[Federal Register Volume 72, Number 78 (Tuesday, April 24, 2007)]
[Notices]
[Pages 20375-20389]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-7534]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 30, 2007 to April 12, 2007. The last 
biweekly notice was published on April 10, 2007 (72 FR 17944).

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

[[Page 20376]]

determination. Within 60 days after the date of publication of this 
notice, the licensee 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 for a hearing and a petition for leave to intervene.
    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, the 
licensee 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 
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 persons 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, http://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 a hearing or a petition for leave to intervene must 
be filed 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; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, <a href="/cdn-cgi/l/email-protection#a9e1ccc8dbc0c7ceedc6cac2ccdde9c7dbca87cec6df"><span class="__cf_email__" data-cfemail="561e3337243f38311239353d33221638243578313920">[email protected]</span></a>; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and

[[Page 20377]]

petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
<a href="/cdn-cgi/l/email-protection#fab5bdb9b79b9396b99f948e9f88ba948899d49d958c"><span class="__cf_email__" data-cfemail="08474f4b456961644b6d667c6d7a48667a6b266f677e">[email protected]</span></a>. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this 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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#50203422103e22337e373f26"><span class="__cf_email__" data-cfemail="9dedf9efddf3effeb3faf2eb">[email protected]</span></a>.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: March 22, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to incorporate a revised limit for 
the variable low reactor coolant system pressure-temperature core 
protection safety limit. The revised limit is associated with the 
introduction of AREVA NP's Mark-B-HTP fuel design, which will require 
more restrictive Safety Limits and more restrictive Limiting Safety 
System Settings for the Reactor Protection System. The proposed limits 
are developed in accordance with the method described in the Nuclear 
Regulatory Commission (NRC)-approved Topical Report BAW-10179P-A, 
``Safety Criteria and Methodology for Acceptable Cycle Reload 
Analyses.'' The revised limits will maintain the same magnitude of 
departure from nucleate boiling (DNB) protection.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Technical Specification (TS) limits and reactor 
protection system (RPS) trip setpoints are developed in accordance 
with the methods and assumptions described in NRC-approved AREVA NP 
Topical Reports BAW-10179 P-A, ``Safety Criteria and Methodology for 
Acceptable Cycle Reload Analyses'' and BAW-10187 P-A, ``Statistical 
Core Design for B&W-Designed 177 FA Plants.'' The core thermal-
hydraulic code (LYNXT) and CHF correlation (BHTP) have been approved 
for use with these methods and the Mark-B-HTP fuel type. The 
proposed change preserves the design DNB Ratio safety criterion that 
there shall be at least a 95% [percent] probability at a 95% 
confidence level that the hot fuel rod in the core does not 
experience a departure from nucleate boiling during normal operation 
or events of moderate frequency. The corresponding core-wide 
protection on a pin-by-pin basis is greater than 99.9%. The margin 
retained for penalties such as transition core effects, by imposing 
a Thermal Design Limit in all DNB analyses supporting the proposed 
change, has been shown to be sufficient to offset the mixed core 
conditions at TMI Unit 1, where the Mark-B-HTP fuel design will be 
co-resident with earlier Mark-B fuel designs. The setpoint 
calculation methodology utilized, and the surveillance requirements 
established, are in accordance with approved industry standards and 
NRC criteria.
    The proposed setpoint change does not involve a significant 
increase in the consequences of an accident previously evaluated 
because the proposed change does not alter any assumptions 
previously made in the radiological consequence evaluations, or 
affect mitigation of the radiological consequences of an accident 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS limit and reactor protection system (RPS) trip 
setpoint provide a core protection safety limit and variable low 
pressure trip setpoint developed in accordance with NRC-approved 
methods and assumptions. No new accident scenarios, failure 
mechanisms or single failures are introduced as a result of the 
proposed change. All systems, structures, and components previously 
required for the mitigation of an event remain capable of fulfilling 
their intended design function.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed RPS trip setpoint ensures core protection safety 
limits will be preserved during power operation. The proposed safety 
limit and setpoint are developed in accordance with NRC-approved 
methods and assumptions. The margin retained for penalties such as 
transition core effects, by imposing a Thermal Design Limit in all 
DNB analyses supporting the proposed change, has been shown to be 
sufficient to offset the mixed core conditions at TMI Unit 1. The 
setpoint calculation methodology utilized, and the surveillance 
requirements established, are in accordance with approved industry 
standards and NRC criteria.
    Therefore, the proposed changes do not involve a significant 
reduction in any 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. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief: Harold K. Chernoff.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert 
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of amendment request: February 27, 2007.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification 4.2.1, Fuel Assemblies, to add a 
temporary exemption to allow the insertion of up to four lead fuel 
assemblies, which contain non-Zircaloy based cladding, into the Unit 1 
core for one cycle of operation. These lead fuel assemblies are 
currently installed in the Unit 2 core under a previous exemption and 
are scheduled to be discharged during the 2007 refueling outage.
    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. The licensee has determined 
that the proposed change:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.

[[Page 20378]]

    Calvert Cliffs Technical Specification 4.2.1, Fuel Assemblies, 
states that fuel rods are clad with either Zircaloy or 
ZIRLO<SUP>TM</SUP>. Calvert Cliffs Nuclear Power Plant, Inc. 
proposes to re-insert up to four fuel assemblies into Calvert Cliffs 
Unit 1 that have some fuel rods clad in zirconium alloys that do not 
meet the definition of Zircaloy or ZIRLO<SUP>TM</SUP>. A temporary 
exemption to the regulations has been requested to allow these fuel 
assemblies to be re-inserted into Unit 1. The proposed change to the 
Calvert Cliffs Technical Specifications will allow the use of 
cladding materials that are not Zircaloy or ZIRLO<SUP>TM</SUP> for 
one fuel cycle once the temporary exemption is approved. The 
proposed change to the Technical Specification is effective only as 
long as the temporary exemption is effective. The addition of what 
will be an approved temporary exemption for Unit 1 to Technical 
Specification 4.2.1 does not change the probability or consequences 
of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed change does not add any new equipment, modify any 
interfaces with existing equipment, change the equipment's function, 
or change the method of operating the equipment. The proposed change 
does not affect normal plant operations or configuration. Since the 
proposed change does not change the design, configuration, or 
operation, it could not become an accident initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind] of accident from any accident 
previously evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The proposed change will add an approved temporary exemption to 
the Calvert Cliffs Technical Specifications allowing the 
installation of up to four lead fuel assemblies. The assemblies use 
advanced cladding materials that are not specifically permitted by 
existing regulations or Calvert Cliffs' Technical Specifications. A 
temporary exemption to allow the installation of these assemblies 
has been requested. The addition of an approved temporary exemption 
to Technical Specification 4.2.1 is an administrative change to 
allow the installation of the lead fuel assemblies under the 
provisions of the temporary exemption. The license amendment is 
effective only as long as the exemption is effective. This amendment 
does not change the margin of safety since it only adds a reference 
to an approved, temporary exemption to the Technical Specifications.
    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: Carey Fleming, Esquire, Senior Counsel--
Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt 
Street, 17th floor, Baltimore, MD 21202.
    NRC Acting Branch Chief: John P. Boska.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendment request: February 27, 2007.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification 5.6.5, Core Operating Limits 
Report (COLR), to add the supporting topical report (WCAP-15604-NP, 
Revision 2-A, ``Limited Scope High Burnup Lead Test Assemblies,'' 
September 2003) to the list of references. The topical report provides 
guidance for operation with a limited number of lead fuel assemblies to 
be irradiated to a higher burnup limit than currently allowed for 
Calvert Cliffs fuel assemblies.
    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. The licensee has determined 
that the proposed change:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change would modify the Calvert Cliffs Units 1 and 
2 Technical Specification 5.6.5.b, Core Operating Limits Report by 
adding an approved topical report to the existing list of topical 
reports. The topical report provides the technical basis that 
supports irradiating a limited number of lead fuel assemblies to a 
higher burnup limit than currently approved for Calvert Cliffs. The 
proposed change is administrative in nature and has no impact on any 
plant configurations or on system performance that is relied upon to 
mitigate the consequences of an accident.
    In the safety evaluation report approving the requested topical 
report (WCAP-15604-NP, Revision 2-A), the Nuclear Regulatory 
Commission concluded that it is acceptable for an individual power 
licensee to irradiate a limited number of lead fuel assemblies to a 
maximum burnup to 75 GWD/MTU [gigawatt days per metric ton of 
uranium] provided that certain conditions are met. Calvert Cliffs 
meets those required conditions. Because those required conditions 
are met and only a limited number of fuel assemblies are included in 
this change, the probability or consequences of an 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. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed change does not add any new equipment, modify any 
interfaces with existing equipment, change the equipment's function, 
or change the method of operating the equipment. The proposed change 
does not affect normal plant operations or configuration. Since the 
proposed change does not change the plant design, operation, or 
configuration, it could not become an accident initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind] of accident from any accident 
previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The proposed change will add a reference to an approved topical 
report to allow a limited number of lead fuel assemblies to be 
irradiated to a higher burnup level than is currently allowed at 
Calvert Cliffs. The higher burnup limit has been evaluated and 
approved in the topical report being referenced. Calvert Cliffs 
conforms to the requirements of the topical report. The addition of 
an approved reference to the Technical Specifications is 
administrative in nature and has no impact on the margin of safety 
for any plant configuration or on system performance that is relied 
upon to mitigate the consequences on an accident.
    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: Carey Fleming, Esquire, Senior Counsel--
Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt 
Street, 17th floor, Baltimore, MD 21202.
    NRC Acting Branch Chief: John P. Boska.

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: February 2, 2007.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TS) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of Three Mile Island

[[Page 20379]]

(TMI) Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
combustible gas control were imposed by Order for many facilities and 
were added to or included in the TS for nuclear power reactors 
currently licensed to operate. The revised Title 10 of the Code of 
Federal Regulations (10 CFR) 50.44, ``Standards for Combustible Gas 
Control System in Light-Water-Cooled Power Reactors,'' eliminated the 
requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50374), on possible 
amendments to eliminate requirements regarding containment hydrogen 
recombiners and the removal of requirements from TS for containment 
hydrogen and oxygen monitors, including a model safety evaluation and 
model No Significant Hazards Consideration (NSHC) Determination, in 
accordance with the Consolidated Line Item Improvement Process. The NRC 
staff subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 25, 2003 (68 FR 55416). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
February 2, 2007.
    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.
    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG 1.97 Category 1, is intended for key variables that most directly 
indicate the accomplishment of a safety function for design-basis 
accident events. The hydrogen monitors no longer meet the definition 
of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 
50.44 the Commission found that Category 3, as defined in RG 1.97, 
is an appropriate categorization for the hydrogen monitors because 
the monitors are required to diagnose the course of beyond design-
basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3 and removal of 
the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.
    Criteria 2: The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From Any Previously 
Evaluated.
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3: The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    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, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: March 19, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) 3.8.1 entitled ``AC Sources-
Operating'' to change the minimum Emergency Diesel Generator (EDG) 
output voltage acceptance criterion from 3740 to 3873 volts. 
Specifically, the proposed change would revise the Surveillance 
Requirements (SRs) 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 3.8.1.14, and 
3.8.1.17.
    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:


[[Page 20380]]


    1. The proposed change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    The increase in the minimum EDG output voltage acceptance 
criterion value in TS 3.8.1 surveillance requirements does not 
adversely affect any of the parameters in the accident analyses. The 
change increases the minimum allowed EDG output voltage acceptance 
criterion to ensure that sufficient voltage is available to operate 
the required Emergency Safety Feature (ESF) equipment under accident 
conditions. The increase in the minimum allowed EDG output voltage 
in the TS surveillance requirements ensures that adequate voltage is 
available to support the assumptions made in the Design Bases 
Accident (DBA) analyses. DBA analyses assume that onsite standby 
emergency power will provide an adequate power source to operate 
safe shutdown equipment and to mitigate consequences of design bases 
accidents. This conservative change of the acceptance criterion 
enhances the testing requirements of the onsite emergency diesel 
generators and ensures the reliability of this power source. 
Changing the acceptance criterion does not affect the probability of 
evaluated accidents and it provides better assurance of EDG 
reliability in mitigating consequences of accidents. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from an accident previously evaluated.
    The change in the value of the minimum EDG output voltage 
acceptance criterion supports the assumptions in the accident 
analyses that sufficient voltage will be available to operate ESF 
equipment on the Class 1E buses when these buses are powered from 
the onsite emergency diesel generators. The maximum EDG output 
voltage of 4580 volts is not affected by this change. The change in 
the minimum EDG output voltage from 3740 to 3873 volts ensures the 
reliability of the onsite emergency power source. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This proposed license amendment involves a change in the minimum 
EDG output voltage acceptance criterion in TS 3.8.1 surveillance 
requirements. The surveillance frequency and the different test 
requirements are unchanged. The change provides a better assurance 
that the onsite power source is able to satisfy the design 
requirements assumed in the accident analyses to safely shutdown the 
reactor and mitigate the consequences of design bases accidents. 
Therefore, the proposed change 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: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Branch Chief: L. Raghavan.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: November 8, 2006.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) Action and Surveillance 
Requirements (SRs) for instrumentation identified in TSs 3.3.1 and 
3.3.2. In particular, the proposed amendment adds actions to address 
the inoperability of one or more automatic bypass removal channels; 
revises the terminology used in the notation of TS Tables 2.2-1 and 
3.3-1 relative to the implementation and automatic removal of certain 
Reactor Protection System (RPS) trip bypasses; revises the frequency 
for performing surveillance of the automatic bypass removal function 
logic; and incorporates two administrative changes.
    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:

    Criterion 1: Does the proposed amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes to Technical Specifications 2.2.1, 3.3.1 
and 3.3.2 do not adversely impact structure, system, or component 
design or operation in a manner that would result in a change in the 
frequency of occurrence of accident initiation. The proposed 
technical specification changes do not involve accident initiators, 
do not change the configuration or method of operation of any plant 
equipment that is used to mitigate the consequences of an accident, 
and do not alter any conditions assumed in the plant accident 
analyses. The proposed amendment does not change the function or the 
manner of operation of the RPS or ESFAS [engineered safety features 
actuation system] trip bypass features. Adding actions to be taken 
for an inoperable automatic bypass removal function places 
additional restriction on plant operation in this condition and does 
not alter the setpoint or the logic of the operating bypasses and 
automatic bypass removals. Clarifying the frequency of the SR 
associated with testing the automatic bypass removal function does 
not alter the setpoint or the manner of operation of the operating 
bypasses and automatic bypass removals. More accurately reflecting 
the input process variable of the operating bypasses and automatic 
bypass removals of the affected reactor trips does not alter the 
setpoint nor the manner of operation of the operating bypasses and 
automatic bypass removals. With respect to the incorporation of the 
administrative changes, the proposed changes are spelling 
corrections and do not alter any of the requirements of the affected 
TS. Therefore, this change does not impact the consequences of any 
accident. Based on this discussion, the proposed amendment does not 
increase the probability or consequence of an accident previously 
evaluated.
    Criterion 2: Does the proposed amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new or different accidents result from clarifying actions for 
an inoperable automatic bypass removal function, clarifying 
surveillance requirements for the automatic bypass removal function, 
and more accurately reflecting the parameter being measured for 
automatic bypass removal by referring to logarithmic power, the 
input process variable. The results of previously performed accident 
analyses remain valid. The proposed amendment does not introduce 
accident initiators or malfunctions that would cause a new or 
different kind of accident. The proposed amendments are 
administrative in nature and will not change the physical plant or 
the modes of plant operation defined in the facility operating 
license. The changes do not involve the addition or modification of 
equipment nor do they alter the design or operation of plant 
systems. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?
    Response: No.
    The proposed change does not alter the function or manner of 
operation of the operating bypasses and automatic bypass removals of 
the affected reactor trips. The proposed changes do not affect any 
of the assumptions used in the accident analysis, nor do they affect 
any operability requirements for equipment important to plant 
safety. Therefore, the proposed amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.

[[Page 20381]]

    NRC Branch Chief: Harold K. Chernoff.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: January 4, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification for Limiting Conditions for 
Operation (LCOs) and Surveillance Requirements (SRs) for control rod 
operability, scram insertion times, and control rod accumulators. 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 changes extend the frequency and revise the 
methodology for testing control rod scram times, and identify a new 
category of ``slow'' control rods for assessing control rod 
operability. The frequency of control rod scram testing is not an 
initiator of any accident previously evaluated. The frequency of 
surveillance testing does not affect the ability to mitigate any 
accident previously evaluated, because the tested component is still 
required to be operable. The proposed test methodology is consistent 
with industry approved methods and ensures control rod operability 
requirements for the number and distribution of operable, slow, and 
stuck control rods continue to satisfy scram reactivity rate 
assumptions used in plant safety analysis.
    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 do not involve any physical alteration of 
the plant (no new or different type of equipment is being installed) 
and do not involve a change in the design, normal configuration, or 
basic operation of the plant. The proposed changes do not introduce 
any new accident initiators. The proposed changes do not involve 
significant changes in the fundamental methods governing normal 
plant operation and do not require unusual or uncommon operator 
actions. The proposed changes provide assurance that the plant will 
not be operated in a mode or condition that violates the assumptions 
or initial conditions in the safety analyses and that SSCs 
[structures, systems, and components] remain capable of performing 
their intended safety functions as assumed in the same analyses. 
Consequently, the response of the plant and the plant operator to 
postulated events will not be significantly different.
    Therefore, the proposed TS change does 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.
    Margin of safety is related to confidence in the ability of the 
fission product barriers to perform their design functions during 
and following an accident situation. The proposed changes address 
control rod scram test performance and acceptance criteria as well 
as control rod operability requirements. The scam test acceptance 
criteria and control rod operability restrictions are based on 
industry approved methodology and will continue to ensure control 
rod scram design functions and reactivity insertion assumptions used 
in safety analyses continue to be protected. The proposed changes 
also extend the frequency of testing control rod scram times while 
at-power from 120 days to 200 days. The proposed change ensures 
scram testing is performed and that test results verify acceptable 
operation of the control rods.
    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 50.929(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Travis C. McCullough, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    Branch Chief: John P. Boska (Acting).

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

    Date of amendment request: March 15, 2007.
    Description of amendment request: The proposed amendment would 
revise containment systems surveillance requirements for Technical 
Specification (TS) 3/4.6.2, ``Depressurization, Cooling, and pH Control 
Systems.'' The proposed amendment would revise the frequency for ANO-2 
TS Surveillance Requirement 4.6.2.1.d to require verification that 
spray nozzels are unobstructed following maintenance that could result 
in a nozzel blockage (loss of foreign material exclusion control) 
rather than performing an air or smoke flow test through each spray 
header every 5 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. Do[es] the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Containment Spray System (CSS) is not an initiator of any 
analyzed event. The proposed change does not have a detrimental 
impact on the integrity of any plant 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 CSS is designed to address the consequences of a Loss of 
Coolant Accident (LOCA) or a Main Steam Line Break (MSLB). The 
Containment 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. Do[es] 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. Do[es] the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The system is not susceptible to corrosion-induced obstruction 
or obstruction from sources external to the system. Strict controls 
are established to ensure the foreign material is not introduced 
into the CSS 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 reduction in a 
margin of safety.


[[Page 20382]]


    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 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: March 1, 2007.
    Description of amendment request: The proposed change would revise 
the Grand Gulf Nuclear Station, Unit 1 (GGNS) Technical Specifications 
(TS) to add a note to the Required Actions of TS 3.6.1.3, ``Primary 
Containment Isolation Valves (PCIVs),'' Actions A.1 and B.1. GGNS TS 
3.6.1.3 requires specific actions to be taken for inoperable PCIVs. The 
TS Required Actions include isolating the affected penetration by use 
of a closed and deactivated automatic valve, closed manual valve, blind 
flange, or check valve with flow through the valve secured. The new 
note would allow a relief valve to be used without being deactivated, 
to comply with TS 3.6.1.3, Actions A.1 and B.1, provided it has a 
relief setpoint of at least 1.5 times containment design pressure 
(i.e., at least 23 pounds per square inch gauge) and meets one of the 
following criteria:
    1. The relief valve is 1-inch nominal size or less, or
    2. The flow path is into a closed system whose piping pressure 
rating exceeds the containment design pressure rating.
    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.
    Primary Containment Isolation Valves (PCIVs) are accident 
mitigating features designed to limit releases from the containment 
following an accident. The TS specify actions to be taken to 
preserve the containment isolation function if a PClV is inoperable. 
These actions include isolating the penetration flow path by 
specific methods including, closed and de-activated automatic 
valves, closed manual valves, blind flanges, and check valves with 
flow through the valve secured. The current TS Actions do not 
specifically recognize a closed relief valve as an acceptable method 
of isolating a penetration flow path. Thus, special measures may 
need to be taken to comply with the TS Required Actions, such as 
replacing the relief valve with a blind flange or de-activating the 
relief valve by installing a gag. While such actions may provide 
additional assurance of preserving the containment isolation 
function, it may also have adverse safety affects such as disabling 
the overpressure protective safety feature, causing additional 
safety system unavailability time, and increasing occupational dose.
    The proposed change would allow certain relief valves to be used 
for isolating the penetration flow path without being de-activated. 
The proposed TS changes do not alter the design, operation, or 
capability of PCIVs. Relief valves are designed to be normally 
closed to preserve the piping boundary integrity yet automatically 
open on an abnormal process pressure to protect the piping from 
overpressure conditions. Relief valves may also serve as passive 
containment isolation devices (i.e., they do not require mechanical 
movement to perform the isolation function). The proposed TS changes 
preserve both the containment isolation and piping overpressure 
protection functions.
    The failure of a relief valve to remain closed during or 
following an accident is considered a low probability because relief 
valves are passive isolation devices that do not require mechanical 
movement to perform the isolation function and the relief setpoint 
provides sufficient margin to preclude the potential for premature 
opening due to containment post-accident pressures. Additional 
criteria are established to provide defense-in-depth protection. 
Relief valves that are one-inch or smaller provide an additional 
physical barrier in that, even in the unlikely event that a relief 
valve were to fail to remain fully closed during or following an 
accident, the size restriction would limit leakage such that a large 
early release would not occur. By definition, penetrations one-inch 
and smaller do not contribute to large early releases. Larger relief 
valves may be used as isolation devices provided that the 
containment penetration flow path through the relief valve would be 
contained in a closed system. In the unlikely event that a relief 
valve were to fail to remain closed, the leakage would be into a 
system which forms a closed loop outside primary containment and any 
containment leakage would return to primary containment through this 
closed loop.
    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 introduce any new modes of plant 
operation or adversely affect the design function or operation of 
safety features. The proposed TS change allows use of existing plant 
equipment as compensatory measures to maintain the containment 
isolation design intent when the normal isolation features are 
inoperable. Since relief valves used for this purpose will not be 
disabled by gags or blind flanges, the system piping overpressure 
protection design feature will also be preserved.
    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 safety margin associated with this change is that associated 
with preserving the containment integrity. NUREG-0800, the Standard 
Review Plan, recognizes that relief valves with relief setpoints 
greater than 1.5 times containment design pressure are acceptable as 
containment isolation devices. Closed relief valves with relief 
setpoints of this margin provide an isolation alternative that is 
less susceptible to a single failure (i.e., inadvertent opening) yet 
still preserves the overpressure protection that the component was 
intended to provide. The failure of a relief valve to remain closed 
during or following an accident is considered a low probability 
because relief valves are passive isolation devices that do not 
require mechanical movement to perform the isolation function and 
the relief setpoint provides sufficient margin to preclude the 
potential for premature opening due to containment post-accident 
pressures. Defense-in-depth containment leakage protection is 
provided by additional TS criteria that limit the use of relief 
valves to those one-inch or less in size or those where containment 
leakage would be into a closed system whose piping pressure rating 
exceeds the containment design pressure rating. Relief valves that 
are one-inch or smaller provide an additional physical barrier in 
that, even in the unlikely event that a relief valve were to fail to 
remain closed during or following an accident, the size restriction 
would limit leakage such that a large early release would not occur. 
In the unlikely event that a relief valve larger than one-inch were 
to fail to remain closed, the leakage would be into a system which 
forms a closed loop outside primary containment and any containment 
leakage would return to primary containment through this closed 
loop.
    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 
Council--

[[Page 20383]]

Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: David Terao.

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: February 9, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) 3.3.2, ``Engineered Safety 
Feature Actuation System Instrumentation,'' TS 3.5.2, ``Emergency Core 
Cooling System--Operating,'' TS 3.6.5, ``Containment Air Temperature,'' 
and TS 5.5.12, ``Containment Leakage Rate Testing Program.'' The 
revised TSs would be consistent with a proposed change to the 
Recirculation Spray System (RSS) pump start signal due to a 
modification to the containment sump screens.
    The proposed amendment would also replace the use of LOCTIC with 
the Modular Accident Analysis Program-Design Basis Accident (MAAP-DBA) 
for calculating containment pressure, temperature, and condensation 
rates for input to the SWNAUA code. The calculation methodology change 
would ultimately change the aerosol removal coefficients used in dose 
consequence analysis.
    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 to the RSS pump start signal, 
the upper containment temperature technical specification (TS) 
limit, the peak containment internal pressure, the nomenclature for 
automatic switchover to the containment sump, and the containment 
sump screen visual inspection surveillance requirement do not 
involve any system or component that are accident initiators. The 
RSS is used for accident mitigation only. The Refueling Water 
Storage Tank (RWST) level and containment pressure instrumentation 
will continue to comply with all applicable regulatory requirements 
and design criteria (e.g., train separation, redundancy, single 
failure, etc.) following approval of the proposed changes. The 
design functions performed by the RSS and the containment are not 
changed by this license amendment request.
    Delaying the start of the RSS pumps and the change to the upper 
containment temperature affect the long-term containment pressure 
and temperature profiles. The environmental qualification of safety-
related equipment inside containment will be confirmed to be 
acceptable and accident mitigation systems will continue to operate 
within design temperatures and pressures. Delaying the RSS pump 
start reduces the emergency diesel generator loading in the early 
stage of a design basis accident and maintaining the staggered 
loading of the RSS pump starts avoids overloading on each emergency 
diesel generator at Unit 1. Staggered loading of the emergency 
diesel generator is not required for Unit 2.
    The methodology change to calculate containment pressure, 
temperature and condensation rates for input to the SWNAUA code will 
not involve a significant increase in the probability of an accident 
previously evaluated because this change in methodology does not 
impact accident initiators.
    The loss of coolant accident (LOCA) has been evaluated using the 
guidance provided in Regulatory Guide 1.183, ``Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors.'' The radiological consequences of the 
remaining design basis accidents are not significantly impacted by 
the proposed changes. As demonstrated by the supporting analyses, 
the estimated dose consequences at the Exclusion Area Boundary 
(EAB), Low Population Zone (LPZ), and control room remain within the 
acceptance criteria of 10 CFR 50.67 as supplemented by Regulatory 
Guide 1.183 and Standard Review Plan Section 15.0.1. In addition, 
the supporting analyses also demonstrates that the dose consequences 
in the Emergency Response Facility remain compliant with paragraph 
IV.E.8 of Appendix E, to 10 CFR part 50, Emergency Planning and 
Preparedness for Production and Utilization Facilities, regulatory 
guidance provided in Supplement 1 of NUREG-0737. The revised 
radiological analyses results in a slight increase in control room 
and off-site doses; however, the radiological analyses and 
evaluations developed in support of this application demonstrate 
that the proposed changes will not impact compliance with applicable 
regulatory requirements and will not involve a significant increase 
in the consequences of an accident previously evaluated. The slight 
increase in control room and off-site doses is more than offset by 
the increased assurance of adequate NPSH [net positive suction head] 
to the RSS pumps and Emergency Core Cooling System operability.
    The safety analysis acceptance criteria will continue to be met 
following the proposed changes to the RSS pump start signal, visual 
sump inspection, TS containment upper temperature limit, peak 
containment internal pressure, nomenclature for automatic switchover 
to the containment sump and the change to the control room and off-
site dose consequences analyses.
    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. One of the proposed changes alters the RSS pump 
start circuitry by initiating the pump start from a coincident 
Containment Pressure High-High/[RWST] Level Low signal instead of 
from a timer. The RSS pump instrumentation will be included as part 
of the Engineered Safety Feature Actuation System (ESFAS) 
instrumentation in the TS and will be subject to the ESFAS 
surveillance requirements following approval of the proposed 
changes. The design of the RSS pump start instrumentation complies 
with all applicable regulatory requirements and design criteria. The 
failure modes have been analyzed to ensure that the revised RSS pump 
start circuitry can withstand a single active failure without 
affecting the RSS design functions. The RSS is an accident 
mitigation system only, so no new accident initiators are created.
    It is not expected that the change in containment temperature 
will have a significant impact on equipment qualification. However, 
any equipment that must be replaced or re-qualified will be 
addressed prior to operation with the proposed change to RSS pump 
start. As a result any such equipment will not introduce new failure 
modes, accident initiators, or malfunctions that would cause a new 
or different kind of accident.
    The remaining changes do not change plant equipment design or 
function and therefore will not introduce new failure modes, 
accident initiators, or malfunctions that would cause a new or 
different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No. The changes to the RSS pump start signal and the 
upper containment temperature limit affect the containment response 
and the LOCA dose analyses. Analyses demonstrate that containment 
design basis limits are satisfied and post-LOCA offsite and control 
room dose criteria will continue to be met following approval of the 
proposed changes.
    The change to the containment sump visual inspection will not 
involve a significant reduction in a margin of safety because the 
revised surveillance will continue to provide adequate assurance the 
sump screens are not blocked with debris and that signs of corrosion 
will be detected.
    The change to peak containment internal pressure will not result 
[in] a significant reduction in a margin of safety because the new 
pressure is lower for each of the units.
    Although the control room and off-site doses slightly increase 
(due to a combination of the change to the start signal and the 
proposed methodology change), the increase will not involve a 
significant reduction in a margin of safety because operator and 
public exposure limits will continue to meet applicable regulatory 
requirements.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.


[[Page 20384]]


    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 W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Acting Branch Chief: John P. Boska.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

    Date of amendment request: March 8, 2007.
    Description of amendment request: The proposed amendment would add 
Technical Specification (TS) Limiting Condition for Operation (LCO) 
3.0.8 to allow a delay time for entering a supported system TS when the 
inoperability is due solely to an inoperable snubber, if risk is 
assessed and managed. The proposed change is consistent with TS Task 
Force (TSTF) change TSTF-372-A, Revision 4, ``Addition of LCO 3.0.8, 
Inoperability of Snubbers.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on November 24, 
2004 (69 FR 68412). The licensee affirmed the applicability of the 
model 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 is presented below:
    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable snubber 
if risk is assessed and managed. The postulated seismic event requiring 
snubbers is a low-probability occurrence and the overall TS system 
safety function would still be available for the vast majority of 
anticipated challenges. Therefore, the probability of an accident 
previously evaluated is not significantly increased, if at all. The 
consequences of an accident while relying on allowance provided by 
proposed LCO 3.0.8 are no different than the consequences of an 
accident while relying on the TS required actions in effect without the 
allowance provided by proposed LCO 3.0.8. Therefore, the consequences 
of an accident previously evaluated are not significantly affected by 
this change. The addition of a requirement to assess and manage the 
risk introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident From Any Previously Evaluated.
    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is assessed 
and managed, will not introduce new failure modes or effects and will 
not, in the absence of other unrelated failures, lead to an accident 
whose consequences exceed the consequences of accidents previously 
evaluated. The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Thus, this change does not create the possibility of a new or different 
kind of accident from an accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic event 
requiring snubbers is a low-probability occurrence and the overall TS 
system safety function would still be available for the vast majority 
of anticipated challenges. The risk impact of the proposed TS changes 
was assessed following the three-tiered approach recommended in 
Regulatory Guide 1.177. A bounding risk assessment was performed to 
justify the proposed TS changes. This application of LCO 3.0.8 is 
predicated upon the licensee's performance of a risk assessment and the 
management of plant risk. The net change to the margin of safety is 
insignificant. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Acting Branch Chief: John P. Boska.

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

    Date of amendment request: February 15, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) Surveillance Requirement (SR) 
3.8.4.2 to correct errors inadvertently introduced by Amendment No. 
146. SR 3.8.4.2 currently requires that each battery charger be 
verified to supply greater than or equal to 150 amps for 250-volt DC 
subsystems, and greater than or equal to 50 amp for 125-volt DC 
subsystems. The licensee proposed to correct the errors by 
differentiating that the Division 1 battery chargers are verified to 
supply greater than or equal to 150 amps and the Division 2 battery 
chargers are verified to supply greater than or equal to 110 amps. The 
licensee stated that the Division 2 battery charger output current 
limiter is field-adjusted to supply 120 to 125 amps in order to stay 
within the electrical circuit breaker ratings in the associated 
distribution cabinet.
    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 (NSHC). The NRC staff reviewed the licensee's analysis, 
and has performed its own analysis as follows:

    (1) Does the proposed amendment involve a significant increase 
in the probability or consequence of an accident previously 
evaluated?
    No. The proposed amendment would only correct the battery 
chargers' DC supply current limits specified by SR 3.8.4.2. The 
current limits of the battery chargers were not considered to be a 
precursor to, and does not affect the probability of, an accident. 
In addition, there is no design or operation change associated with 
the proposed amendment. Therefore, the proposed amendment does not 
increase the probability of an accident previously evaluated.
    The corrected DC supply current limits of the battery chargers 
will ensure that the batteries will be charged under as-designed 
conditions. The corrected limits will not decrease the functionality 
of the Division 1 or Division 2 battery chargers, or the 
functionality of the batteries the battery chargers support. 
Therefore, the plant systems required to mitigate accidents will 
remain capable of performing their design functions. As a result, 
the proposed amendment will not lead to a significant change in the 
consequences of any accident.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of

[[Page 20385]]

accident from any accident previously evaluated?
    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 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 revised acceptance value. Thus, 
the proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed amendment would only change the current supply 
limits of the battery chargers. There will be no modification of any 
TSs limiting condition for operation, no change to any limit on 
previously analyzed accidents, no change to how previously analyzed 
accidents or transients would be mitigated, no change in any 
methodology used to evaluate consequences of accidents, and no 
change in any operating procedure or process. Therefore, the 
proposed amendment does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
the NRC staff's own analysis above, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the proposed amendment involves no significant hazards 
consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: March 30, 2007.
    Description of amendment requests: The proposed change will revise 
Technical Specifications (TSs) Surveillance Requirement (SR) 3.3.7.3.b, 
``Loss of Voltage Function'' to a narrower voltage band and lower 
operating time for channel calibration testing, by replacing the 
undervoltage relays with the reset time significantly lower.
    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 revises the Technical Specifications 
Surveillance Requirement 3.3.7.3.b allowable set point values of the 
Loss of Voltage Function for the channel calibration testing. This 
proposed change will allow Southern California Edison (SCE) to 
increase margin and conservatism for the loss of voltage relay 
settings and overall loop uncertainties while performing Loss of 
Voltage Signal (LOVS) channel calibration testing.
    The loss of voltage function is detected by the LOVS relays 
installed on the 4.16 kV Safety Related busses. Normally, these 
devices are not considered to be accident initiators. The proposed 
change narrows the voltage operating band and lowers the allowable 
upper limit for this loss of voltage detection by use of the 
electronic type Basler BE1-27 under-voltage relays. However, the 
reset time of the relay [will be reduced] significantly. [Therefore, 
t]he proposed change does not impact 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 [an] accident previously evaluated?
    Response: No.
    The proposed allowable values for the LOVS relays voltage 
settings and the minimum operating voltage of the of[f]site power 
will provide acceptable level of protection for the plant equipment.
    3. Does the proposed change involve [a] significant reduction in 
a margin of safety?
    Response: No.
    The proposed loss of voltage function is designed to ensure that 
plant equipment will not operate beyond its normal operating range 
for satisfactory operation of all the safety related equipment. The 
proposed loss of voltage function values will not affect the 
existing protection criterion for the plant equipment and will not 
reduce margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Thomas G. Hiltz.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 16, 2007.
    Description of amendment request: The proposed amendment would 
permanently revise Technical Specification 2.2.1, Table 2.2-1, 
Functional Unit 17.A, Turbine Trip Low Trip System Pressure allowable 
value. The proposed revision was previously approved for one operating 
cycle at each unit.
    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 revises the allowable value for reactor trip 
as a result of a turbine trip on low trip system pressure. This 
change will not alter any plant components, systems, or processes 
and will only provide a more appropriate value to assess operability 
of the associated pressure switches. Since the plant features and 
operating practices are not altered, the possibility of an accident 
is not affected. This reactor trip is not directly credited in SQN's 
[Sequoyah Nuclear Plant's] accident analysis and is maintained as an 
anticipatory trip to enhance the overall reliability of the reactor 
trip system. As such, there is not a specific safety limit 
associated with this function and the generation of a reactor trip 
based on low trip system pressure is above the required actuations 
to ensure acceptable mitigation of accidents. As the proposed change 
will continue to provide an acceptable anticipatory trip signal, the 
offsite dose potential is not affected by this change. 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.
    As described above, this change will not alter any plant 
equipment or operating practices that have the ability to create a 
new potential for accident generation. The proposed change revises 
the operability limits for a function that generates a trip signal 
when appropriate conditions exist to require accident mitigation 
response. This type of function does not have the ability to create 
an accident as its purpose and function is to mitigate events. 
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 revise an allowable value for a reactor 
trip initiator that results from a turbine trip condition. This 
change will not alter the setpoint, and the calibration of the 
associated pressure switches will continue to be set at the current 
value. The allowable value change is in response to accuracy aspects 
of the

[[Page 20386]]

instrumentation and does not alter the ability of this trip function 
to operate when and as needed to mitigate accident conditions. 
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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#cabaaeb88aa4b8a9e4ada5bc"><span class="__cf_email__" data-cfemail="64140016240a16074a030b12">[email protected]</span></a>.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: January 19, 2007, as 
supplemented by letters dated March 13 and 22, 2007.
    Brief description of amendment: The amendment modifies Technical 
Specifications 5.5.9 and 5.6.8 to add steam generator alternate repair 
criteria and additional steam generator reporting criteria at H. B. 
Robinson Steam Electric Plant, Unit No. 2.
    Date of issuance: April 9, 2007.
    Effective date: This license amendment is effective as of the date 
of issuance and shall be implemented within 60 days.
    Amendment No.: 214.
    Renewed Facility Operating License No. DPR-23. Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: January 30, 2007 (72 FR 
4300). The March 13 and 22, 2007, supplemental letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 9, 2007.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: March 28, 2006, as supplemented 
by letters dated October 26, and December 4, 2006, and January 26, 
2007.
    Brief description of amendment: The amendment revises Millstone 
Power Station, Unit No. 3 Technical Specifications (TS) to delete 
redundant surveillance requirements pertaining to post-maintenance/
post-modification testing.
    Date of Issuance: March 29, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 237.
    Facility Operating License No. NPF-49: Amendment revised the TS.
    Date of initial notice in Federal Register: May 23, 2006 (71 FR 
29673). The supplemental letters provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2007.
    No significant hazards consideration comments received: No.

Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: April 11, 2006, as supplemented 
October 24, 2006.
    Brief description of amendments: The amendments revised the 
Technical Specifications requirements related to steam generator tube 
integrity consistent with the NRC-approved Revision 4 to Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler TSTF-449, ``Steam Generator Tube Integrity.'' These amendments 
also remove license conditions that become outdated with these TS 
changes. In addition, the amendments revised the organizational 
description in TS 5.2.1, which is solely administrative and unrelated 
to steam generator tube integrity.
    Date of Issuance: April 2, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 355, 357, 356.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: January 3, 2007 (72 FR 
149). The supplement dated October 24, 2006, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated April 2, 2007. No significant hazards 
consideration comments received: No.

[[Page 20387]]

Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: April 11, 2006, as supplemented 
by letter dated March 14, 2007.
    Brief description of amendments: The amendments added Technical 
Specification (TS) Limiting Condition for Operation (LCO) 3.0.8 to 
allow a delay time for entering a supported system TS when the 
inoperability is due solely to an inoperable snubber, if risk is 
assessed and managed with an approved Bases Control Program that is 
consistent with the TS Bases Control Program described in Section 5.5 
of the applicable vendor's Standard Technical Specifications.
    Date of Issuance: April 2, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 356, 358, 357.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the licenses and the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 2007 (72 FR 
151). The supplement provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register on January 3, 2007 (72 FR 151). The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated April 2, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 18, 2005, as 
supplemented by letters dated August 18 and September 28, 2006, and 
February 15, February 23, and March 7, 2007.
    Brief description of amendment: The amendments would revise the 
existing steam generator tube surveillance program using Technical 
Specification Task Force Traveler No. 449 (TSTF-449), Revision 4, 
``Steam Generator Tube Integrity'' as a basis. The amendments would 
also revise TS 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' regarding the required SG inspection scope for Byron 
Station, Unit No. 2, during outage number 13 and subsequent operating 
cycle. A similar approval was granted for Braidwood Station, Unit 2 by 
letter from the NRC dated October 24, 2006.
    Date of Issuance: March 30, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 150/150, 144/144.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: May 23, 2006 (71 FR 
29676). The August 18 and September 28, 2006 and February 15, February 
23, and March 7, 2007 supplements, contained clarifying information and 
did not change the staff's initial proposed finding of no significant 
hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 30, 2007.
    No significant hazards consideration comments received: No.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: March 23, 2006, as supplemented by 
letters dated August 16 and November 28, 2006.
    Description of amendment request: The amendment revises the 
Seabrook Station, Unit No. 1 Technical Specifications (TSs) 
Definitions, TS 3.4.5, ``Steam Generator (SG) Tube Integrity,'' and TS 
3.4.6.2, ``Reactor Coolant System Operational Leakage'' consistent with 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Traveler TSTF-449, ``Steam Generator Tube Integrity,'' 
Revision 4. Additionally the amendment creates TS 6.7.6.k. ``Steam 
Generator (SG) Program'' and TS 6.8.1.7, ``Steam Generator Tube 
Inspection Report,'' consistent with TSTF-449, Revision 4.
    Date of Issuance: March 28, 2007.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 115.
    Facility Operating License No. NPF-86: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: April 25, 2006 (71 FR 
23955). The licensee's August 16 and November 28, 2006, supplements 
provided clarifying information that did not change the scope of the 
proposed amendment as described in the original notice of proposed 
action published in the Federal Register, and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2007.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-266, Point Beach Nuclear 
Plant, Unit 1, Town of Two Creeks, Manitowoc County, Wisconsin

    Date of application for amendments: July 11, 2006, as supplemented 
January 19, March 9 and 26, 2007.
    Brief description of amendments: The amendment revises Technical 
Specification (TS) 5.5.8, ``Steam Generator Program,'' to change the 
inspection and repair criteria for portions of the tubes within the 
hot-leg region of the tubesheet for a single operating cycle. In 
addition, an administrative change corrects a page number in the TS 
Table of Contents and deletes two blank pages in TS Section 5.0.
    Date of Issuance: April 4, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 226.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Technical Specifications/License.
    Date of initial notice in Federal Register: August 29, 2006 (71 FR 
51230). The supplements dated January 19, March 9 and 26, 2007, 
contained clarifying information and did not change the staff's initial 
proposed finding of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 2007.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: December 20, 2006.
    Brief description of amendment: The amendment removed annotations 
referencing Technical Data Book (TDB)-VIII, ``Equipment Operability 
Guidance,'' and annotations referencing Technical Specification 
Interpretations (TSIs) from the NRC Authority File of the Technical 
Specifications (TSs). These documents are used by Omaha Public Power 
District (OPPD) personnel

[[Page 20388]]

for additional guidance in applying certain Limiting Conditions of 
Operation requirements to specific equipment and/or situations. OPPD 
has annotated references to these documents in the TS copies used at 
the Fort Calhoun Station, Unit No.1 (FCS); however, these annotations 
were inadvertently included into the NRC Authority File and are not 
officially part of the FCS TS. The amendment also corrected a 
discrepancy in TS 2.10.4(1)(c).
    Date of Issuance: April 3, 2007.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 249.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 30, 2007 (72 FR 
4308).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated April 3, 2007.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, 
Unit No. 2, Salem County, New Jersey

    Date of application for amendment: April 6, 2006, as supplemented 
by letters dated January 19, and February 27, 2007.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) related to steam generator tube integrity 
consistent with Revision 4 to Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler 449 (TSTF-449), 
``Steam Generator Tube Integrity.''
    Date of Issuance: March 29, 2007.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 262.
    Facility Operating License No. DPR-75: The amendment revised the 
TSs and the License.
    Date of initial notice in Federal Register: July 18, 2006 (71 FR 
40753). The letters dated January 19, and February 27, 2007, provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the 
application beyond the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2007.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: June 2, 2006, as supplemented 
by letter dated October 19, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.8.1, ``AC [alternating current] Sources--
Operating,'' and TS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting 
Air,'' to increase the required amount of stored diesel fuel oil to 
support a change to Ultra Low Sulfur Diesel fuel from California diesel 
fuel presently in use. This change in the type of fuel oil is mandated 
by California air pollution control regulations.
    Date of Issuance: April 4, 2007.
    Effective date: As of its issuance and shall be implemented within 
60 days of issuance.
    Amendment Nos.: Unit 2--211; Unit 3--203.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 18, 2006 (71 FR 
40754). The supplemental letter dated October 19, 2006, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 2007.
    No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: February 21, 2006.
    Brief description of amendments: The amendments revised Technical 
Specifications 1.1, ``Definitions,'' and 3.4.16, ``RCS [Reactor Coolant 
System] Specific Activity.'' The revisions replaced the current 
Limiting Condition for Operation (LCO) 3.4.16 limit on RCS 
grossspecific activity with limits on RCS Dose Equivalent I-131 (DEI) 
and Dose Equivalent Xe-133 (DEX). The conditions and required actions 
for LCO 3.4.16 not being met, and surveillance requirements for LCO 
3.4.16, are revised. The modes of applicability for LCO 3.4.16 are 
extended. TS Figure 3.4.16-1 on the limit for DEI with respect to rated 
thermal power is deleted.
    Date of issuance: March 29, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 137/137.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: February 27, 2007 (72 
FR 8805).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 2007.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: August 17, 2006.
    Brief description of amendment: The amendment revised Technical 
Specifications (TSs) 2.1.1, ``Reactor Core SLs [Safety Limits],'' 
3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' 3.4.1, RCS 
[Reactor Coolant System] Pressure, Temperature, and Flow Departure from 
Nucleate Boiling (DNB) Limits,'' and 5.6.5, ``Core Operating Limits 
Report (COLR).'' The changes (1) relocated certain operating cycle-
specific core operating limits, including TS Figure 2.1.1-1, ``Reactor 
Core Safety Limits,'' from the TSs to the plant COLR, (2) added two new 
safety limits for departure from nucleate boiling ratio and peak fuel 
centerline temperature, and (3) added topical reports to TS 5.6.5 and 
had the reports cited by only the report title and number. TS 5.6.5 was 
expanded to include the limits from TSs 2.1.1, 3.3.1, and 3.4.1.
    Date of Issuance: April 2, 2007.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance. The final TS Bases changes 
including the licensee's application dated August 17, 2006, will be 
processed under the licensee's program for updates to the TS Bases, in 
accordance with TS 5.5.14, at the time this amendment is implemented. 
The final changes to the COLR including those in the licensee's 
application dated August 17, 2006, will be submitted to the NRC in 
accordance with the update process covered by TS 5.6.5.d.
    Amendment No.: 183.
    Facility Operating License No. NPF-30: The amendment revised the

[[Page 20389]]

Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 16, 2007 (72 FR 
1781).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 2, 2007.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50 280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: May 26, 2006, as supplemented 
on January 19, 2007.
    Brief Description of amendments: These amendments revised the 
Technical Specification (TS) requirements related to steam generator 
tube integrity and Reactor Coolant System leakage definitions and 
requirements. The TSs were revised to implement TS Task Force (TSTF) 
Standard TS Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity,'' (TSTF-449, Rev. 4) with minor deviations to be consistent 
with Surry's custom TSs.
    Date of Issuance: March 29, 2007.
    Effective date: As of date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 251, 250.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments changed the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 15, 2006 (71 FR 
46941). The supplement dated January 19, 2007, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated March 29, 2007.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: July 5, 2006, as supplemented 
on September 21 and November 20, 2006.
    Brief Description of amendments: These amendments revised the main 
control room (MCR) and emergency switchgear room (ESGR) air-
conditioning system (ACS) Technical Specifications to reflect the 
completion of permanent modifications to the equipment and associated 
power supply configuration. The revisions include the addition of 
requirements and/or action statements addressing the inoperability of 
two or more air handling units (AHUs) on a unit, as well as AHU powered 
from an H emergency bus. The proposed change, paralleling requirements 
in the Improved Technical Specifications, also adds MCR and ESGR ACS 
requirements during refueling operations and irradiated fuel movement 
in the fuel building. In addition, the proposed change clarified the 
service water requirements for the ACS chillers that serve the MCR and 
ESGRs.
    Date of Issuance: April 2, 2007.
    Effective date: As of date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 252, 251.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments changed the licenses and the technical specifications.
    Date of initial notice in Federal Register: September 26, 2006 (71 
FR 56193). The supplements dated September 21 and November 20, 2006, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated April 2, 2007.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 16th day of April 2007.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E7-7534 Filed 4-23-07; 8:45 am]
BILLING CODE 7590-01-P


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