Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 4764-4781 [E9-1568]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 16 (Tuesday, January 27, 2009)]
[Notices]
[Pages 4764-4781]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-1568]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0016]


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 December 31, 2008 to January 13, 2009. The 
last biweekly notice was published on January 13, 2009 (74 FR 1712).

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.

[[Page 4765]]

    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, TWB-05-B01M, Division of Administrative 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. 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.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Public File 
Area 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.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve all adjudicatory documents 
over the Internet or in some cases to mail copies on electronic storage 
media. Participants may not submit paper copies of their filings unless 
they seek 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 Viewer \TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ 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

[[Page 4766]]

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 electronic filing 
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, 
Monday through Friday. The help electronic filing Help Desk can be 
contacted by telephone at 1-866-672-7640 or by e-mail at 
MSHD.Resource@nrc.gov.
    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).
    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.resource@nrc.gov.
Arizona Public Service Company, 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: July 2, 2008.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 4.2.2, ``Control Element Assemblies,'' to 
support replacement of the full strength control element assemblies 
(CEAs) with a new design beginning with the 14th refueling outage 
(U3R14) for Palo Verde Nuclear Generating Station (PVNGS), Unit 3 in 
the spring of 2009. Additionally, Arizona Public Service Company (APS) 
will be updating the TS by removing the registered trademark 
``Inconel'' while retaining the generic terminology ``Alloy 625'' and 
deleting the references to part-length CEAs in TS 4.2.2.
    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?
    Replacement of full-strength compression sleeve control element 
assemblies with full-strength silver (Ag)-indium (In)-Cadmium (Cd) 
control element assemblies.
    Response: No.
    The proposed change involves a new design for the full-strength 
Control Element Assemblies (CEA) that replaces a portion of B4C 
pellets (including the compression sleeve) in the tips of the CEA 
fingers with hollow silver-indium-cadmium slugs.
    The following events are related to inadvertent movement of the 
CEAs; however, they are not initiated by the CEAs.
     Uncontrolled Control Element Assembly Withdrawal from a 
Subcritical or Low (Hot Zero) Power Condition.
     Uncontrolled Control Element Assembly Withdrawal at 
Power.
     Single Full-Strength Control Element Assembly Drop.
     Control Element Assembly Ejection.
    These previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components (SSC) other than the CEA 
itself. The proposed change to the CEA design does not have a 
detrimental impact on the integrity of any plant SSC that initiates 
an analyzed event. Additionally, the CEAs mitigate other events. In 
these events, the chrome plating on the portion of the clad exterior 
and the added weight has been conservatively accounted for in the 
SCRAM [safety control rod axe man] calculation. The change does not 
adversely affect the protective and mitigative capabilities of the 
plant, nor does the change affect the initiation or probability of 
occurrence of any accident. The SSCs will continue to perform their 
intended safety functions.
    The proposed change in CEA design has resulted in a slight (less 
than 1%) reduction of total reactivity.
    Computer modeling events which exhibit sensitivity to time 
dependent rod worth (sheared shaft/seized rotor, loss of flow from 
SAFDL [specified acceptable fuel design limits] and total loss of 
reactor coolant flow) demonstrate that all acceptance criteria 
continued to be met.
    Therefore this change will not significantly increase the 
probability or consequences of any accident previously evaluated.
    The removal of the registered trademark name ``Inconel''.
    Response: No.
    This change is considered editorial. Inconel is a registered 
trademark of Special Metals Corporation, while Alloy 625 is a 
generic alloy designation from the Unified Numbering System. 
Retaining the already referenced term ``Alloy 625'' does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated, as the material properties and 
application of Alloy 625 have not changed.
    Deletion of the references to part-length control element 
assemblies.
    Response: No.
    This change is considered editorial. The removal of this 
information does not involve a significant increase in the 
probability or consequences of an accident previously evaluated as 
the part-length CEAs were

[[Page 4767]]

replaced in accordance with License Amendment 152, dated March 23, 
2004 (Agency Document Access and Management System (ADAMS) Accession 
No. ML040860573) and the information is no longer applicable.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Replacement of full-strength compression sleeve control element 
assemblies with full-strength silver(Ag)-indium(In)-Cadmium(Cd) 
control element assemblies.
    Response: No.
    There are three differences in the replacement CEAs as compared 
to the current CEAs.
    First, there is a very slight change in the outside diameter of 
a portion of the cladding on the replacement CEAs due to chrome 
plating on the lower portion of cladding. Analysis demonstrates that 
this change will not cause interference between the CEA cladding and 
the guide tube inside diameter in the buffer region. Secondly, there 
is a slight increase in weight with the Ag-In-Cd CEAs. However, this 
difference has been analyzed with respect to the performance 
capability of the CEDMs [Control Element Drive Mechanisms] and found 
to be within design capabilities and design analyses. Finally, the 
upper edges of the spider bosses have been chamfered to prevent 
damage to the self-latching mechanisms that can occur if the CEA 
hangs up when lifting through the upper guide structure cut outs. 
This change is for ease of maintenance and has no impact on 
operation of the CEAs.
    Therefore, the Ag-In-Cd CEAs are identical to the compression 
sleeve CEAs in terms of form, fit and function and the proposed 
change will not introduce any new failure mechanisms, malfunctions, 
or accident initiators not already considered in the design and 
licensing bases. The possibility of a new or different malfunction 
of safety-related equipment is not created. No new accident 
scenarios, transient precursors, or limiting single failures are 
introduced as a result of these changes. There will be no adverse 
effects or challenges imposed on any safety-related system as a 
result of these changes. Therefore, the possibility of a new or 
different accident from any accident previously evaluated is not 
created as a result of any dimensional change.
    The removal of the registered trademark name ``Inconel''.
    Response: No.
    This change is considered editorial. Inconel is a registered 
trademark of Special Metals Corporation, while Alloy 625 is a 
generic alloy designation from the Unified Numbering System. 
Retaining the already referenced term ``Alloy 625'' does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, as the material properties and 
application of Alloy 625 have not changed.
    Deletion of the references to part-length control element 
assemblies.
    Response: No.
    This change is considered editorial. The removal of this 
information does not create the possibility of a new or different 
kind of accident from any accident previously evaluated as the part-
length CEAs were replaced in accordance with License Amendment 152, 
dated March 23, 2004 (Agency Document Access and Management System 
(ADAMS) Accession No. ML040860573) and the information is no longer 
applicable.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Replacement of full-strength compression sleeve control element 
assemblies with full-strength silver(Ac)-indium(In)-Cadmium(Cd) 
control element assemblies.
    Response: No.
    Reactor core safety limits are established in the PVNGS 
Technical Specifications to prevent overheating of the fuel and 
cladding that would result in the release of fission products to the 
reactor coolant during steady state operation, normal operational 
transients, and anticipated operational occurrences. The margin to 
these safety limits is not affected by the CEA design changes under 
consideration.
    Overheating of the fuel is prevented by maintaining steady 
state, peak linear heat rate (LHR) below the level at which fuel 
centerline melting occurs. If the local LHR is high enough to cause 
the fuel centerline temperature to reach the melting point of the 
fuel, expansion of the pellet caused by centerline melting may cause 
the pellet to stress the cladding to the point of failure, allowing 
an uncontrolled release of activity to the reactor coolant.
    Compliance with the DNBR [departure from nucleate boiling ratio] 
and fuel centerline melt specified acceptable fuel design limits 
(SAFDLs) is assured through the CEA insertion limits and alignment 
technical specifications, and through the power distribution limit 
technical specifications.
    There is no change to the operation of the full-strength CEAs 
due to the change from compression sleeve CEAs to Ag-In-Cd CEAs. 
Since the Ag-In-Cd CEAs may be used to control power distribution 
similar to the compression sleeve CEAs, power distributions will 
still be controlled and maintained within the limits necessary to 
assure SAFDLs are met.
    The proposed change in CEA design has resulted in a slight (less 
than 1%) reduction in total reactivity.
    Computer modeling results of events which exhibit sensitivity to 
time dependent rod worth (sheared shaft/seized rotor, loss of flow 
from SAFDL and total loss of reactor coolant flow) demonstrate that 
all acceptance criteria continued to be met.
    Therefore, since SAFDLs continue to be met, the change from 
compression sleeve CEAs to Ag-In-Cd CEAs does not involve a 
significant reduction in a margin of safety.
    The removal of the registered trademark name ``Inconel''.
    Response: No.
    The removal of the registered trademark name ``Inconel'' [ ] is 
considered editorial. Inconel is a registered trademark of Special 
Metals Corporation, while Alloy 625 is a generic alloy designation 
from the Unified Numbering System. Retaining the already referenced 
term ``Alloy 625'' does not involve a significant reduction in the 
margin of safety as the material properties and application of Alloy 
625 have not changed.
    Deletion of the references to part-length control element 
assemblies.
    Response: No.
    This change is considered editorial. The removal of this 
information does not involve a significant reduction in the margin 
of safety as the part-length CEAs were replaced in accordance with 
Amendment 152, dated March 23, 2004 (Agency Document Access and 
Management System (ADAMS) Accession No. ML040860573) and the 
information is no longer applicable.

    The NRC staff has reviewed the licensee's analysis 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 no significant hazards consideration.
    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: Michael T. Markley.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina
    Date of amendments request: October 6, 2008.
    Description of amendments request: The proposed change would remove 
work hour controls and/or references to the NRC Generic Letter 82-12 
from the administrative control sections of the technical 
specifications. On April 17, 2007, the NRC approved a final rule that 
amended 10 CFR Part 26 and, among other changes, established 
requirements for managing worker fatigue at operating nuclear power 
plants. Subpart I, ``Managing Fatigue,'' specifically addresses 
managing worker fatigue by designating individual break requirements, 
work hour limits, and annual reporting requirements. Subpart I was 
published in the Federal Register on March 31, 2008 (73 FR 16966), with 
a required implementation period of 18 months. Compliance is, 
therefore, required by October 1, 2009. In order to support compliance 
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove 
these work hour controls from Technical Specification 5.2.2.e at the 
Brunswick Steam Electric Plant, Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 4768]]

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 remove TS [technical specification] 
controls on working hours for personnel who perform safety related 
functions. The TS controls are superseded by the worker fatigue 
requirements in 10 CFR Part 26. Removal of the TS requirements will 
be performed concurrently with the implementation of the 10 CFR Part 
26, Subpart I requirements. The proposed changes 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 the mitigation of accidents or transient events.
    Therefore, it is concluded that these 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 remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. 
Work hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not restrict work hours and thereby create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter plant configuration, require 
that new plant equipment be installed, alter assumptions made about 
accidents previously evaluated, add any initiators, or effect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected.
    Therefore, it is concluded that this change does 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 remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. The 
proposed changes do not involve any physical changes to plant or the 
manner in which plant systems are operated, maintained, modified, 
tested, or inspected. The proposed changes do not alter the manner 
in which safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this change. The proposed 
changes will not result in plant operation in a configuration 
outside the design basis. The proposed changes will not adversely 
affect systems that respond to safely shutdown the plant and to 
maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed. 
Therefore, it is concluded that these 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina
    Date of amendment request: October 6, 2008.
    Description of amendments request: The proposed change would remove 
work hour controls and/or references to the NRC Generic Letter 82-12 
from the administrative control sections of the technical 
specifications. On April 17, 2007, the NRC approved a final rule that 
amended 10 CFR Part 26 and, among other changes, established 
requirements for managing worker fatigue at operating nuclear power 
plants. Subpart I, ``Managing Fatigue,'' specifically addresses 
managing worker fatigue by designating individual break requirements, 
work hour limits, and annual reporting requirements. Subpart I was 
published in the Federal Register on March 31, 2008 (73 FR 16966), with 
a required implementation period of 18 months. Compliance is, 
therefore, required by October 1, 2009. In order to support compliance 
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove 
these work hour controls from Technical Specification 5.2.2.e at the H. 
B. Robinson Steam Electric Plant, Unit 2.
    Basis for proposed no significant hazards consideration 
determination 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 remove TS [technical specification] 
controls on working hours for personnel who perform safety related 
functions. The TS controls are superseded by the worker fatigue 
requirements in 10 CFR Part 26. Removal of the TS requirements will 
be performed concurrently with the implementation of the 10 CFR Part 
26, Subpart I requirements. The proposed changes 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 the mitigation of accidents or transient events.
    Therefore, it is concluded that these 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 remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. 
Work hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not restrict work hours and thereby create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter plant configuration, require 
that new plant equipment be installed, alter assumptions made about 
accidents previously evaluated, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected.
    Therefore, it is concluded that this change does 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 remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. The 
proposed changes do not involve any physical changes to the plant or 
the manner in which plant systems are operated,

[[Page 4769]]

maintained, modified, tested, or inspected. The proposed changes do 
not alter the manner in which safety limits, limiting safety system 
settings or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed changes will not result in plant operation in a 
configuration outside the design basis. The proposed changes will 
not adversely affect systems that respond to safely shut down the 
plant and to maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed. 
Therefore, it is concluded that these 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina
    Date of amendment request: October 6, 2008.
    Description of amendment request: The proposed change would remove 
work hour controls and/or references to the NRC Generic Letter 82-12 
from the administrative control sections of the technical 
specifications. On April 17, 2007, the NRC approved a final rule that 
amended 10 CFR Part 26 and, among other changes, established 
requirements for managing worker fatigue at operating nuclear power 
plants. Subpart I, ``Managing Fatigue,'' specifically addresses 
managing worker fatigue by designating individual break requirements, 
work hour limits, and annual reporting requirements. Subpart I was 
published in the Federal Register on March 31, 2008 (73 FR 16966), with 
a required implementation period of 18 months. Compliance is, 
therefore, required by October 1, 2009. In order to support compliance 
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove 
these work hour controls from Technical Specification 6.2.2.f at the 
Shearon Harris Nuclear Power Plant, Unit 1.
    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 remove TS [technical specification] 
controls on working hours for personnel who perform safety related 
functions. The TS controls are superseded by the worker fatigue 
requirements in 10 CFR Part 26. Removal of the TS requirements will 
be performed concurrently with the implementation of the 10 CFR Part 
26, Subpart I requirements. The proposed changes 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 the mitigation of accidents or transient events.
    Therefore, it is concluded that these 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 remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. 
Work hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not restrict work hours and thereby create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter plant configuration, require 
that new plant equipment be installed, alter assumptions made about 
accidents previously evaluated, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected.
    Therefore, it is concluded that this change does 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 remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. The 
proposed changes do not involve any physical changes to the plant or 
the manner in which plant systems are operated, maintained, 
modified, tested, or inspected. The proposed changes do not alter 
the manner in which safety limits, limiting safety system settings 
or limiting conditions for operation are determined. The safety 
analysis acceptance criteria are not affected by this change. The 
proposed changes will not result in plant operation in a 
configuration outside the design basis. The proposed changes will 
not adversely affect systems that respond to safely shut down the 
plant and to maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed. 
Therefore, it is concluded that these 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas
    Date of amendment request: November 13, 2008.
    Description of amendment request: The proposed change will modify 
Technical Specification (TS) 3.3.1.1, ``Reactor Protective 
Instrumentation.'' Specifically, Table 4.3-1 and the associated Notes 7 
and 8 will be revised to clarify and streamline the reactor coolant 
system (RCS) flow verification requirements associated with the 
departure from nucleate boiling ratio (DNBR) reactor trip signal.
    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 CPC [Core Protection Calculator] reactor protective function 
is not considered an accident initiator. The primary function is to 
initiate an automatic reactor trip signal when specific plant 
conditions are reached, thereby limiting the consequences of an 
accident. The proposed change acts to eliminate unnecessary 
conservatisms and accordingly increase operational margin by 
eliminating the requirement to use

[[Page 4770]]

calorimetric flow measurement in the CPC flow verification. This 
method of verification will normally only be used in the future 
during periods when the COLSS [Core Operating Limits Supervisory 
System] RCP [Reactor Coolant Pump] [Delta] p flow measurement is 
unavailable. Regardless of the method of verification used, the CPC 
will continue to be verified to have an indicated RCS flow equal to 
or conservative relative to the measured RCS flow on a once per 12-
hour basis. In so doing, the CPC will continue to act to generate a 
reactor trip on low DNBR as originally designed in order to ensure 
the DNBR reactor core Safety Limit is not exceeded.
    The relocation of measurement uncertainty references to the TS 
Bases does not reduce the requirements to account for uncertainties 
in any Limiting Safety System Setting (LSSS) designed to protect 
reactor core Safety Limits. The necessary uncertainties will 
continue to be applied as required and will be controlled in 
accordance with TS 6.5.14, Technical Specification Bases Control 
Program, and station procedures.
    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 does not result in any physical plant 
modifications or changes in the way the plant is operated. In 
addition, the CPCs are unrelated to any type of accident initiator 
previously evaluated.
    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 increases operating margin when the COLSS 
RCP [Delta]p flow measurement is available for use while unaffecting 
the CPC ability to initiate an automatic reactor trip on low DNBR 
prior to the DNBR reactor core safety limit being exceeded. 
Relocating the references to measurement uncertainties to the TS 
Bases likewise has no impact on the CPC design function and the 
uncertainties will continue to be applied as required and controlled 
in accordance with TS 6.5.14, Technical Specification Bases Control 
Program, and station procedures.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana
    Date of amendment request: December 8, 2008
    Description of amendment request: The proposed amendment adds a 
license condition to allow a one-time extension of surveillance 
requirements involving the 18-month channel calibration and logic 
system functional tests for one channel of the reactor water level 
instrumentation system. The extension is to account for the effects of 
rescheduling the next refueling outage from early to late 2009.
    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 requested action is a one-time extension to the performance 
interval of certain TS [Technical Specification] surveillance 
requirements. The performance of the surveillances, or the failure 
to perform the surveillances, is not a precursor to an accident. 
Performing the surveillances or failing to perform the surveillances 
does not affect the probability of an accident. Therefore, the 
proposed delay in performance of the surveillance requirements in 
this amendment request does not increase the probability of an 
accident previously evaluated.
    A delay in performing the surveillances does not result in a 
system being unable to perform its required function. Additionally, 
the defense in depth of the system design provides additional 
confidence that the safety function is maintained. In the case of 
this one-time extension request, the relatively short period of 
additional time that the systems and components will be in service 
before the next performance of the surveillance will not affect the 
ability of those systems to operate as designed. Therefore, the 
systems required to mitigate accidents will remain capable of 
performing their required function. No new failure modes have been 
introduced because of this action and the consequences remain 
consistent with previously evaluated accidents. Therefore, the 
proposed delay in performance of the surveillance requirement in 
this amendment request does not involve a significant increase in 
the consequences of an accident.
    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 amendment does not involve a physical alteration of 
any system, structure, or component (SSC), or a change in the way 
any SSC is operated. The surveillance intervals of the level 
instrumentation are currently evaluated for 30 months, which bounds 
the requested interval extension. The proposed amendment does not 
involve operation of any SSCs in a manner or configuration different 
from those previously recognized or evaluated. No new failure 
mechanisms will be introduced by the one-time surveillance extension 
being requested.
    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 amendment is a one-time extension of the 
performance-interval of certain TS surveillance requirements. 
Extending the surveillance requirements does not involve a 
modification of any TS Limiting Conditions for Operation. Extending 
the surveillance frequency does not involve a change to any limit on 
accident consequences specified in the license or regulations. 
Extending the surveillance frequency does not involve a change to 
how accidents are mitigated or a significant increase in the 
consequences of an accident. Extending the surveillance frequency 
does not involve a change in a methodology used to evaluate 
consequences of an accident. Extending the surveillance frequency 
does not involve a change in any operating procedure or process. The 
surveillance intervals of the level instrumentation are currently 
evaluated for 30 months which bounds the requested interval 
extension. The components involved in this request have exhibited 
reliable operation based on the results of the most recent 
performances of their 18-month surveillance requirements and the 
associated functional surveillances.
    Based on the limited additional period of time that the systems 
and components will be in service before the surveillance is next 
performed, as well as the operating experience that these 
surveillances are typically successful when performed, it is 
reasonable to conclude that the margin of safety associated with the 
surveillance requirement will not be affected by the requested 
extension.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 4771]]

    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.
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: December 4, 2008.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16, 
``RCS Specific Activity,'' and Surveillance Requirements 3.4.16.1 and 
3.4.16.3. The proposed changes would replace the current TS 3.4.16 
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 
definition that would replace the current E Bar average disintegration 
energy definition. In addition, the current dose equivalent I-131 
definition would be reformatted. The availability of this TS revision 
was announced in the Federal Register on March 15, 2007 (72 FR 12217) 
as part of the consolidated line item improvement process. The licensee 
affirmed the applicability of the model no significant hazards 
consideration determination in its application.
    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 adopted by the licensee 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. 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.
    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
analysis adopted by the licensee 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 amendments involve 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.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois
    Date of amendment request: September 2, 2008.
    Description of amendment request: The proposed amendments would 
relocate Surveillance Requirements (SR) 3.8.3.6 from the technical 
specifications (TSs) to a licensee-controlled document. SR 3.8.3.6 
requires Emergency Diesel Generator fuel oil storage tanks to be 
drained, sediment removed, and cleaned on a 10-year interval. The 
change is consistent with the current revision (i.e., Rev. 3) of the 
Improved Standard Technical Specifications (ISTS), NUREG 1434, 
``Standard Technical Specifications General Electric Plants, BWR/6.'' 
The SR was removed from the ISTS under Technical Specification Task 
Force (TSTF) Traveler No. 2, ``Relocate the 10-Year Sediment Cleaning 
of the Fuel Oil Storage Tank to Licensee Control,'' approved by the 
Nuclear Regulatory Commission on July 16, 1998.
    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 FOSTs [fuel oil storage tanks] provide the storage for the 
DG [diesel generator] fuel oil, assuring an adequate volume is 
available for each DG to operate for seven days in the event of a 
loss of offsite power concurrent with a loss of coolant accident. 
The relocation of the SR to drain and clean the FOSTs to a licensee-
controlled document will not impact any of the previously analyzed 
accidents. Sediment in the tank, or failure to perform this SR, does 
not necessarily result in an inoperable storage tank. Fuel oil 
quantity and quality are assured by other TS SRs that remain 
unchanged. These SRs help ensure tank sediment is minimized and 
ensure that any degradation of the tank wall surface that results in 
fuel oil volume reduction is detected and corrected in a timely 
manner. Future changes to the licensee-controlled document will be 
evaluated pursuant to the requirements of 10 CFR 50.59, ``Changes, 
tests, and experiments,'' to ensure that such changes do not result 
in more than a minimal increase in the probability or consequences 
of an accident previously evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration or the manner in which the plant is 
operated and maintained. The proposed change does not adversely 
affect the ability of structures, systems or components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits.
    The proposed change does not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed change does not increase the types and amounts 
of radiological effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS change does not involve the addition or 
modification of any plant equipment. Also, the proposed change will

[[Page 4772]]

not alter the design configuration, or method of operation of plant 
equipment beyond its normal functional capabilities. The 
requirements retained in the TS continue to require testing of the 
diesel fuel oil to ensure the proper functioning of the DGs. The 
proposed TS change does not create any new credible failure 
mechanisms, malfunctions or accident initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change does not alter or exceed a design basis or 
safety limit. The requirements retained in the TS continue to 
require testing of the diesel fuel oil to ensure the DGs are able to 
perform their intended function.
    Therefore, the proposed changes 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. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania
    Date of amendment request: September 24, 2008.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TSs) to allow the BVPS-2 containment 
spray additive sodium hydroxide (NaOH) to be replaced by sodium 
tetraborate (NaTB).
    Basis for proposed no significant hazards consideration 
determination: As required by10 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.
    Use of NaTB in lieu of NaOH would not involve a significant 
increase in probability of a previously evaluated accident because 
the containment spray additive is not an initiator of any analyzed 
accident. The NaTB would be stored and delivered by a passive method 
that does not have potential to affect plant operations. Any 
existing NaOH delivery system equipment which remains in place but 
is removed from service would meet existing seismic, electrical and 
containment isolation requirements. Therefore the change in 
additive, including removal of NaOH equipment from service, would 
not result in any failure modes that could initiate an accident.
    The spray additive is used to mitigate the consequences of a 
LOCA [loss-of-coolant accident]. Use of NaTB as an additive in lieu 
of NaOH would not involve a significant increase in the consequences 
of a previously evaluated accident because the amount of NaTB 
specified in the proposed TS would achieve a pH of 7 or greater, 
consistent with the current licensing basis. This pH is sufficient 
to achieve long-term retention of iodine by the containment sump 
fluid for the purpose of reducing accident related radiation dose 
following a LOCA.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Regarding the proposed use of NaTB in lieu of NaOH, the NaTB 
would be stored and delivered by a passive method that does not have 
potential to affect plant operations. Any existing NaOH delivery 
system equipment remaining in place but which is removed from 
service would meet existing seismic, electrical and containment 
isolation requirements. Hydrogen generation would not be 
significantly impacted by the change. Therefore, no new failure 
mechanisms, malfunctions, or