Reactor Coolant System Specific 
Activity'' limit on reactor coolant system (RCS) gross specific 
activity with a new limit on RCS noble gas specific activity. The noble 
gas specific activity limit would be based on a new dose equivalent Xe-
133 (DEX) definition that would replace the current E Bar average 
disintegration energy definition. In addition, the current dose 
equivalent I-131 (DEI) definition would be revised to allow the use of 
additional thyroid dose conversion factors (DCFs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary 
coolant gross activity is not within limit is not an initiator for 
any accident previously evaluated. The current variable limit on 
primary coolant iodine concentration is not an initiator to any 
accident previously evaluated. As a result, the proposed change does 
not significantly increase the probability of an accident. The 
proposed change will limit primary coolant noble gases to 
concentrations consistent with the accident analyses. The proposed 
change to the Completion Time has no impact on the consequences of 
any design basis accident since the consequences of an accident 
during the extended Completion Time are the same as the consequences 
of an accident during the current Completion Time. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change in specific activity limits does not alter 
any physical part of the plant nor does it affect any plant 
operating parameter. The change does not create the potential for a 
new or different kind of accident from any previously calculated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the limits on noble gas 
radioactivity in the primary coolant. The proposed change is 
consistent with the assumptions in the safety analyses and will 
ensure the monitored values protect the initial assumptions in the 
safety analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Entergy Nuclear Operations, P. O. Box 31995, Jackson, 
Mississippi 39286-1995.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: March 13, 2008.
    Description of amendment request: The proposed change will relocate 
Technical Specification (TS) 3.4.7, ``Reactor Coolant System 
Chemistry,'' to the Technical Requirements Manual (TRM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change acts to relocate current Reactor Coolant 
System (RCS) chemistry limits and monitoring requirements from the 
TSs to the TRM. Monitoring and maintaining RCS chemistry minimizes 
the potential for corrosion of RCS piping and components. Corrosion 
effects are considered a long-term impact on RCS structural 
integrity. Because RCS chemistry will continue to be monitored and 
controlled, relocating the current TS requirements to the TRM will 
not present an adverse impact to the RCS and, subsequently, will not 
impact the probability or consequences of an accident previously 
evaluated. Furthermore, once relocated to the TRM, changes to RCS 
chemistry limits or monitoring requirements will be controlled in 
accordance with 10 CFR 50.59.
    Therefore, the proposed change does not involve a significant 
increase in the

[[Page 25040]]

probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not result in any plant modifications 
or changes in the way the plant is operated. The proposed change 
only acts to relocate current RCS chemistry limits and monitoring 
requirements from the TSs to the TRM.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will maintain limits on RCS chemistry 
parameters and will continue to provide associated monitoring 
requirements. Once relocated to the TRM, changes to RCS chemistry 
limits or monitoring requirements will be controlled in accordance 
with 10 CFR 50.59. In addition, the RCS chemistry limits are not a 
structure, system, or component which operating experience or 
probabilistic risk assessment has shown to be significant to public 
health and safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Entergy Nuclear Operations, P.O. Box 31995, Jackson, 
Mississippi 39286-1995.
    NRC Branch Chief: Thomas G. Hiltz.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: December 12, 2007.
    Description of amendment request: The proposed changes are 
administrative in nature and provide editorial changes to the technical 
specifications (TSs). The proposed changes involve: (1) Correcting the 
index; (2) removing cycle specific requirements or notes that have 
since expired and are no longer applicable; (3) deleting references to 
previously deleted requirements; (4) changing references to the 
location of previously relocated information; and (5) other editorial 
corrections. These proposed changes correct minor inconsistencies that 
have been introduced over time as a result of previous changes to the 
TSs or involve changes that are solely editorial in nature.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature and do not 
impact the physical configuration or function of plant structures, 
systems, or components (SSCs) or the manner in which SSCs are 
operated, maintained, modified, tested, or inspected. The proposed 
changes do not impact the initiators or assumptions of analyzed 
events, nor do they impact mitigation of accidents or transient 
events.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature and do not 
alter plant configuration, require that new plant equipment be 
installed, alter assumptions made about accidents previously 
evaluated, or impact the function of plant SSCs or the manner in 
which SSCs are operated, maintained, modified, tested, or inspected. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are administrative in nature and do not 
involve any physical changes to plant SSCs or the manner in which 
SSCs are operated, maintained, modified, tested, or inspected. The 
proposed changes do not involve a change to any safety limits, 
limiting safety system settings, limiting conditions of operation, 
or design parameters for any SSC. The proposed changes do not impact 
any safety analysis assumptions and do not involve a change in 
initial conditions, system response times, or other parameters 
affecting an accident analysis. Therefore, the proposed changes do 
not involve a significant reduction in a margin of safety.

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

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS),

    Units 2 and 3, York and Lancaster Counties, Pennsylvania 
    Date of amendment request: July 13, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications to support application of 
Alternative Source Term (AST) methodology at PBAPS Units 2 and 3. The 
fission product release from the reactor core into containment is 
referred to as the ``source term,'' and is characterized by the 
composition and magnitude of the radioactive material, the chemical and 
physical properties of the material, and the timing of the release from 
the reactor core as discussed in Technical Information Document (TID) 
14844, ``Calculation of Distance Factors for Power and Test Reactor 
Sites.'' Since the publication of TID 14844, advances have been made in 
understanding the composition and magnitude, chemical form, and timing 
of fission product releases from severe nuclear power plant accidents. 
In light of these insights, NUREG-1465, ``Accident Source Terms for 
Light-Water Nuclear Power Plants,'' was published in 1995 with revised 
ASTs for use in the licensing of future light-water reactors.
    The Nuclear Regulatory Commission (NRC), in Title 10 of the Code of 
Federal Regulations, Section 50.67 (10 CFR 50.67), ``Accident source 
term,'' subsequently allowed the use of the ASTs described in NUREG-
1465 at operating plants. This request to apply the AST methodology is 
made in accordance with 10 CFR 50.67, with the exception that TID 14844 
will continue to be used as the radiation dose basis for equipment 
qualification at PBAPS Units 2 and 3. Application of the AST 
methodology at PBAPS Units 2 and 3 requires that radiation dose limits 
specified in 10 CFR 50.67 are adhered to for the exclusion area 
boundary, the low population zone outer boundary, and the facility 
control room.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 25041]]

consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The implementation of alternative source term (AST) assumptions 
has been evaluated in revisions to the analyses of the following 
limiting design basis accidents (DBAs) at Peach Bottom Atomic Power 
Station (PBAPS):
     Loss-of-Coolant Accident,
     Fuel Handling Accident,
     Control Rod Drop Accident, and
     Main Steam Line Break Accident.
    Based upon the results of these analyses, it has been 
demonstrated that, with the requested changes, the dose consequences 
of these limiting events are within the regulatory guidance provided 
by the NRC for use with the AST. This guidance is presented in 10 
CFR 50.67 and associated Regulatory Guide 1.183, and Standard Review 
Plan Section 15.0.1. The Alternative Source Term is an input to 
calculations used to evaluate the consequences of an accident, and 
does not by itself affect the plant response, or the actual pathway 
of the radiation released from the fuel. It does, however, better 
represent the physical characteristics of the release, so that 
appropriate mitigation techniques may be applied. Therefore, the 
consequences of an accident previously evaluated are not 
significantly increased.
    The equipment affected by the proposed changes is mitigative in 
nature, and relied upon after an accident has been initiated. 
Application of the Alternative Source Term (AST) does not involve 
any physical changes to the plant design. While the operation of 
various systems do change as a result of these proposed changes, 
these systems are not accident initiators. Application of the AST is 
not an initiator of a design basis accident. The proposed changes to 
the Technical Specifications (TS), while they revise certain 
performance requirements, do not involve any physical modifications 
to the plant. As a result, the proposed changes do not affect any of 
the parameters or conditions that could contribute to the initiation 
of any accidents. As such, removal of operability requirements 
during the specified conditions will not significantly increase the 
probability of occurrence for an accident previously analyzed. Since 
design basis accident initiators are not being altered by adoption 
of the Alternative Source Term analyses, the probability of an 
accident previously evaluated is not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed 
and there are no physical modifications to existing equipment 
associated with the proposed changes). Similarly, it does not 
physically change any structures, systems or components involved in 
the mitigation of any accidents; thus, no new initiators or 
precursors of a new or different kind of accident are created. New 
equipment or personnel failure modes that might initiate a new type 
of accident are not created as a result of the proposed amendment.
    As such, the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Safety margins and analytical conservatisms have been evaluated 
and have been found acceptable. The analyzed events have been 
carefully selected and margin has been retained to ensure that the 
analyses adequately bound postulated event scenarios. The dose 
consequences due to design basis accidents comply with the 
requirements of 10 CFR 50.67 and the guidance of Regulatory Guide 
1.183. The proposed amendment is associated with the implementation 
of a new licensing basis for PBAPS Design Basis Accidents (DBAs). 
Approval of the change from the original source term to a new source 
term taken from Regulatory Guide 1.183 is being requested. The 
results of the accident analyses, revised in support of the proposed 
license amendment, are subject to revised acceptance criteria. The 
analyses have been performed using conservative methodologies, as 
specified in Regulatory Guide 1.183. Safety margins have been 
evaluated and analytical conservatism has been utilized to ensure 
that the analyses adequately bound the postulated limiting event 
scenario. The dose consequences of these DBAs remain within the 
acceptance criteria presented in 10 CFR 50.67, ``Accident Source 
Term'', and Regulatory Guide 1.183.
    The proposed changes continue to ensure that the doses at the 
exclusion area boundary (EAB) and low population zone boundary 
(LPZ), as well as the Control Room, are within corresponding 
regulatory limits.
    Therefore, operation of PBAPS in accordance with the proposed 
changes will not involve a significant reduction in a margin of 
safety.

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

FPL Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: March 31, 2008.
    Description of amendment request: FPL Energy Point Beach, LLC, 
requests adoption of an approved change to the Standard Technical 
Specifications (STS) for pressurized-water reactor (PWR) plants (NUREG-
1430, NUREG-1431, & NUREG-1432) and plant-specific technical 
specifications (TS), to replace the current limits on primary coolant 
gross specific activity with limits on primary coolant noble gas 
activity. The noble gas activity would be based on dose equivalent 
Xenon-133 and would take into account only the noble gas activity in 
the primary coolant. In addition, the current dose equivalent I-131 
definition would be revised to allow the use of additional thyroid dose 
conversion factors. The changes are consistent with Nuclear Regulatory 
Commission (NRC)-approved Industry/Technical Specification Task Force 
(TSTF) Standard Technical Specification Change Traveler, TSTF-490, 
Revision 0.
    Basis for proposed no-significant-hazards-consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary coolant 
gross activity is not within limit is not an initiator for any accident 
previously evaluated. The current variable limit on primary coolant 
iodine concentration is not an initiator to any accident previously 
evaluated. As a result, the proposed change does not significantly 
increase the probability of an accident. The proposed change will limit 
primary coolant noble gases to concentrations consistent with the 
accident analyses. The proposed change to the Completion Time has no 
impact on the consequences of any design basis accident since the 
consequences of an accident during the extended Completion Time are the 
same as the consequences of an accident during the Completion Time. As 
a result, the consequences of any accident previously evaluated are not 
significantly increased.
    Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Accident Previously 
Evaluated.
    The proposed change in specific activity limits does not alter any 
physical part of the plant nor does it affect any plant operating 
parameter.

[[Page 25042]]

The change does not create the potential for a new or different kind of 
accident from any previously calculated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The proposed change revises the limits on noble gas radioactivity 
in the primary coolant. The proposed change is consistent with the 
assumptions in the safety analyses and will ensure the monitored values 
protect the initial assumptions in the safety analyses. Based upon the 
reasoning presented above, the requested change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the analysis and based on this review, 
it appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Antonio Fernandez, Esquire, Senior Attorney, 
FPL Energy Point Beach, LLC, P. O. Box 14000, Juno Beach, FL 33408-
0420.
    NRC Branch Chief: Lois M. James.

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

    Date of amendment request: March 31, 2008.
    Description of amendment request: The licensee proposed to increase 
the current maximum power level authorized by Section 2.C(1) of the 
renewed facility operating license from 1,775 megawatts thermal (Mwt) 
to 1,870 Mwt, an approximately five percent increase from the current 
licensed thermal power. The current maximum power level of 1,775 Mwt 
was approved in 1998, an increase of 6.3 percent from the original 
licensed thermal power of 1670 Mwt. Thus, when approved, the licensee's 
proposed amendment would take the maximum power level to about 12 
percent above the original license thermal power. The licensee's 
application addresses in details each of the following major technical 
areas: Extended power uprate, containment analysis methods change, 
increase in credit for containment overpressure for low head emergency 
core cooling system (ECCS) pumps, and reactor internal pressure 
differentials (RIPDs) for the steam dryer.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC). 
The licensee's NSHC analysis, addressing each technical area listed 
above, is reproduced below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

Extended Power Uprate

    Response: No.
    The probability (frequency of occurrence) of [d]esign [b]asis 
[a]ccidents occurring is not affected by the increased power level, 
because Monticello Nuclear Generating Plant (MNGP) continues to 
comply with the regulatory and design basis criteria established for 
plant equipment. A probabilistic risk assessment demonstrates that 
the calculated core damage frequencies do not significantly change 
due to [e]xtended [p]ower [u]prate (EPU). Scram setpoints (equipment 
settings that initiate automatic plant shutdowns) are established 
such that there is no significant increase in scram frequency due to 
EPU. No new challenges to safety-related equipment result from EPU.
    The changes in consequences of postulated accidents, which would 
occur from 102 percent of the EPU [rated thermal power] RTP compared 
to those previously evaluated, are acceptable. The results of EPU 
accident evaluations do not exceed the NRC[-] approved acceptance 
limits. The spectrum of postulated accidents and transients has been 
investigated, and are shown to meet the plant's currently licensed 
regulatory criteria. In the area of fuel and core design, for 
example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and 
other applicable Specified Acceptable Fuel Design Limits (SAFDL) are 
still met. Continued compliance with the SLMCPR and other SAFDLs 
will be confirmed on a cycle[-]specific basis consistent with the 
criteria accepted by the NRC.
    Challenges to the [r]eactor [c]oolant [p]ressure [b]oundary were 
evaluated at EPU conditions (pressure, temperature, flow, and 
radiation) and were found to meet their acceptance criteria for 
allowable stresses and overpressure margin. Challenges to the 
containment have been evaluated, and the containment and its 
associated cooling systems continue to meet the current licensing 
basis. The increase in the calculated post[-] LOCA suppression pool 
temperature above the currently assumed peak temperature was 
evaluated and determined to be acceptable. Radiological release 
events (accidents) have been evaluated, and have been shown to meet 
the guidelines of 10 CFR 50.67.

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks, variable RHR [residual heat 
removal] heat exchanger capability K-value, and mechanistic heat and 
mass transfer from the suppression pool surface to the wetwell 
airspace after 30 seconds for the long[-]term design[-] basis [-
]accident loss of coolant accident (DBA-LOCA) containment analysis 
are not relevant to accident initiation, but rather, pertain to the 
method used to accurately evaluate postulated accidents. The use of 
these elements does not, in any way, alter existing fission product 
boundaries, and provides a conservative prediction of the 
containment response to DBA-LOCAs. Therefore, the containment 
analysis method change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Increase in Credit for Containment Overpressure for Low Head 
Emergency Core Cooling System (ECCS) Pumps

    Response: No.
    These changes update parameters used in the MNGP safety analyses 
and expand the range and scope of the analyses. This will result in 
a more realistic analysis of available containment overpressure 
under design [-]basis accident conditions. The updated analyses 
affect only the evaluation of previously reviewed accidents. No 
plant structure, system, or component (SSC) is physically affected 
by the updated and expanded analyses. No method of operation of any 
plant SSC is affected. Therefore, there is no significant increase 
in the probability or consequence of a previously evaluated 
accident.

Reactor Internal Pressure Differentials (RIPDs) for the Steam Dryer

    Response: No.
    The revised steam dryer RIPDs are used in evaluating loads in 
reactor vessel internals for various conditions (i.e., during 
normal, upset and faulted conditions). The values more accurately 
represent the actual plant configuration. No plant structure, 
system, or component (SSC) is physically affected by the updated and 
expanded analyses. No method of operation of any plant SSC is 
affected. Therefore, there is no significant increase in the 
probability or consequence of a previously evaluated accident.
    The analyses supporting the above evaluations were performed at 
the EPU power level of 2,004 Mwt, which bounds this license 
amendment request to operate at 1,870 Mwt. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the pro