Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 31097-31108 [E7-10590]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 107 (Tuesday, June 5, 2007)]
[Notices]
[Pages 31097-31108]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-10590]


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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 May 11, 2007, to May 23, 2007. The last 
biweekly notice was published on May 22, 2007 (72 FR 28717).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. 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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and

[[Page 31098]]

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, HearingDocket@nrc.gov; 
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 
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 
OGCMailCenter@nrc.gov. 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, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Pope County, Arkansas

    Date of amendment request: April 24, 2007.
    Description of amendment request: The proposed amendment will 
delete the Fuel Handling Area Ventilation System (FHAVS) and associated 
Ventilation Filter Testing Program (VFTP) requirements that are 
included in the ANO-1 Technical Specifications (TSs) 3.7.12 and 5.5.11 
and the ANO-2 TSs 3.9.11 and 6.5.11. These requirements will be 
relocated to a licensee-controlled document, the unit-specific 
Technical Requirements Manuals (TRM), which are controlled under 10 CFR 
50.59, ``Changes, tests, and experiments.''
    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 change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The FHAVS is not involved in the initiation of any accidents. 
The system maintains a suitable environment for equipment operation 
and personnel access. They are also designed to filter any gaseous 
radioactivity that may occur during normal or accident conditions 
(i.e., a fuel handling accident). On this basis, the system is 
currently classified and designed as an Engineered Safety Features 
(ESF) air cleanup system. The FHAVS is used during movement of 
irradiated fuel, crane operation with loads over the Spent Fuel Pool 
(SFP), fuel shipments, and spent resin transfer to pull possible 
airborne radioactivity from the Train Bay by re-positioning manual 
dampers.
    Revised ANO-1 and ANO-2 analysis of the dose consequences of 
a[n] FHA, to both the public and to the control room operator, 
demonstrate that doses remain well within regulatory acceptance 
limits without crediting filtration.
    Thus there is no required safety function for the ANO-1 or ANO-2 
FHAVS.
    Therefore, the proposed change[s] [do] not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The FHAVS is not involved in the initiation of any accidents. It 
was designed to

[[Page 31099]]

filter any gaseous radioactivity that may occur during normal or 
accident conditions (i.e., a fuel handling accident). No physical 
modifications are planned to the ANO-1 or ANO-2 FHAVS.
    Revised ANO-1 and ANO-2 analysis of the dose consequences of 
a[n] FHA, to both the public and to the control room operator, 
demonstrate that doses remain well within regulatory acceptance 
limits without crediting filtration. Thus, there is no required 
safety function for the ANO-1 or ANO-2 FHAVS.
    Therefore, the proposed change[s] [do] not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The FHAVS was designed to filter any gaseous radioactivity that 
may occur during normal or accident conditions (i.e., a fuel 
handling accident). No physical modifications are planned to the 
ANO-1 or ANO-2 FHAVS.
    Revised ANO-1 and ANO-2 analysis of the dose consequences of 
a[n] FHA, to both the public and to the control room operator, 
demonstrate that doses remain well within regulatory acceptance 
limits without crediting filtration. The margin of safety, as 
defined in Standard Review Plan 15.7.4, Revision 1, and GDC [General 
Design Criterion] 19 has not been significantly reduced.
    Therefore, the proposed change[s] [do] not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: April 24, 2007.
    Description of amendment request: The proposed amendment will 
revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification 
(TS) 5.2.1, ``Fuel Assemblies,'' to add Optimized ZIRLO\TM\ as an 
acceptable fuel rod cladding material.
    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 NRC approved topical report WCAP-12610-P-A and CENPD-404-P-
A, Addendum 1-A ``Optimized ZIRLO\TM\,'' prepared by Westinghouse 
Electric Company, LLC (Westinghouse), addresses Optimized ZIRLO\TM\ 
and demonstrates that Optimized ZIRLO\TM\ has essentially the same 
properties as currently licensed ZIRLO\TM\. The fuel cladding itself 
is not an accident initiator and does not affect accident 
probability. Use of Optimized ZIRLO\TM\ fuel cladding has been shown 
to meet all 10 CFR 50.46 design criteria and, therefore, will not 
increase the consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Use of Optimized ZIRLO\TM\ clad fuel will not result in changes 
in the operation or configuration of the facility. Topical report 
WCAP-12610-P-A and CENPD-404-P-A demonstrated that the material 
properties of Optimized ZIRLO\TM\ are similar to those of standard 
ZIRLO\TM\. Therefore, Optimized ZIRLO\TM\ fuel rod cladding will 
perform similarly to those fabricated from standard ZIRLO\TM\, thus 
precluding the possibility of the fuel becoming an accident 
initiator and causing a new or different type of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the Optimized ZIRLO\TM\ are not significantly 
different from those of standard ZIRLO\TM\. Optimized ZIRLO\TM\ is 
expected to perform similarly to standard ZIRLO\TM\ for all normal 
operating and accident scenarios, including both loss-of-coolant 
accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where 
the slight difference in Optimized ZIRLO\TM\ material properties 
relative to standard ZIRLO\TM\ could have some impact on the overall 
accident scenario, plant-specific LOCA analyses using Optimized 
ZIRLO\TM\ properties will be performed prior to the use of fuel 
assemblies with fuel rods containing Optimized ZIRLO\TM\. These LOCA 
analyses will demonstrate that the acceptance criteria of 10 CFR 
50.46 will be satisfied when Optimized ZIRLO\TM\ fuel rod cladding 
is implemented.
    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--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: May 8, 2007.
    Description of amendment request: The proposed amendment will 
revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification 
(TS) 3.1.1.4, ``Moderator Temperature Coefficient (MTC),'' to change 
the surveillance frequency to be based on effective full-power days 
instead of boron concentration.
    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 continues to perform the SRs [surveillance 
requirements] to determine MTC at test intervals associated with the 
beginning and middle of the cycle. The results of the test[s] will 
continue to verify that the predicted MTC is consistent with the 
measured [MTC].
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not result in any plant modifications 
or changes in the way the plant is operated. The revised SRs for 
confirming the MTC predicted values will continue to be performed at 
intervals associated with the beginning and middle of the cycle.
    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 does not result in any changes to the test 
method or to the frequency of the test. The change of the test 
interval to use EFPD [effective full-power

[[Page 31100]]

days] instead of RCS [reactor coolant system] boron concentration 
still provides assurance that the predicted MTC is consistent with 
the measured [MTC].
    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--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

    Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 
and 2, Ogle County, Illinois.
    Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 
2 and 3, Grundy County, Illinois.
    Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 
2, LaSalle County, Illinois.
    Docket No. 50-352 and 50-353, Limerick Generating Station, Units 1 
and 2, Montgomery County, Pennsylvania.
    Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois.
    Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 
50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, 
York and Lancaster Counties, Pennsylvania.
    AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois.
    Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean 
County, New Jersey.
    Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, 
Dauphin County, Pennsylvania.
    Date of amendment request: April 12, 2007.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements related to control 
room envelope (CRE) habitability in accordance with Technical 
Specification Task Force (TSTF) Traveler TSTF-448, Revision 3, 
``Control Room Habitability.''
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change 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 any accident previously 
evaluated.
    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does 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 the margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

Exelon Generation Company, LLC, Docket 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 amendment request: April 4, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.16, ``Containment Leakage Rate 
Testing Program,'' to reflect a one-time deferral of the containment 
Type A, integrated leak rate test from once in 10 years to once in 15 
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. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes will revise Braidwood Station and Byron 
Station TS 5.5.16, ``Containment Leakage Rate Testing Program'' to 
reflect a one-time, five-year extension of the containment Type A 
test date to enable the implementation of a 15-year test interval.
    The containment is designed to contain radioactive material that 
may be released from the reactor core following a design basis

[[Page 31101]]

Loss of Coolant Accident (LOCA). The test interval associated with 
Type A testing is not a precursor of any accident previously 
evaluated. Type A testing does provide assurance that the 
containment will not exceed allowable leakage rate criteria 
specified in the TS and will continue to perform its design function 
following an accident. A risk assessment of the proposed changes has 
concluded that there is an insignificant increase in total 
population dose rate and an insignificant increase in the 
conditional containment failure probability.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes for a one-time, five-year extension of the 
Type A tests for Braidwood Station and Byron Station will not affect 
the control parameters governing unit operation or the response of 
plant equipment to transient and accident conditions. The proposed 
changes do not introduce any new equipment, modes of system 
operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The Braidwood Station and Byron Station containment consists of 
the concrete containment building, its steel liner, and the 
penetrations through this structure. The structure is designed to 
contain radioactive material that may be released from the reactor 
core following a design basis LOCA. Additionally, this structure 
provides shielding from the fission products that may be present in 
the containment atmosphere following accident conditions.
    The containment is a reinforced concrete structure with a 
cylindrical wall, a flat foundation mat, and a shallow dome roof. 
The inside surface of the containment is lined with a carbon steel 
liner to ensure a high degree of leak tightness during operating and 
accident conditions. The cylinder wall is pre-stressed with a post[-
] tensioning system in the vertical and horizontal directions, and 
the dome roof is pre-stressed utilizing a three way post-tensioning 
system.
    The concrete containment building is required for structural 
integrity of the containment under Design Basis Accident (DBA) 
conditions. The steel liner and its penetrations establish the 
leakage limiting boundary of the containment. Maintaining the 
containment OPERABLE limits the leakage of fission product 
radioactivity from the containment to the environment.
    The integrity of the containment penetrations and isolation 
valves is verified through Type B and Type C local leak rate tests 
(LLRTs) and the overall leak tight integrity of the containment is 
verified by a Type A integrated leak rate test (ILRT) as required by 
10 CFR 50, Appendix J, ``Primary Reactor Containment Leakage Testing 
for Water-Cooled Power Reactors.'' These tests are performed to 
verify the essentially leak tight characteristics of the containment 
at the design basis accident pressure.
    The existing 10-year Type A test interval is based on past 
performance. Previous Type A leakage tests conducted at Braidwood 
Station Units 1 and 2, and Byron Station Units 1 and 2 indicate that 
leakage from containment has been less than the 10 CFR 50 Appendix J 
leakage limit.
    The proposed changes for a one-time extension of the Type A 
tests do not affect the method for Type A, B or C testing or the 
test acceptance criteria. Type B and C testing will continue to be 
performed at the frequency required by the Braidwood Station and 
Byron Station Technical Specifications. The containment inspections 
that are performed in accordance with the requirements of the ASME 
Boiler and Pressure Vessel Code, Section XI and 10 CFR 50.65, 
``Requirements for monitoring the effectiveness of maintenance at 
nuclear power plants,'' provide a high degree of assurance that the 
containment will not degrade in a manner that is only detectable by 
Type A testing.
    In NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' the NRC indicated that a 20-year extension for Type A 
testing resulted in an imperceptible increase in risk to the public. 
The NUREG-1493 study also concluded that, generically, the design 
containment leak rate contributes a very small amount to the 
individual risk [and] have a minimal affect on this risk. EGC has 
conducted risk assessments to determine the impact of a change to 
the Braidwood Station and Byron Station Type A test schedule from a 
baseline value of once in 10 years to once in 15 years for the risk 
measures of Large Early Release Frequency (LERF), Total Population 
Dose, and Conditional Containment Failure Probability (CCFP). The 
results of the risk assessments indicate that the proposed changes 
to the Braidwood Station and Byron Station Type A test schedule has 
a minimal impact on public risk.
    Therefore, based on previous Type A test results for the 
Braidwood Station and Byron Station containments, the current 
containment surveillance programs at each station, and the results 
of the EGC risk assessments, the proposed changes do not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: August 7, 2006, as supplemented by 
letters dated January 22, and May 14, 2007, which included a revised no 
significant hazards consideration determination (NSHCD). This NSHCD is 
from the May 14, 2007, supplement.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station Unit No. 1 (Seabrook) Facility Operating 
License (FOL) and Technical Specifications (TSs). The proposed changes 
would correct a joint-owner name in the operating license, remove a 
license condition from Appendix C to the FOL that is no longer 
applicable, and remove the list of Bases sections from the TS Index. 
Additionally, the proposed amendment would remove two manual valves 
from TS table 3.3.9, ``Remote Shutdown System,'' and add the 
requirement that only one charging pump is permitted to be aligned for 
injection into the reactor coolant system (RCS) in Modes 4, 5, and 6 to 
TS 3.4.9.3, ``Overpressure Protection Systems.'' The additional 
requirement proposed for TS 3.4.9.3 would allow for two pumps to be 
aligned for injection under administrative controls for up to one hour 
to permit swap over operations. The proposed changes would also remove 
a 1-hour reporting requirement for portable makeup pump system storage 
from TS 3.7.4, ``Service Water System/Ultimate Heat Sink,'' correct an 
error in TS 4.7.4.3, related to the service water pumphouse water level 
and delete a footnote from TS 3.7.6.2, ``Air Conditioning,'' that was 
only applicable to Cycle 7. The proposed changes would also delete a 
redundant reporting requirement in TS 6.6, ``Safety Limit Violation.'' 
Lastly, the proposed amendment would modify TS 6.7.6, ``Radioactive 
Effluent Controls Program,'' to clarify the TS with respect to the 
performance of dose projections.
    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. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The probability or consequences of accidents previously 
evaluated in the UFSAR [Updated Final Safety Analysis Report] are 
unaffected by this proposed change. There is no change to any 
equipment response or accident mitigation scenario, and this change

[[Page 31102]]

results in no additional challenges to fission product barrier 
integrity. The proposed change does not alter the design, 
configuration, operation, or function of any plant system, 
structure, or component. As a result, the outcomes of previously 
evaluated accidents are unaffected.
    This change limits availability of the charging pumps to one 
pump when in Mode 4 with the temperature of any RCS cold leg is less 
than or equal to 290 [deg]F, in Mode 5, and in Mode 6 with the 
reactor vessel head on and the vessel head closure bolts not fully 
de-tensioned. Nonetheless, imposing this limitation does not alter 
the configuration or operation of the charging pumps from that 
specified in current administrative controls. Technical 
Specification (TS) 3/4.5.3, ECCS [Emergency Core Cooling System] 
Subsystems--Tavg Less Than 350 [deg]F, presently stipulates that 
only one charging pump is maintained operable in Mode 4. Similarly, 
Technical Requirement 26, Boration Systems, requires that all but 
one operable charging pump be demonstrated inoperable in Modes 4, 5, 
and 6. Also, the Seabrook Station Updated Final Safety Analysis 
Report (UFSAR) describes the configuration of the charging pumps 
during shutdown conditions: Prior to decreasing RCS temperature 
below 350 [deg]F, the safety injection pumps and the non-operating 
charging pumps are made inoperable. Consequently, the change does 
not alter the configuration or operation of the charging pumps from 
the procedures presently described in the UFSAR; rather, it only 
relocates an existing limitation from the UFSAR to the technical 
specifications. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    This proposed change also revises the minimum water level in the 
service water system pump house required for operability of the 
service water system. The value currently specified in the technical 
specifications has been in error since 1986 and will be corrected 
with this change. Increasing the minimum required water level from 
five feet to 25.1 feet does not alter the configuration or operation 
of the service water system. Following discovery of this 
discrepancy, administrative controls established a minimum water 
level of approximately 25 feet. Moreover, monitoring of the service 
water pump house level during 2005 observed that the level, which is 
controlled by the ocean tides, is normally greater than 26 feet. 
During this period the minimum and maximum pump house water levels 
were 26.3 and 48.57 feet, respectively. This administrative change 
has no affect on the actual operation or configuration of the 
service water system. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed revision to TS Table 3.3-9, Remote Shutdown System, 
eliminates valves MS-V127 and MS-V128 from the table. Located in the 
main steam supply line to the turbine-driven emergency feedwater 
(TDEFW) pump, these are locked open, manually operated, valves. 
Supplement 4 of NUREG 0896, Safety Evaluation Report, discusses the 
modifications made to the Emergency Feedwater System (EFW) to 
address problems experienced with the EFW steam supply lines during 
hot functional testing. A design change, installed in 1991, changed 
MS-V127 and MS-V128 to normally open valves, replaced the valves' 
pneumatic actuators with gear-operated manual operators, and re-
assigned the EFW actuation and containment isolation functions of 
these valves to new automatic isolation valves (MS-V393 and MS-V394) 
in the TDEFW pump steam supply line. As a result, the elimination of 
MS-V127 and MS-V128 from TS Table 3.3-9 does not alter the design, 
configuration, operation, or function of these valves with regard to 
operation of the EFW system because in the existing design these 
normally open valves are not required to re-position to support 
operation of the TDEFW pump. Automatic valves MS-V393 and MS-V394, 
which actuate to initiate operation of the TDEFW pump, are 
appropriately under the control of TS Table 3.3-9. This proposed 
change does not alter the design, configuration, operation, or 
function of the EFW steam supply valves. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The other changes in this proposed amendment correct errors, 
remove an outdated license condition, remove an inconsistency 
between indexes, and revise a reporting requirement. These changes 
are administrative in nature and do not impact the design, 
configuration, operation, or function of any plant system, 
structure, or component. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes (1) relocate an existing limitation from 
the UFSAR to the technical specifications regarding availability of 
the charging pumps, (2) revise the minimum water level in the 
service water system pump house required for operability of the 
service water system, (3) eliminate valves MS-V127 and MS-V128 from 
TS Table 3.3-9, and (4) make administrative changes to the TS that 
correct errors, remove an outdated license condition and an 
inconsistency between indexes and revises a reporting requirement. 
No new accident scenarios, failure mechanisms, or limiting single 
failures are introduced as a result of the proposed change. The 
proposed change does not challenge the performance or integrity of 
any safety-related system. The ability of any operable structure, 
system, or component to perform its designated safety function is 
unaffected by this change. The proposed change neither installs or 
removes any plant equipment, nor alters the design, physical 
configuration, or mode of operation of any plant structure, system, 
or component. No physical changes are being made to the plant, so no 
new accident causal mechanisms are being introduced. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
on the availability, operability, or performance of safety-related 
systems and components. The proposed change relocates an existing 
limitation from the UFSAR to the technical specifications regarding 
availability of the charging pumps during operation in Mode 4 with 
the temperature of any RCS cold leg is less than or equal to 290 
[deg]F, in Mode 5, and in Mode 6 with the reactor vessel head on and 
the vessel head closure bolts not fully de-tensioned. Nonetheless, 
imposing this limitation does not alter the configuration or 
operation of the charging pumps from those specified in current 
administrative controls and the UFSAR. The proposed change includes 
revising the minimum water level in the service water system pump 
house required for operability of the service water system. This 
change replaces a non-conservative, incorrect value in the TS with a 
minimum required water level that is consistent with the design 
basis for the system. The elimination of MS-V127 and MS-V128 from TS 
Table 3.3-9 does not alter the design, configuration, operation, or 
function of these valves with regard to operation of the EFW system 
because in the existing design these normally open valves are not 
required to re-position to support operation of the TDEFW pump. 
Automatic valves MS-V393 and MS-V394, which actuate to initiate 
operation of the TDEFW pump, are appropriately under the control of 
TS Table 3.3-9. Last, the proposed amendment makes administrative 
changes to the TS that correct errors, remove an outdated license 
condition and an inconsistency between indexes and revises a 
reporting requirement.
    The proposed changes do not alter the design, configuration, 
operation, or function of any plant system, structure, or component. 
The ability of any operable structure, system, or component to 
perform its designated safety function is unaffected by this change. 
Therefore, the margin of safety as defined in the TS is not reduced 
and the proposed change does not involve a significant reduction in 
a margin of safety.

    Based upon the reasoning presented above 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: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Harold K. Chernoff.
    Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power 
Plant (HBPP), Unit 3 Humboldt County, California.
    Date of amendment request: April 4, 2007.

[[Page 31103]]

    Description of amendment request: The licensee has proposed 
amending the existing license to allow the results of near-term 
surveys, performed on a portion of the plant site, to be included in 
the eventual Final Status Survey (FSS) for license termination.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow survey results for a specific 
area within the licensed site area, performed prior to Humboldt Bay 
Power Plant (HBPP) Unit 3 decommissioning and dismantlement 
activities, to be used in the overall licensed site area Final 
Status Survey (FSS) for license termination. The FSS will be 
performed following completion of HBPP Unit 3 decommissioning and 
dismantlement activities. This proposed change would not change 
plant systems or accident analysis, and as such, would not affect 
initiators of analyzed events or assumed mitigation of accidents. 
Therefore, the proposed change does not increase the probability or 
consequences of an accident previously evaluated.
    (2) Does the change create the possibility of a new or different 
kind of accident from any accident evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant or require existing equipment to be operated in a manner 
different from the present design. Implementation of a cross 
contamination prevention and monitoring plan will be done in 
accordance with plant procedures and licensing bases documents.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident evaluated.
    (3) Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed change has no effect on existing plant equipment, 
operating practices, or safety analysis assumptions. 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: Mr. Antonio Fernandez, Esquire, Pacific Gas 
& Electric Company, Post Office Box 7442, San Francisco, CA 94120.
    NRC Acting Branch Chief: Kristina Banovac.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: April 17, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) and license to establish more 
effective and appropriate action, surveillance, and administrative 
requirements related to ensuring the habitability of the control room 
envelop (CRE) in accordance with Nuclear Regulatory Commission (NRC) 
approved TS Task Force (TSTF) Standard Technical Specification change 
traveler TSTF-448, Revision 3, ``Control Room Habitability.'' The NRC 
staff issued a ``Notice of Availability of Technical Specification 
Improvement to Modify Requirements Regarding Control Room Envelope 
Habitability Using the Consolidated Line Item Improvement Process'' 
associated with TSTF-448, Revision 3, in the Federal Register on 
January 17, 2007 (72 FR 2022). The notice included a model safety 
evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request. In its 
application dated April 17, 2007, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC 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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

[[Page 31104]]

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: April 15, 2007.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications (TSs) and license to establish more 
effective and appropriate action, surveillance, and administrative 
requirements related to ensuring the habitability of the control room 
envelop (CRE) in accordance with Nuclear Regulatory Commission (NRC) 
approved TS Task Force (TSTF) Standard Technical Specification change 
traveler TSTF-448, Revision 3, ``Control Room Habitability.'' The NRC 
staff issued a ``Notice of Availability of Technical Specification 
Improvement to Modify Requirements Regarding Control Room Envelope 
Habitability Using the Consolidated Line Item Improvement Process'' 
associated with TSTF-448, Revision 3, in the Federal Register on 
January 17, 2007 (72 FR 2022). The notice included a model safety 
evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request. In its 
application dated April 15, 2007, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC 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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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 amendment revises 
Technical Specifications (TSs) 3.8.1, ``AC [alternating current] 
Sources--Operating,'' 3.8.4, ``DC [direct current] Sources--
Operating,'' 3.8.5, ``DC Sources--Shutdown,'' 3.8.6, ``Battery Cell 
Parameters,'' 3.8.7, ``Inverters--Operating,'' and 3.8.9, 
``Distribution Systems--Operating.'' This change will also add a new 
Battery Monitoring and Maintenance Program, Section 5.5.2.16. The 
proposed TS changes will provide operational flexibility supported by 
DC electrical subsystem design upgrades that are in progress. These 
upgrades will provide increased capacity batteries, additional battery 
chargers, and the means to cross-connect DC subsystems while meeting 
all design battery loading requirements. With these modifications in 
place, it will be feasible to perform routine surveillances as well as 
battery replacements online.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to Technical Specifications (TS) 3.8.4 and 
3.8.6 would allow extension of the Completion Time (CT) for 
inoperable Direct Current (DC) distribution subsystems to manually 
cross-connect DC distribution buses of the same safety train of the 
operating unit for a period of 30 days. Currently the CT only allows 
for 2 hours to ascertain the source of the problem before a 
controlled shutdown is initiated. Loss of a DC subsystem is not an 
initiator of an event. However, complete loss of a Train A 
(subsystems A and C) or Train B (subsystems B and D) DC system would 
initiate a plant transient/plant trip.
    Operation of a DC Train in cross-connected configuration does 
not affect the quality of DC control and motive power to any system. 
Therefore, allowing the cross-connect of DC distribution systems 
does not significantly increase the probability of an accident 
previously evaluated in Chapter 15 of the Updated Final Safety 
Analysis Report (UFSAR).
    The above conclusion is supported by Probabilistic Risk 
Assessment (PRA) evaluation which encompasses all accidents,

[[Page 31105]]

including UFSAR Chapter 15. The Frequency for Surveillance 
Requirements in TS 3.8.4.3 is changed from 24 months to 30 months. 
San Onofre Nuclear Generating Station (SONGS) experience has 
indicated that there have been no battery failures using the 24-
month test frequency for battery service tests, and extending the 
interval to 30 months is not expected to affect SONGS' capability to 
detect battery health and capacity. Also, the routine test frequency 
of 30 months will better dove-tail with the scheduling of the more 
rigorous 60-month interval battery performance of modified 
performance discharge tests.
    Enhancements from TSTF-360, Rev. 1 and IEEE 450 have been 
incorporated into Limiting Conditions for Operation (LCOs) 3.8.4, 
3.8.5, and 3.8.6. These changes do not impact the probability or 
consequences of an accident previously evaluated.
    Further changes are made of an editorial nature or provide 
clarification only. For example, discussions regarding electrical 
`Trains' and `Subsystems' will be in more conventional terminology. 
LCOs affected by editorial changes include 3.8.1, 3.8.4, 3.8.5, 
3.8.6, 3.8.7, and 3.8.9.
    The changes being proposed in the TS do not affect assumptions 
contained in other safety analyses or the physical design of the 
plant, nor do they affect other Technical Specifications that 
preserve safety analysis assumptions.
    Therefore