Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 13169-13188 [06-2383]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 49 (Tuesday, March 14, 2006)]
[Notices]
[Pages 13169-13188]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-2383]


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NUCLEAR REGULATORY COMMISSION


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

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 February 16, 2006 to March 2, 2006. The last 
biweekly notice was published on February 28, 2006 (71 FR 10071).

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,

[[Page 13170]]

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, Rules and 
Directives 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 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

[[Page 13171]]

the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-
mail to pdr@nrc.gov.

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

    Date of amendments request: February 14, 2006.
    Description of amendments request: The amendments would revise 
Technical Specifications (TS) 3.6.3 to allow a blind flange to be used 
for containment isolation in each of the two flow paths of the 42 inch 
refueling purge valves in Modes 1 through 4 without remaining in TS 
3.6.3 Condition D.
    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 probability of an accident previously evaluated would not be 
affected by the proposed changes to allow the use of blind flanges 
for containment isolation in each of the two 42 inch refueling purge 
valve flow paths. The blind flanges are passive components that 
could not initiate an accident.
    The consequences of an accident previously evaluated would not 
be increased because the blind flanges would provide containment 
isolation assumed in the accident analyses instead of the 42 inch 
refueling purge valves. The blind flanges are passive devices not 
susceptible to an active failure or malfunction that could result in 
a loss of isolation or leakage that exceeds limits assumed in the 
safety analysis. The blind flanges are leak rate tested in 
accordance with the containment leakage rate testing program that is 
required by TS surveillance requirement (SR) 3.6.1.1 and TS 5.5.16. 
The blind flanges are sealed using two separate concentric O-rings 
and are leak rate tested after installation by pressurizing the 
space between the O-rings through a test connection and measuring 
the leakage. In addition, the outboard 42 inch refueling purge valve 
packing leakage is measured by pressurizing the stuffing box through 
the leak off line after flange installation and after any 
maintenance on the packing. The sum of the individual leakage rates 
is compared to the acceptance criteria. The blind flanges are 
verified to be in position at a frequency of 31 days in accordance 
with TS SR 3.6.3.3.
    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 accident from any accident previously evaluated?
    Response: No.
    A new or different kind of accident from any accident previously 
evaluated would not be created by the proposed changes to allow the 
use of blind flanges for containment isolation in each of the two 42 
inch refueling purge valve flow paths. The blind flanges are passive 
components that could not create an 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.
    No margin of safety is affected by the proposed changes to allow 
the use of blind flanges for containment isolation in each of the 
two 42 inch refueling purge valve flow paths. The blind flanges 
would provide containment isolation assumed in the accident analyses 
instead of the 42 inch refueling purge valves. The blind flanges are 
passive devices not susceptible to an active failure or malfunction 
that could result in a loss of isolation or leakage that exceeds 
limits assumed in the safety analysis. The blind flanges are leak 
rate tested in accordance with the containment leakage rate testing 
program that is required by TS SR 3.6.1.1 and TS 5.5.16. The blind 
flanges are leak rate tested after installation by pressurizing the 
space between the O-rings through a test connection and measuring 
the leakage. In addition, the outboard 42 inch refueling purge valve 
packing leakage is measured by pressurizing the stuffing box through 
the leak off line after flange installation and after any 
maintenance on the packing. The sum of the individual leakage rates 
is compared to the acceptance criteria. The blind flanges are 
verified to be in position at a frequency of 31 days in accordance 
with SR 3.6.3.3.
    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 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: January 31, 2006.
    Description of amendment request: The proposed amendment would 
address an inconsistency that was inadvertently introduced during 
conversion to improved technical specifications (TSs) when ``1 per 
room'' replaced ``2'' as the required channels per trip system for the 
reactor water cleanup (RWCU) area ventilation differential 
temperature--high isolation function.
    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 change clarifies the requirement to maintain 
isolation capability for the RWCU Area Ventilation Differential 
Temperature--High isolation instrumentation by addition of a note to 
TS 3.3.6.1 Condition B, modification of TS 3.3.6.1 Surveillance 
Requirements Notes, and by clarifying the number of instruments 
required to be available in TS Table 3.3.6.1-1, ``Primary 
Containment Isolation Instrumentation,'' Function 5.c, by the 
addition of note (d). This ensures, during surveillance testing and 
normal operation, there will always be at least one instrument 
monitoring for a small leak in all RWCU locations. No changes in 
operating practices or physical plant equipment are created as a 
result of 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 type of accident from any accident previously evaluated?
    Response: No.
    The proposed change clarifies the requirement to maintain 
isolation capability for the RWCU Area Ventilation Differential 
Temperature--High isolation instrumentation by addition of a note to 
TS 3.3.6.1 Condition B, modification of TS 3.3.6.1 Surveillance 
Requirements Notes, and by clarifying the number of instruments 
required to be available in TS Table 3.3.6.1-1, ``Primary 
Containment Isolation Instrumentation,'' Function 5.c, by the 
addition of note (d). This ensures, during surveillance testing and 
normal operation, there will always be at least one instrument 
monitoring for a small leak in all RWCU locations. No physical 
change in plant equipment will result from this proposed change. 
Therefore, the proposed change does not create the possibility of a 
new or different type 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 clarifies the requirement to maintain 
isolation capability for the RWCU Area Ventilation Differential 
Temperature--High isolation

[[Page 13172]]

instrumentation by addition of a note to TS 3.3.6.1 Condition B, 
modification of TS 3.3.6.1 Surveillance Requirements Notes, and by 
clarifying the number of instruments required to be available in TS 
Table 3.3.6.1-1, ``Primary Containment Isolation Instrumentation,'' 
Function 5.c, by the addition of note (d). This ensures, during 
surveillance testing and normal operation, there will always be at 
least one instrument monitoring for a small leak in all RWCU 
locations. 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: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Branch Chief: Timothy J. Kobetz, Acting.

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

    Date of amendment request: January 30, 2006.
    Description of amendment request: The license amendment request 
would modify the currently approved radiological accident analyses 
(RAA) and associated Technical Specifications (TS) to account for the 
difference between the control room emergency zone (CREZ) unfiltered 
in-leakage (UFI) assumed in the current RAA and the CREZ UFI that was 
measured during testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. There are no system, structural, or component (SSC) 
alterations due to these changes. The radiological accident analyses 
inputs modified by this request are not accident initiators and do 
not affect the frequency of occurrence of previously analyzed 
transients.
    The radiological accident analyses have demonstrated acceptable 
results using the revised inputs for all affected accidents. 
Further, the proposed changes do not alter or prevent the ability of 
structures, systems or components to perform their intended function 
to mitigate the consequences of accidents previously evaluated in 
the Updated Safety Analysis Report.
    Therefore, the 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?
    No. There are no physical changes to the plant SSCs and there is 
no adverse impact on component or system interactions due to the 
proposed changes. The modes of operation of the plant remain 
unchanged and the design functions of all the safety systems remain 
in compliance with the applicable safety analysis acceptance 
criteria. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The radiological accident analysis inputs modified by this 
request were incorporated into the revised radiological accident 
analyses. The revised radiological analyses satisfy all applicable 
acceptance criteria. There is no adverse effect on plant safety due 
to this proposed license amendment. Therefore, the change does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    Acting NRC Branch Chief: T. Kobetz.

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

    Date of amendment request: February 6, 2006.
    Description of amendment request: The proposed amendment adds a 
license condition to extend certain Technical Specification (TS) 
surveillance test intervals on a one-time basis to account for the 
effects of an extended forced outage in the spring of 2005.
    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 requested action is a one-time extension to the performance 
interval of a limited number of TS surveillance requirements. The 
performance of these surveillances, or the failure to perform these 
surveillances, is not a precursor to an accident. Performing these 
surveillances or failing to perform these surveillances does not 
affect the probability of an accident. Therefore, the proposed delay 
in performance of the surveillance requirements in this amendment 
request does not increase the probability of an accident previously 
evaluated.
    A delay in performing these surveillances does not result in a 
system being unable to perform its required function. In the case of 
this one-time extension request, the relatively short period of 
additional time that the systems and components will be in service 
before the next performance of the surveillance will not affect the 
ability of those systems to operate as designed. Therefore, the 
systems required to mitigate accidents will remain capable of 
performing their required function. No new failure modes have been 
introduced because of this action and the consequences remain 
consistent with previously evaluated accidents. Therefore, the 
proposed delay in performance of the surveillance requirements in 
this amendment request does not involve a significant increase in 
the consequences of an accident.
    Therefore, operation of the facility in accordance with the 
proposed license amendment would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed 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 one-time surveillance requirement deferrals 
being requested.
    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?
    Response: No.
    The proposed amendment is a one-time extension of the 
performance interval of a limited number of TS surveillance 
requirements. Extending these surveillance requirements does not 
involve a modification of any TS Limiting Conditions for Operation. 
Extending these surveillance requirements does not involve a change 
to any limit on accident consequences specified in the license or 
regulations. Extending these surveillance requirements does not 
involve a change to how accidents are mitigated or a significant 
increase in the consequences of an accident. Extending these 
surveillance requirements does not involve a change in a methodology 
used to evaluate consequences of an accident. Extending these 
surveillance requirements does not involve a change in any operating 
procedure or process.

[[Page 13173]]

    The instrumentation and components involved in this request have 
exhibited reliable operation based on the results of the most recent 
performance of their 18-month surveillance requirements.
    Based on the limited additional period of time that the systems 
and components will be in service before the surveillances are next 
performed, as well as the operating experience that these 
surveillances are typically successful when performed, it is 
reasonable to conclude that the margins of safety associated with 
these surveillance requirements will not be affected by the 
requested extension.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    Acting NRC Branch Chief: T. Kobetz.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: December 19, 2005.
    Description of amendment request: The amendment proposes to revise 
the Technical Specifications (TS) to make the temporary changes to TS 
Table 3.3.8.1-1, previously approved by Amendment No. 147, permanent. 
TS Table 3.3.8.1-1 would be revised to delete the temporary note, 
correct the number of Required Channels per Division for the Loss of 
Power (LOP) time delay functions, and delete the requirement to perform 
Surveillance Requirement (SR) 3.3.8.1.2, the monthly Channel Functional 
Test, on certain LOP time delay functions.
    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 regarding the number of required channels 
per division for the LOP time delay functions are administrative in 
nature. The changes do not alter the instrumentation design or their 
physical configuration, and will not affect their operation or 
manner of control. The proposed changes correct an inconsistency 
between a TS Table and the RBS [River Bend Station, Unit 1] design 
basis. The TS required number of voltage sensors per division and 
associated channel components that monitor voltage conditions and 
provide the 4.16 kV bus undervoltage protection are unchanged.
    The exclusion of the time delay functions from the monthly 
Channel Functional Test is proposed because the test creates a loss 
of function for the LOP instrumentation and is, therefore, 
undesirable during unit operations. The test also introduces the 
potential for an unintended plan transient, so the elimination of 
the requirement reduces the potential for such transients.
    The channel functional test will continue to be performed every 
31 days for the sensor channels. In addition, the LOP time delay 
functions will continue to be functionally tested and calibrated 
every 18 months as required by SR 3.3.8.1.3 and SR 3.3.8.1.4. 
Therefore, the required LOP instrumentation will continue to be 
tested in a manner and at a frequency necessary to provide 
confidence that the instrumentation can perform its intended safety 
function.
    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 changes do not alter the instrumentation design or their 
physical configuration, and will not affect their operation or 
manner of control. The proposed TS changes do not introduce any new 
failure mechanisms, malfunctions, or accident initiators not 
considered in the design and licensing bases.
    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 changes have no affect on any safety analysis 
assumptions or methods of performing safety analyses. The changes do 
not adversely affect system OPERABILITY or design requirements and 
the equipment continues to be tested in a manner and at a frequency 
necessary to provide confidence that the equipment can perform its 
intended safety functions. [Regulation] 10 CFR 50.36(c)(3) requires 
the TS to include Surveillance Requirements relating to test, 
calibration, or inspection to assure that the necessary quality of 
systems and components is maintained, that facility operation will 
be within safety limits, and that the limiting conditions for 
operation will be met. The channel functional test will continue to 
be performed every 31 days for the sensor channels. In addition, the 
LOP time delay functions will continue to be functionally tested and 
calibrated every 18 months as required by SR 3.3.8.1.3 and SR 
3.3.8.1.4.
    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: Mark Wetterhahn, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: January 26, 2006.
    Description of amendment request: The proposed amendment will 
modify Technical Specification (TS) requirements to support the 
implementation of Average Power Range Monitor (APRM), Rod Block 
Monitor, TS/Maximum Extended Operating Domain (ARTS/MEOD).
    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?
    The proposed changes revise thermal limit structure employed to 
comply with TS Section 3.2 LCOs [limiting conditions for operation]. 
The proposed changes will replace the flow-biased APRM scram and rod 
block trip setdown requirements with power and flow dependent 
adjustments to the Minimum Critical Power Ratio (MCPR) and Maximum 
Average Planar Linear Heat Generation Rate (MAPLHGR) or Linear Heat 
Generation Rate (LHGR) thermal limits. The adjustments to the 
thermal limits have been determined using NRC approved analytical 
methods as required by Technical Specifications 5.6.5.b and topical 
reports as specified in the Core Operating Limits Report (COLR). The 
proposed changes will not affect any accident initiating mechanism. 
Adjustments to thermal limits will be determined using NRC approved 
methodologies. The power and flow dependent adjustments will ensure 
that the MCPR safety limit will not be violated as a result of any 
anticipated operational occurrence (AOO), that the fuel thermal and 
mechanical design bases will be maintained, and that the 
consequences of the postulated loss of coolant accident (LOCA) will 
remain within acceptable limits. There are no changes to radioactive 
source terms or release pathways. Operation within the expanded 
operating domain has been evaluated and the affect on plant 
accidents was found to be

[[Page 13174]]

within acceptable parameters. The proposed changes do not result in 
any significant change in the availability of logic systems or 
safety-related systems themselves. Required protective functions 
will be maintained. The proposed changes do not degrade plant 
design, operation, or the performance of any safety system assumed 
to function in the accident analysis.
    Therefore, the proposed changes do 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?
    The proposed changes do not introduce any new accident 
initiators or failure mechanisms because the changes and the affects 
on existing structures, systems and components have been evaluated 
and found to not have any adverse affects. The proposed changes 
eliminate the requirement for setdown of the flow-biased APRM scram 
and rod block trip setpoints or APRM adjustments under specified 
conditions and will substitute adjustments to the MCPR and MAPLHGR 
or LHGR thermal limits. Because the thermal limits will continue to 
be met, no transient event will escalate into a new or different 
type of accident due to the initial starting conditions permitted by 
the adjusted thermal limits.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident than those previously evaluated.
    3. [Does the proposed change] involve a significant reduction in 
a margin of safety?
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. There is no affect on the conclusions of 
any safety analysis. Replacement of the APRM setpoint requirement 
with power and flow dependent adjustments to the MCPR and MAPLHGR or 
LHGR thermal limits will continue to ensure that margins to the fuel 
cladding Safety Limit are preserved during operation at other than 
rated conditions. The fuel cladding safety limit will not be 
violated as a result of any anticipated operational occurrence. The 
flow and power dependent adjustments will be determined using NRC 
approved methodologies. The flow and power dependent adjustments 
will also ensure that all fuel thermal-mechanical design bases shall 
remain within the licensing limits. The proposed changes do not 
involve any increase in calculated off-site dose consequences. 
Operability of protective instrumentation and the associated systems 
is assured, and performance of equipment will not be significantly 
affected.
    Therefore, there is no significant reduction in the margin of 
safety as a result of the proposed changes.

    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. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: January 26, 2006.
    Description of amendment request: The proposed license amendment 
replaces the existing Reactor Vessel Material Surveillance Program with 
the Boiling Water Reactor Vessel and Internals Project (BWRVIP) 
Integrated Surveillance Program (ISP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the licensing basis continues to assure 
that applicable regulatory requirements are met and the same 
assurance of reactor pressure vessel integrity continues to be 
provided. The proposed change to the License and licensing basis 
follow the NRC Safety Evaluation approving the implementation of the 
ISP. The proposed change ensures that the reactor pressure vessel 
will continue to be operated within the design, operational, and 
testing limits.
    The proposed change does not modify the reactor coolant pressure 
boundary, (i.e., there are no changes in operating pressure, 
materials, or seismic loading). The proposed change does not 
adversely affect the integrity of the reactor coolant pressure 
boundary such that its function in the control of radiological 
consequences is affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a modification to the 
design of plant structures, systems, or components. Thus, no new 
modes of operation are introduced by the proposed change. The 
proposed change will not create any failure mode not bounded by 
previously evaluated accidents.
    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 implementation of ISP has been previously approved 
by the NRC and found to provide an acceptable alternative to plant-
specific reactor vessel material surveillance programs. Operation of 
JAFNPP within the program ensures that the reactor vessel materials 
will continue to behave in a non-brittle manner, thereby preserving 
the original safety design bases.
    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. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Richard J. Laufer.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: February 15, 2006.
    Description of amendment request: The proposed change will 
specifically credit the measurement tank weir flow instrumentation for 
the containment fan cooler condensate flow monitoring system in place 
of the one containment fan cooler condensate flow switch currently 
required by Technical Specification 3.4.5.1, ``Reactor Coolant System 
Leakage--Leakage Detection Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Reactor Coolant System (RCS) leakage detections systems are 
passive monitoring systems; therefore, the proposed changes do not 
affect reactor operations or accident analyses and have no 
radiological consequences. The change maintains conservative 
restrictions on RCS leakage detections systems consistent with 
Regulatory Guide 1.45 [``Reactor Coolant Pressure Boundary Leakage 
Detection Systems''] and 10 CFR [Part] 50, Appendix A, General 
Design Criteri[on] 30.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 13175]]

    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 introduces no new mode of plant operation or 
any plant modification. The RCS leakage detection instrumentation is 
not part of plant control instruments or engineered safety feature 
actuation circuits but is used solely for monitoring purposes. The 
change does not vary or affect any plant operating condition or 
parameter.
    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.
    There will be no adverse affects on margins of safety since more 
stringent requirements will be applied to the third method (CFC 
[Containment Fan Cooler] condensate flow monitoring) of detecting 
RCS leakage. The third required RCS leakage detection method will 
now be capable of detecting a one gallon per minute leak within one 
hour.
    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: N. S. Reynolds, Esquire, Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

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, Unit Nos. 1 
and 2, Will County, Illinois

    Date of amendment request: October 3, 2005.
    Description of amendment request: The proposed amendments would 
revise the reactor coolant system pressure and temperature limits 
report (PTLR) requirements. Specifically, the amendment would revise 
the TS Section 1.1, ``Definitions,'' description of the PTLR by 
deleting reference to specifications containing limits in the PTLR; (2) 
revise the administrative controls TS 5.6.6, ``Reactor Coolant System 
(RCS) Pressure and Temperature Limits Report (PTLR),'' by requiring the 
NRC approval documents to be identified by date and topical reports to 
be identified by number and title in accordance with Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-419; ``Revise PTLR Definition and References in ISTS 
5.6.6, RC PTLR,'' and (3) add Westinghouse Electric Company, LLC, WCAP-
16143, ``Reactor Vessel Closure Head/Vessel Flange Requirements 
Evaluation for Byron/Braidwood Units 1 and 2,'' to the list of 
analytical methods provided in TS 5.6.6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the definition of PTLR is considered to 
be an editorial change because the requirements of TS 5.6.6 continue 
to specify the Limiting Conditions for Operation that address 
operation within the P-T [pressure temperature] limits.
    The proposed changes to reference only the Topical Report number 
and title do not alter the use of the analytical methods used to 
determine the pressure temperature (P-T) limits or Low Temperature 
Overpressure Protection (LTOP) System setpoints that have been 
reviewed and approved by the NRC. This method of referencing Topical 
Reports would allow the use of current Topical Reports to support 
limits in the PTLR without having to submit an amendment to the 
operating license provided there is no change to the approved 
methodology. TS 5.6.6.b requires that the analytical methods used to 
determine the P-T limits be those previously reviewed and approved 
by the NRC. Implementation of revisions to Topical Reports would 
still be reviewed in accordance with 10 CFR 50.59, ``Changes, tests 
and experiments,'' and where required receive NRC review and 
approval.
    The use of WCAP-16143, following approval by the NRC, for 
generation of P-T limits will continue to ensure that reactor 
pressure vessel integrity is maintained under all conditions.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
The proposed changes do not increase the types or amounts of 
radioactive effluent that may be released offsite, nor significantly 
increase individual or cumulative occupational/public radiation 
exposures. The proposed changes are consistent with safety analysis 
assumptions and resultant consequences.
    Based on the above discussion, the proposed changes do not 
involve an 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 to the definition of PTLR is considered to 
be an editorial change because the requirements of TS 5.6.6 continue 
to specify the Limiting Conditions for Operation that address 
operation within the P-T limits.
    The proposed changes to reference only the Topical Report Number 
and title do not alter the use of the analytical methods used to 
determine the P-T limits or LTOP setpoints that have been reviewed 
and approved by the NRC. This method of referencing Topical Reports 
would allow the use of current Topical Reports to support limits in 
the PTLR without having to submit an amendment to the operating 
license provided there is no change to the approved methodology. TS 
5.6.6.b requires that the analytical methods used to determine the 
P-T limits be those previously reviewed and approved by the NRC. 
Implementation of revisions to Topical Reports would still be 
reviewed in accordance with 10 CFR 50.59 and where required receive 
NRC review and approval.
    The use of WCAP-16143, following approval by the NRC, for 
generation of P-T limits will continue to ensure that reactor 
pressure vessel integrity is maintained under all conditions.
    The proposed changes will allow the use of a new NRC-approved 
methodology for the calculation of P-T limits. However, the changes 
do not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) and do not introduce 
a new mode of plant operation. Safety functions associated with P-T 
limits and LTOP setpoints will continue to function as previously 
assumed in accident analyses.
    Based on this evaluation, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the definition of PTLR is considered to 
be an editorial change because the requirements of TS 5.6.6 continue 
to specify the Limiting Conditions for Operation that address 
operation within the P-T limits. The proposed changes to reference 
only the Topical Report Number and title do not alter the use of the 
analytical methods used to determine the P-T limits or LTOP 
setpoints that have been reviewed and approved by the NRC. This 
method of referencing Topical Reports would allow the use of current 
Topical Reports to support limits in the PTLR without having to 
submit an amendment to the operating license provided there is no 
change to the approved

[[Page 13176]]

methodology. TS 5 .6.6.b requires that the analytical methods used 
to determine the P-T limits be those previously reviewed and 
approved by the NRC. Implementation of revisions to Topical Reports 
would still be reviewed in accordance with 10 CFR 50.59 and where 
required receive NRC review and approval.
    The P-T limits provide assurance that the reactor pressure 
vessel is maintained. The use of WCAP-16143, following approval by 
the NRC, for generation of P-T limits will continue to ensure that 
reactor pressure vessel integrity is maintained under all 
conditions.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. Changes to setpoints at which protective 
actions are initiated that are allowed by the use of WCAP-16143 are 
evaluated in accordance with 10 CFR 50.59 and where required receive 
NRC review and approval. Sufficient equipment remains available to 
actuate upon demand for the purpose of mitigating an analyzed event.
    Based on this evaluation, 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. J. Bradley Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Mindy Landau, Acting.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment requests: January 25, 2006.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 1.1, ``Definitions,'' and TS 
3.4.16, ``RCS Specific Activity.'' The proposed amendments would 
replace the current TS 3.4.16 limit on reactor coolant system (RCS) 
gross specific activity with a new limit on RCS noble gas specific 
activity. The noble gas specific activity limit would be based on a new 
DOSE EQUIVALENT XE-133 definition (corresponding to the Xenon-133 
isotope) that would replace the current--AVERAGE DISINTEGRATION ENERGY 
definition. In addition, the current DOSE EQUIVALENT I-131 definition 
(corresponding to the Iodine-131 isotope) would be revised to allow the 
use of alternate, NRC-approved thyroid dose conversion factors.
    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 to add a new thyroid dose conversion factor 
reference to the definition of DOSE EQUIVALENT I-131, eliminate the 
definition of E--AVERAGE DISINTEGRATION ENERGY, add a new definition 
of DOSE EQUIVALENT XE-133, replace the Technical Specification (TS) 
3.4.16 limit on reactor coolant system (RCS) gross specific activity 
with a limit on noble gas specific activity in the form of a 
Limiting Condition for Operation (LCO) on DOSE EQUIVALENT XE-133, 
replace TS Figure 3.4.16-1 with a maximum limit on DOSE EQUIVALENT 
I-131, extend the Applicability of LCO 3.4.16, and make 
corresponding changes to TS 3.4.16 to reflect all of the above are 
not accident initiators and have no impact on the probability of 
occurrence for any design[-]basis accidents.
    The proposed changes will have no impact on the consequences of 
a design[-basis accident because they will limit the RCS noble gas 
specific activity to be consistent with the values assumed in the 
radiological consequence analyses. The changes will also limit the 
potential RCS iodine concentration excursion to the value currently 
associated with full power operation, which is more restrictive on 
plant operation than the existing allowable RCS iodine specific 
activity at lower power levels.
    Therefore, the proposed changes do 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 accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not alter any physical part of the plant 
nor do they affect any plant operating parameters besides the 
allowable specific activity in the RCS. The changes that impact the 
allowable specific activity in the RCS are consistent with the 
assumptions assumed in the current radiological consequence 
analyses.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The acceptance criteria related to the proposed changes involve 
the allowable control room and offsite radiological consequences 
following a design[-]basis accident. The proposed changes will have 
no impact on the radiological consequences of a design[-]basis 
accident because they will limit the RCS noble gas specific activity 
to be consistent with the values assumed in the radiological 
consequence analyses. The changes will also limit the potential RCS 
iodine specific activity excursion to the value currently associated 
with full power operation, which is more restrictive on plant 
operation than the existing allowable RCS iodine specific activity 
at lower power levels.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of amendment request: October 28, 2005.
    Description of amendment request: The amendment would revise the 
Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications (TS) 
TS 3.8.1 to incorporate changes implementing requirements for an 
Alternate AC (AAC) power supply.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No.
    The proposed change revised two action statements and relocated 
a surveillance requirement. The first AOT [allowable outage time] 
extension permits one EDG [emergency diesel generator] to be 
inoperable for up to 14 days, but the AAC [alternate alternating 
current] source will have to be available. This proposed change will 
be primarily used for scheduled preventative maintenance while the 
plant is online. If used for corrective maintenance, the AAC source 
will have to be capable of providing power within one hour, 
otherwise the existing 72-hour AOT would apply. This assures that 
adequate power remains available to the ESF buses to enable the 
plant to safely shut down, maintain a safe shutdown condition, and/
or mitigate the effects of a design basis accident.
    The second AOT extension provides an additional two hours to 
complete the

[[Page 13177]]

verification of supported equipment for operability. This additional 
time allows for a planned and systematic approach to performing this 
verification. Since there are other more immediate ways for the 
control room staff to be notified of the inoperable status of ESF 
[engineered safety feature] equipment, (annunciators, BISI, status 
lights), the TS requirement is not critical in knowing the status of 
the plant. Should some equipment be discovered inoperable, the 
extended AOT provides for some opportunity to restore the status to 
operable.
    The deletion of a surveillance requirement that requires 
performing a vendor recommended maintenance at a specific frequency 
does not impact the ability of the EDG to perform its intended 
function for the mission time assumed in the accident analysis. EDG 
maintenance will continue to be performed and controlled under 
station procedures. The risk associated with the maintenance will be 
assessed under the provisions of 10 CFR 50.65 [Requirements for 
monitoring the effectiveness of maintenance at nuclear power 
plants], section (a) 4. The TS frequency was initially established 
to coincide with refueling outages, the only time that one EDG could 
be inoperable for any extended time. However, multiple plants have 
extended the time between refueling outages to 24 months with no 
discernable impact on reliability or availability. In addition, the 
Fairbanks-Morse diesel engine owners group has evaluated the 
maintenance requirements and determined that the TS required 
frequency should be based on performance and inspection results, not 
an arbitrary period that coincides with the best opportunity to 
perform the work. The Maintenance Rule requires evaluation for 
additional corrective actions and increased monitoring for scoped 
systems if the reliability and/or availability fall below pre-
established criteria. This approach ensures appropriate actions in a 
timely manner are taken to ensure that equipment relied upon for 
accident mitigation is available when required.
    There are no changes in operational limits or physical design of 
the onsite electric power systems. The proposed changes do not 
change the function or operation of plant equipment or affect the 
response of the equipment if called upon to operate. The EDGs are 
not the initiators of previously evaluated accidents. The EDGs are 
designed to mitigate the consequences of accidents. The risk 
informed assessment that was performed concluded that the increase 
in plant risk is small and consistent with the guidance in 
Regulatory Guide 1.174, [``An Approach for Using Probabilistic Risk 
Assessment in Risk Informed Decisions on Plant-Specific Changes to 
the licensing Basis'']. This assessment considers the possibility of 
an accident occurring during the extended period that the EDG would 
be unavailable. The proposed changes allow for additional 
operational flexibility and will not cause a significant increase in 
the probability or consequences of an accident previously evaluated. 
In actuality, the installation and availability of the AAC will have 
an overall net reduction in core damage frequency.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No.
    The proposed change to extend the EDG AOT to 14 days is based 
upon the installation of an AAC power source and the significant 
reduction in core damage frequency that results. There are no 
significant changes in installed plant equipment or operation of 
safety related equipment. The accident analysis considered the 
credible accidents and bounded those that apply.
    The installation of the AAC and the extended AOT for one EDG to 
be inoperable remain bounded by previous evaluations.
    The AOT extension to provide additional time to perform the 
redundant equipment verification is based on the other methods 
available for the Control Room staff to be made aware of a change in 
ESF equipment status and the safety benef