[Federal Register Volume 71, Number 49 (Tuesday, March 14, 2006)]
[Notices]
[Pages 13169-13188]
From the Federal Register Online via the Government Publishing 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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
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
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact
[[Page 13171]]
the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-
mail to [email protected].
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 benefit of performing this
verification in an unhurried manner. This verification has been
extended by other plants, both those who have converted to ITS and
those that have not. No plant modifications are required and
operator training is unaffected. The verification process does not
utilize any new or complex software and any new accident is bounded
by a Loss of Site Power or Station Blackout analysis.
The deletion of a surveillance requirement to perform the
manufacturer's recommended inspection and maintenance is based on
the recommendations from the vendor and the Fairbanks Morse owners
group. The recommendation is to continue to perform the inspections
and maintenance but the frequency should not be based on the
refueling outage frequency. The effectiveness of the maintenance
will be assured through monitoring under the Maintenance Rule
program which would require evaluation and corrective actions should
the EDG not meet its performance criteria for reliability and
availability.
The EDG performs a function of supplying power when the normal
ESF sources are unavailable. This is a function that mitigates the
effects of the event and the proposed changes cannot cause the
possibility of an accident that was not previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
No.
The proposed change to extend the EDG AOT to 14 days from the
current 72 hours will assure that an alternative source of power for
the ESF onsite distribution system is available and ready. The AAC
and interfacing equipment are designed to maintain independence and
separation, particularly during faulted conditions. The plant
equipment will continue to respond per the design and analysis. The
performance capability of the EDGs will not be affected.
Installation of the AAC will have a net reduction in the core damage
frequency. In addition, administrative controls will ensure that
there are adequate compensatory measures that can and will be taken
during extended EDG maintenance activities to reduce overall risk.
The AOT extension to provide additional time to perform the
redundant equipment verification for operability verification allows
some time to discover a problem and make a minor repair prior to
placing the plant in a shutdown transient. The types of corrective
or preventative maintenance associated with an EDG will not change.
Plant operating and emergency procedures will be enhanced with
guidance on when to use the AAC and how to connect up to the ESF
bus.
The deletion of the periodic EDG inspection per the vendor's
recommendation at a proscribed frequency provides significant
flexibility in when to schedule the inspection and preventative
maintenance. The activities would still be performed but the
frequency would be based on equipment performance and owners group
recommendation. The plant analysis only considers the availability
of the EDG. The TS surveillances that assure the EDG remains
operable remain in place at their current frequencies and the
maintenance requirement will assure that the EDG receives sufficient
maintenance to remain operable.
Since the operation of the plant remains largely unaffected and
the EDG or the AAC will supply power to the ESF equipment as needed,
there is no significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. Hamilton Hagood, Jr., South Carolina
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina
29218.
NRC Section Chief: Evangelos C. Marinos.
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: November 29, 2005.
Description of amendment request: The proposed amendment would add
requirements to TS 3/4.7.1.2 to assure continued operability of the
Emergency Feedwater (EFW) System based on LER 1998-004-00, by including
the newly installed six emergency feedwater system automatic isolation
valves into the Surveillance Requirements to assure the capability for
automatic isolation of EFW in the event of a faulted steam generator.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or
[[Page 13178]]
consequences of an accident previously evaluated?
No.
The proposed change addresses necessary changes to the VCSNS
[Virgil C. Summer Nuclear Station] Technical Specification (TS)
4.7.1.2.b and 4.7.1.2.c.2 associated with the installation of six
new automatic isolation valves in the EF[W] system.
The only Final Safety Analysis Report (FSAR) analyzed accident
for which the EF[W] system could contribute as an initiator would be
minor secondary line break, as described in Section 15.3.2. The
addition of isolation valves in the EF[W] piping to the steam
generators [SGs] will not increase the likelihood of a pipe break,
since the addition will be in accordance with the same codes and
standards as the corresponding, existing portions of the system.
Piping stress analyses have demonstrated the addition of these
valves does not result in the need to postulate any additional pipe
breaks.
The accidents analyzed in the FSAR, which rely on EF[W system]
to mitigate consequences, are loss of normal feedwater, loss of off-
site power, and major secondary system pipe ruptures. The addition
of these automatic isolation valves will eliminate the need for
operator action to manually close a flow control valve in response
to a major secondary system line break. The elimination of operator
manual action is accomplished by the addition of a new pneumatically
operated isolation valve in series with each of the six existing
flow control valves.
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 previously evaluated?
No.
This proposed change does not result in changes to actual
operating pressures, flow rates, flow paths, or system interfaces.
There are no alterations to system operability requirements. The
existing system alarm set points are not affected, neither is the
information available to the operators. The addition of six new
isolation valves will not change system design criteria and the
surveillance testing will be the same as for the existing flow
control valves.
This change does not introduce any new or different kind of
failure mechanisms or limiting single failures. Piping analysis has
concluded that no new pipe break locations or break sizes will
result from this change. Equipment protection features are not
impacted, the frequency of pump and valve operation remains the
same. Independence and redundancy are actually improved.
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 change involve a significant reduction in a
margin of safety?
No.
The design basis for the EF[W] system is to assure the required
flow and pressure to remove decay heat from the core under the worst
postulated conditions. An additional function of the system is to
isolate flow to a faulted SG within the time assumed in the safety
analysis. The proposed change eliminates the need for operators to
take actions to manually close the flow control valves in the event
of a single failure.
The proposed change will create a surveillance requirement for
the new isolation valves that is the same as the existing flow
control valves. The acceptance criteria will assure the operability
of these valves. The design and installation of these isolation
valves will maintain the requirements for independence, redundancy,
separation and testability. The margins assumed in the safety
analysis will be enhanced by this proposed change. Due to the
automatic isolation capability, additional water will be available
for the intact SGs and a reduced mass will be available to be
released into the containment building. No credible single failure
will be capable of preventing isolation of a faulted SG upon a high
flow signal.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. Hagood Hamilton, South Carolina Electric
& Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Branch Chief: Evangelos C. Marinos.
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: November 29, 2005.
Description of amendment request: This amendment revises Technical
Specification (TS) 6.9.1.5 and TS 6.9.1.10 by eliminating the
requirements to submit monthly operating reports and occupational
radiation exposure reports. This consolidated line item improvement
process (CLIIP) TS change was noticed in the Federal Register on June
23, 2004, (69 FR 35067). In addition, the TSs are revised beyond the
scope of the CLIIP by the deletion of the TS 6.9.15 requirement to
report exceedence of coolant specific activity limits and an
administrative change to a TS index page.
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:
SCE&G has reviewed the proposed no significant hazards
consideration determination published on June 23, 2004 (69 FR 35067)
as part of the CLIIP. SCE&G has concluded that the proposed
determination presented in the notice is applicable to the VCSNS,
and the determination is hereby incorporated by reference to satisfy
the requirements of 10 CFR 50.91(a).
The deletion of the additional paragraph in 6.9.1.5 is beyond
the scope of the CLIIP and as such is beyond the scope of the no
significant hazards consideration determination published on June
23, 2004. Therefore the following evaluation has been performed.
In accordance with the criteria set forth in 10 CFR 50.92, SCE&G
has evaluated the proposed beyond scope Technical Specification
change and determined it does not represent a significant hazards
consideration. The following is provided to support this conclusion.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No.
The proposed change is the deletion of a paragraph in the
administrative controls section of the facility Technical
Specifications. The paragraph identifies required information that
was to be provided in a report to the staff in the event where the
RCS specific activity exceeded TS limits. This report has been found
to be un-necessary due to reporting requirements located in 10 CFR
50.73 (exceeding a TS limit). Additionally, the TS limits are set
such that there is very little risk to the health and safety of the
public. Before the condition became significant, the NRC would have
been notified due to the 10 CFR 50.73 requirement to report
significant degradations in a principal fission product barrier.
Deletion of an administrative controls paragraph that provides
reporting requirements is not a precursor to an accident. No changes
are being proposed to any installed plant equipment or procedures.
The operating philosophy is unaffected and training is not impacted.
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?
No.
The proposed change is the deletion of a paragraph that was
inserted per the guidance of Generic Letter 85-19. The staff was
concerned that the reporting requirements prior to that time were
too restrictive and relaxed them through the Generic Letter. Since
that time, it was determined that specific reporting could be
performed via requirements in 10 CFR 50.73. Exceeding the TS limit
is now an uncommon condition as proper fuel management and
fabrication techniques should preclude approaching the TS limit.
Revising or even deleting a reporting requirement in the
facility TS will not impact
[[Page 13179]]
how the plant is operated, how data is evaluated, or what
instructions are located in operating and emergency procedures. No
new equipment is being installed and no plant modifications are
resulting from this proposed change. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
No.
The proposed change to delete some specific reporting
requirements in the Administrative Controls section of TS has no
impact on any plant evaluation or analysis. No plant setpoints are
impacted; no alarm or annunciator functions are affected. This
change has been approved for other plants. 10 CFR 50.73 will still
require reporting the condition should it ever occur. 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: J. Hagood Hamilton, South Carolina Electric
& Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: Evangelos C. Marinos.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: September 19, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Limiting Conditions for Operation
(LCO) 3.3.1, ``Reactor Trip system (RTS) Instrumentation'' and TS
Surveillance Requirements (SR) 3.2.4.2, ``Quadrant Power Tilt Ration
(QPTR)'' to avoid confusion as to when a flux map for QPTR is required.
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 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.
Further, 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. Therefore, the
proposed changes do not increase 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 previously evaluated?
No.
The proposed changes do not result in a change in the manner in
which the RTS and ESFAS provide plant protection. The RTS and ESFAS
will continue to have the same set points after the proposed changes
are implemented. There are no design changes associated with the
license amendment.
The changes do not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. In addition,
the changes do not impose any new or different requirements or
eliminate any existing requirements. The changes do not alter
assumptions made in the safety analysis. The proposed changes are
consistent with the safety analysis assumptions and current plant
operating practice.
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 change involve a significant reduction in a
margin of safety?
No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by these changes. Redundant RTS and ESFAS trains
are maintained, and diversity with regard to the signals that
provide reactor trip and engineered safety features actuation is
also maintained. All signals credited as primary or secondary, and
all operator actions credited in the accident analyses will remain
the same. The proposed changes will not result in plant operation in
a configuration outside the design basis.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: December 16, 2005.
Brief description of amendments: The proposed change would revise
Technical Specifications (TSs) 3.3.2, ``ESFAS [Engineered Safety
Features Actuation System] Instrumentation''; 3.5.2, ``ECCS [Emergency
Core Cooling System]--Operating''; and 3.6.7, ``Spray Additive
System.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
None of the changes impact the initiation or probability of
occurrence of any accident.
The consequences of accidents evaluated in the FSAR [Final
Safety Analysis Report] that could be affected by this proposed
change are those involving the pressurization of the containment and
associated flooding of the containment and recirculation of this
fluid within the ECCS or the Containment Spray System (e.g., LOCAs
[loss-of-coolant accidents]).
Although the water level in the containment flood plain will be
higher at the start of ECCS switchover, the maximum water levels
observed for the duration of the accident are unchanged by the
nominal setpoint changes.
The increase in the minimum water delivered to containment by
the RWST [Refueling Water Storage Tank] setpoint change will reduce
the radiological consequences of LOCAs by diluting the radioiodine
concentrations in the recirculating sump fluid which could be
released by Engineered Safety Features (ESF) leakage. This increase
in water will also reduce the maximum pH and its deleterious effects
on equipment and sump performance.
The increase in water level and the change in strainer design
will significantly increase NPSH [net positive suction head] and
headloss margins required to assure long term core cooling.
The change to a minimum pH of 7.1 will not result in a
significant increase in the radiological consequences of a LOCA as
described below.
[[Page 13180]]
The buffering agent will dissolve in the containment sump fluid
resulting from these accidents raising the pH of the fluid, which
would initially be greater than or equal to 4.0 but less than 7.0
during the injection phase of containment spray operation. The
equilibrium spray pH during the recirculation phase resulting from
this change will be greater than or equal to 7.1. The pH range for
the spray will be bounded by the water spray solution which is
borated water with a maximum of 2600 ppm [parts per million] boron
buffered to a final spray solution pH much less than the 10.5 as
described in the current FSAR Section 3.11(B) for the postulated
spray solution environment. The maximum pH is the limiting parameter
for equipment qualification. Since the resulting pH level will be
closer to neutral using the lower limit of 7.1, post-LOCA corrosion
of containment components will not be increased. Post-LOCA hydrogen
generation will be reduced. There will not be an adverse radiation
dose effect on any safety-related equipment. Thus, the potential for
failures of the ECCS or safety-related equipment following a LOCA
will not be increased as a result of the proposed change.
This modification affects the Containment Spray System which is
intended to respond to and mitigate the effects of a LOCA. The
chemical additive baskets serve a passive function to provide a
buffering agent to neutralize the sump solution. Failure of a basket
would not initiate an accident. The Containment Spray System will
continue to function in a manner consistent with the plant design
basis. There will be no degradation in the performance of nor an
increase in the number of challenges to equipment assumed to
function during an accident situation.
As such, these Technical Specification revisions do not affect
the probability of any event initiators. There will be no adverse
changes to normal plant operating parameters, ESF actuation
setpoints, or accident mitigation capabilities.
The proposed change allows a passive Spray Additive System to
replace the active Spray Additive System currently used to mitigate
the consequences of an accident. By substituting a passive system
for an active system, the probability of occurrence of a malfunction
of equipment associated with the Spray Additive System will be
reduced since the number of active components subject to malfunction
is reduced. This TS surveillance change will maintain the
equilibrium sump pH at greater than or equal to 7.1 to minimize
chloride-induced stress corrosion cracking in austenitic stainless
components important to safety located inside containment.
Therefore, the proposed changes will not increase the probability of
an accident or malfunction of equipment important to safety
previously evaluated in the FSAR.
The offsite and control room doses will continue to meet the
requirements of 10 CFR [Part] 100; 10 CFR [Part] 50, Appendix A, GDC
[General Design Criterion] 19; SRP [Standard Review Plan] 15.6.5.11;
and SRP 6.4.11. The deletion of the active Spray Additive System and
replacement with a sump pH control system using TSP-C [Trisodium
Phosphate crystalline] will not increase the reported radiological
consequences of a postulated LOCA. The proposed new pH control
system will provide satisfactory retention of iodine in the sump
water, as well as provide adequate pH control to minimize the
potential of chloride-induced stress corrosion cracking of
austenitic stainless steel components.
The baskets which will contain the trisodium phosphate are
seismically designed and located in the post-accident flood plane
area to ensure mixing with the recirculating fluid. The consequences
of a malfunction of any piece of equipment associated with the
Containment Spray System would not be affected by the change from an
active Spray Additive System to a passive system. The consequences
of a failure in the active Spray Additive System are eliminated by
this passive system. The proposed changes do not increase the
malfunction of equipment important to safety previously evaluated in
the FSAR. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The changes to the new Containment Spray Additive System are
essentially a passive system, i.e., no operator or automatic action
of electrical devices is required to actuate the system. There are
no electrical components being added whose failure could prevent the
new system from functioning. The only new components being added are
the storage baskets for the chemical buffering agent. Seismic
requirements have been included in the design to ensure the
structural integrity of the baskets will be maintained during a
seismic event.
No new accident scenarios, transient precursors, or limiting
single failures are introduced as a result of these changes. There
will be no adverse effect or challenges imposed on any safety-
related system as a result of these changes. The use of dry sodium
phosphates is allowed for adjustment of the post-LOCA sump solution
pH as discussed in SRP 6.1.1. The quantity of trisodium phosphate or
any other buffering agent chosen will provide a minimum equilibrium
sump pH of 7.1 following dissolution and mixing. Therefore, the
possibility of a new or different type of accident is not created.
There are no changes which would cause the malfunction of
safety-related equipment, assumed to be operable in the accident
analyses, as a result of the proposed Technical Specification
changes. No new equipment performance burdens are imposed; however,
there is the potential for an unlikely, but possible, event in which
an initially concentrated solution of buffering agent could be
transported to the stagnant volume of an inactive sump during
blowdown and pool fill. This situation would be short-lived since,
as the recirculated sump fluid is cooled in the RHR [residual heat
removal] heat exchangers, sufficient buoyancy-driven circulation
within containment will result to displace the stagnant solution and
eventually yield a uniform, equilibrium solution. In the current
design, all of the chemical additive is delivered to the
recirculation sump even in the event of the worst single active
failure. The possibility of a malfunction of safety-related
equipment with a different result is not created. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The RWST Low-Low nominal setpoint, in conjunction with the plant
modifications, ensures that both the ECCS and Containment Spray
Systems can be transferred from injection to recirculation without
stopping the pumps and with no credit for containment overpressure.
Analyses have been performed which show that, even with worst case
single active failures, suction to the pumps would not be lost.
The only function of the NaOH spray additive solution is to
provide pH control of the post-accident containment recirculation
sump water, since the borated water from the Refueling Water Storage
Tank (RWST) used as the containment spray pump suction source during
injection is sufficient to remove iodine from the containment
atmosphere following a LOCA. The net effect on the pH control
function of reducing the amount of NaOH or replacing NaOH with the
chemical buffering agent TSP-C is that the equilibrium sump pH will
be lowered to a minimum of 7.1. There will be no change to the
current Technical Specification acceptance limits on RWST volume and
boron concentration. The resulting equilibrium sump pH level from
this change will be closer to neutral; therefore, the post-LOCA
corrosion of containment components will not be increased.
Because the long term pH will be maintained greater than or
equal to 7.1, margin to minimize the potential for stress corrosion
cracking is maintained.
The radiological analysis as discussed in the technical analysis
above, is shown not to be impacted. There will be no change to the
DNBR [departure from nucleate boiling ratio] Correlation Limit, the
design DNBR limits, or the safety analysis DNBR limits discussed in
Bases Section 2.1.1. There will be no effect on the manner in which
Safety Limits or Limiting Safety System Settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection functions. There will be no adverse
impact on DNBR limits, FQ, F-delta-H, LOCA PCT [peak
cladding temperature], peak local power density, or any other margin
of safety. Therefore the proposed change does not involve a
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
[[Page 13181]]
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: December 16, 2005.
Brief description of amendments: The amendment would revise the
Technical Specifications (TS) to adopt NRC-approved Revision 4 to
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity.'' The proposed amendment includes:
--Revised TS definition of Leakage,
--Revised TS 3.4.13, ``RCS [Reactor Coolant System] Operational
Leakage,''
--Added new TS 3.4.17, ``Steam Generator (SG) Tube Integrity,''
--Revised TS 5.5.9, ``Steam Generator Program''
--Added new TS 5.6.9, ``Steam Generator Tube Inspection Report,'' and
--Revised TS 5.6.10, ``Steam Generator Tube Inspection Report'' (for
existing Unit 1 SGs).
The proposed changes are necessary in order to implement the
guidance for the industry initiative on Nuclear Energy Institute (NEI)
Report 97-06, ``Steam Generator Program Guidelines.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the following
NSHC determination in its application dated December 16, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change requires a SG Program that includes
performance criteria that will provide reasonable assurance that the
SG tubing will retain integrity over the full range of operating
conditions (including startup, operation in the power range, hot
standby, cooldown and all anticipated transients included in the
design specification). The SG performance criteria are based on tube
structural integrity, accident induced leakage, and operational
LEAKAGE.
A SGTR [steam generator tube rupture] event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis of a SGTR event, a bounding primary to
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits
in the licensing basis plus the LEAKAGE rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as a MSLB [main steam line
break], rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TS identify
the standards against which tube integrity is to be measured.
Meeting the performance criteria provides reasonable assurance that
the SG tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design basis accident. The performance criteria are only a part of
the SG Program required by the proposed change to the TS. The
program, defined by NEI 97-06, Steam Generator Program Guidelines,
includes a framework that incorporates a balance of prevention,
inspection, evaluation, repair, and leakage monitoring. The proposed
changes do not, therefore, significantly increase the probability of
an accident previously evaluated.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in the plant technical specifications
for operational leakage and for DOSE EQUIVALENT I-131 in primary
coolant to ensure the plant is operated within its analyzed
condition. The analysis of the limiting design basis accident
assumes that primary to secondary leak rate after the accident is 1
gallon per minute with no more than 150 gallons per day in any one
SG, and that the reactor coolant activity levels of DOSE EQUIVALENT
I-131 are at the TS values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection,
[[Page 13182]]
assessment, repair, and plugging. The requirements established by
the SG Program are consistent with those in the applicable design
codes and standards and are an improvement over the requirements in
the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff proposes to determine that the amendments request
involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: December 16, 2005.
Brief description of amendments: The proposed amendments would
revise the Technical Specifications (TSs) consistent with the Nuclear
Regulatory Commission (NRC)-approved Technical Specification Task Force
(TSTF) Standard Technical Specification Change Traveller, TSTF-419,
``Revise PTLR [Pressure and Temperature Limits Report] Definition and
References in ISTS [improved Standard 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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to reference the Topical Report number and
title do not alter the use of the analytical methods used to
determine the P/T [Pressure/Temperature] limits or LTOP [Low
Temperature Overpressure Protection] 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. 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 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. Further, 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.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to reference 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. 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 changes do not involve a
physical alteration of the plant (i.e., no new or different type of
equipment will be installed) or a change in the methods governing
normal plant operation. In addition, the changes do not impose any
new or different requirements or eliminate any existing
requirements. The changes do not alter assumptions made in the
safety analysis. The proposed changes are consistent with the safety
analysis assumptions and current plant operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes to reference 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. 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 proposed changes do not alter
the manner in which safety limits, limiting safety system settings
or limiting conditions for operation are determined. The setpoints
at which protective actions are initiated are not altered by the
proposed changes. Sufficient equipment remains available to actuate
upon demand for the purpose of mitigating an analyzed event.
Therefore, it is concluded that this 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: David Terao.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: January 31, 2006.
Description of amendment request: The proposed change would replace
the current containment methodology with the methodology described in
Topical Report DOM-NAF-3, ``GOTHIC Methodology for Analyzing the
Response to Postulated Pipe Ruptures Inside Containment,'' increase the
containment air partial pressure limits in Technical Specification (TS)
3.8, ``Containment,'' revise the loss-of-coolant (LOCA) accident
alternate source term (AST) analysis, and change the method of starting
the recirculation spray (RS) pumps.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No.
The proposed changes include a physical alteration to the RS
system to start the inside and outside RS pumps on RWST [Refueling
Water Storage Tank] Level Low coincident with CLS [consequence
limiting safeguards] High High containment pressure. The RS system
is used for accident mitigation only, and changes in the operation
of the RS system cannot have an impact on the probability of an
accident. The other changes do not affect equipment and are not
accident initiators. The RWST Level Low instrumentation will comply
with all applicable regulatory requirements and design criteria
(e.g., train separation, redundancy, single failure). Therefore, the
design functions performed by the RS system are not changed.
Delaying the start of the RS pumps affects long-term containment
pressure and
[[Page 13183]]
temperature profiles. The environmental qualification of safety-
related equipment inside containment was confirmed to be acceptable,
and accident mitigation systems will continue to operate within
design temperatures and pressures. Delaying the RS pump start
reduces the emergency diesel generator loading early during a design
basis accident, and staggering the RS pump start avoids overloading
on each emergency bus. The reduction in iodine removal efficiency
during the delay period is offset by changes to other assumptions in
the LOCA dose analysis. The net impact is a reduction in the
predicted offsite doses and control room doses following a design
basis LOCA.
The UFSAR [Updated Final Safety Analysis Report] safety analysis
acceptance criteria continue to be met for the proposed changes to
the RS pump start method, the proposed TS containment air partial
pressure limits, the implementation of the GOTHIC containment
analysis methodology, and the changes to the LOCA dose consequences
analyses. Based on this discussion, the proposed amendments do not
increase the probability or consequence of an accident previously
evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
identified?
No.
The proposed change alters the RS pump circuitry by initiating
the start sequence with a new RWST Level Low signal instead of a
timer after the CLS High High pressure setpoint is reached. The
timers for the outside RS pumps will be used to sequence pump starts
and preclude diesel generator overloading. The RS pump function is
not changed. The RWST Level Low instrumentation will be included as
part of the engineered safeguards features (ESF) instrumentation in
the Surry TS and will be subject to the ESF surveillance
requirements. The design of the RWST Level Low instrumentation
complies with all applicable regulatory requirements and design
criteria. The failure modes have been analyzed to ensure that the
RWST Level Low circuitry can withstand a single active failure
without affecting the RS system design functions. The RS system is
an accident mitigation system only, so no new accident initiators
are created.
The remaining changes to the containment analysis methodology,
the containment air partial pressures, and the LOCA AST analysis
basis do not impact plant equipment design or function. Together,
the changes assure that there is adequate margin available to meet
the safety analysis criteria and that dose consequences are within
regulatory limits. The proposed changes do not introduce failure
modes, accident initiators, or malfunctions that would cause a new
or different kind of accident. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously identified.
3. Does the proposed license amendment involve a significant
reduction in a margin of safety?
No.
The changes to the actuation of the RS pumps and the increased
containment air partial pressure affect the containment response
analyses and the LOCA dose analysis. Analyses have been performed
that show the containment design basis limits are satisfied and the
post-LOCA offsite and control room doses meet the required criteria
for the proposed changes to the containment analysis methodology,
the RS pump start method, the TS containment air partial pressure
limits, and the LOCA AST bases. Therefore, the proposed amendment
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: February 25, 2005.
Brief description of amendment: The amendment deleted Section 2.E
of the Facility Operating License, which requires reporting of
violations of the requirements in Section 2.C of the Facility Operating
License.
Date of Issuance: February 22, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 258.
Facility Operating License No. DPR-16: The amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21453).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 22, 2006.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: March 4, 2005, as supplemented
by letter dated January 25, 2006.
Brief description of amendments: The proposed amendments deleted
Section 2.F (2.G in Unit 3) of the Facility Operating Licenses, which
requires reporting violations of the requirements
[[Page 13184]]
in Section 2.C of the Facility Operating License. The amendments also
make administrative and editorial changes to the Technical
Specifications (TSs). Changes to TS 1.4, ``Frequency,'' and TS 3.4.3,
``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,''
correct editorial errors. The changes to TS 2.1.1, ``Reactor Core SLs
[Safety Limits],'' and TS 3.3.1, ``Reactor Protective System (RPS)
Instrumentation--Operating,'' remove the reference to departure from
nucleate boiling ratios (DNBR) based on operating cycle, since only one
of the listed DNBR values is now valid. TS 3.1.10, ``Special Test
Exceptions (STE)--MODES 1 and 2,'' is changed to correct an
inconsistency between the limiting condition for operation and the TS
Bases. The changes to TS 3.7.2, ``Main Steam Isolation Valves
(MSIVs),'' and TS 3.7.3, ``Main Feedwater Isolation Valves (MFIVs),''
correct the applicability for these specifications. The change to TS
3.8.1, ``AC [Alternating Current] Sources--Operating,'' adds a note to
a surveillance requirement. Changes to TS 3.8.4, ``DC [Direct Current]
Sources--Operating,'' and TS 3.8.6, ``Battery Cell Parameters,'' remove
the reference to AT&T batteries. The changes to TS 5.5.9, ``Steam
Generator (SG) Tube Surveillance Program,'' correct the reference for
NRC notification.
Date of issuance: February 28, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days of the date of issuance.
Amendment Nos.: Unit 1--158, Unit 2 --158, Unit 3--158.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Facility Operating Licenses and the Technical
Specifications.
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24647).
The January 25, 2006, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 28, 2006.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: October 31, 2005.
Brief description of amendment: The amendment modified requirements
by adding to the technical specifications a Limiting Condition for
Operation (LCO) 3.0.8 that provides a delay time for entering a
supported system TS when the inoperability is due solely to an
inoperable snubber, if risk is assessed and managed. In addition, a
change to LCO 3.0.1 was required to reference the addition of LCO
3.0.8.
Date of issuance: February 15, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 172.
Facility Operating License No. NPF-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72670).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 15, 2006.
No significant hazards consideration comments received: No.
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: March 8, 2005, as supplemented by letter
dated January 17, 2006.
Brief description of amendment: The amendment allows a one-time
extension of an additional 4 months beyond the 5-year extension already
granted by the staff to the nominal 10-year interval of the test
interval for the next Appendix J, Type A test.
Date of issuance: February 9, 2006.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 150.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15942). The supplement dated January 17, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne
County, Mississippi
Date of application for amendment: March 30, 2005, as supplemented
by letter dated November 21, 2005.
Brief description of amendment: The amendment incorporated the
following U.S. Nuclear Regulatory Commission (NRC)-approved Technical
Specification Task Force (TSTF) changes that apply to the Boiling Water
Reactor/6 Improved Standard Technical Specifications into GGNS
Technical Specifications (TSs):
------------------------------------------------------------------------
TS section
TSTF No. Description affected
------------------------------------------------------------------------
TSTF-046, Rev. 1.............. Clarify the Surveillance
Containment Isolation Requirement
Valve surveillance to (SR) 3.6.1.3.4,
apply only to SR 3.6.4.2.2,
automatic isolation SR 3.6.5.3.3.
valves.
TSTF-222, Rev. 1.............. Control Rod Scram Time SR 3.1.4.1, SR
Testing. 3.1.4.4.
TSTF-264, Rev. 0.............. Delete flux monitors SR 3.3.1.1.5, SR
specific overlap SRs. 3.3.1.1.6,
Table 3.3.1.1-
1.
TSTF-275, Rev. 0.............. Clarify requirements Table 3.3.5.1-1,
for Diesel Generator Footnote (a).
(DG) start signal on
Reactor Pressure
Vessel (RPV) Level--
Low, Low, Low during
RPV cavity flood-up.
TSTF-276, Rev. 2.............. Revise DG full load SR 3.8.1.9, SR
rejection test. 3.8.1.10, SR
3.8.1.14.
TSTF-300, Rev. 0.............. Eliminate DG Loss of SR 3.8.2.1.
Coolant Accident
(LOCA) Start SRs
while in shutdown
when Emergency Core
Cooling System is not
required.
TSTF-322, Rev. 2.............. Secondary Containment SR 3.6.4.1.3, SR
Integrity SRs. 3.6.4.1.4.
TSTF-400, Rev. 1.............. Clarify SR on bypass SR 3.8.1.13.
of DG automatic trips.
[[Page 13185]]
TSTF-416, Rev. 0.............. SR 3.5.1.2 Notation... Limiting
Condition for
Operation (LCO)
3.5.1, SR
3.5.1.2, LCO
3.5.2, SR
3.5.2.4.
------------------------------------------------------------------------
The amendment also granted delayed performance of the modified SRs
for DG 12 until the next regularly scheduled performance rather than
immediately upon implementation of this amendment, which is still
consistent with NRC-approved TSTF changes. Those SRs are SR 3.8.1.9, SR
3.8.1.10, and SR 3.8.1.14.
Date of issuance: February 2, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance, with the exception of SR 3.8.1.9, SR
3.8.1.10, and SR 3.8.1.14.
Amendment No: 169.
Facility Operating License No. NPF-29: The amendment revises the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29791). The supplemental letter dated November 21, 2005, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 2, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237, Dresden Nuclear
Power Station, Unit 2, Grundy County, Illinois
Date of application for amendment: February 25, 2005.
Brief description of amendment: The amendment deleted the reporting
requirement in the Renewed Facility Operating License related to
reporting violations of other requirements in the operating license.
Date of issuance: February 17, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 210.
Facility Operating License No. DPR-19: The amendments revised the
Facility Operating License.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21456).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 17, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: April 13, 2005, as supplemented
by letter dated December 22, 2005.
Brief description of amendments: The amendment extended the
completion time (CT) for Required Action A.1, ``Restore Residual Heat
Removal Service Water subsystem to OPERABLE status,'' associated with
Technical Specification (TS) Section 3.7.1 from 7 days to 10 days;
established a 6-day (for Division 2 core standby cooling system (CSCS)
maintenance) or 10-day (for Division 1 CSCS maintenance) CT for TS
Section 3.7.2 when one or more required diesel generator cooling water
subsystem(s) are inoperable. The Nuclear Regulatory Commission (NRC)
staff is granting this amendment request with respect to TS Sections
3.7.1 and 3.7.2 only. In the original submittal, the licensee also
requested an extension of the CT for required Action C.4, ``Restore
required Diesel Generator (DG) to OPERABLE status,'' associated with TS
3.8.1 from 72 hours to 6 days; and extension of the CT for required
Action F.1, ``Restore one required Diesel Generator (DG) to OPERABLE
status,'' associated with TS 3.8.1 from 2 hours to 6 days. The NRC
staff needs additional information from the licensee in order to
complete its review and grant this portion of the amendment request.
The staff will address the requests to extend CTs for TS 3.8.1 in a
separate safety evaluation and license amendment, if granted.
Date of issuance: February 23, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 175/161
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33213).
The December 22, 2005, supplement, contained clarifying information
and did not change the NRC staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 23, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of application for amendment: December 14, 2005, as
supplemented by letter dated February 13, 2006.
Brief description of amendment: The amendment modifies the
Technical Specifications (TSs) to incorporate a revised Single Loop
Operation Safety Limit Minimum Critical Power Ratio due to the cycle-
specific analysis.
Date of issuance: March 1, 2006.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 183.
Facility Operating License No. NPF-39 This amendment revised the
TSs.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2590). The supplement dated February 13, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 1, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: December 17, 2004.
Brief description of amendments: The amendments revised Appendix B,
Environmental Protection Plan (non-radiological), of the Limerick
Generating Station Operating Licenses.
Date of issuance: February 17, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 180 and 142.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
[[Page 13186]]
revised the Environmental Protection Plan.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19112).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 17, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: February 25, 2005.
Brief description of amendments: The proposed amendment would
delete the sections of the Facility Operating Licenses that require
reporting of violations of the requirements in Section 2.C of the
Facility Operating Licenses.
Date of issuance: February 17, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 181 and 143.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications/license.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21457).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 17, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Unit Nos. 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: December 21, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) by relocating the Pressure Isolation
Valve Table to the Technical Requirements Manual.
Date of issuance: February 17, 2006.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment Nos.: 182 and 144.
Facility Operating License Nos. NPF-39 and NPF-85. These amendments
revised the TSs.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2590).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 17, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: June 2, 2004, as supplemented
February 11, May 12, October 31, and November 14, 2005.
Brief description of amendments: These amendments approve
conversion of the BVPS-1 and 2 containments from subatmospheric to
atmospheric operating conditions. The proposed changes also approves
the Modular Accident Analysis Program--Design Basis Accident (MAAP-DBA)
computer code for the BVPS-1 and 2 containment integrity analysis and
changes to mass and energy calculation methodologies.
Date of issuance: February 6, 2006.
Effective date: For BVPS-1, the amendment is effective as of the
date of its issuance and shall be implemented prior to Mode 4 entry
during startup from 1R17 which begins on or about February 10, 2006.
For BVPS-2, the amendment is effective as of the date of its issuance
and shall be implemented prior to Mode 4 entry during startup from 2R12
which begins October 2006.
Amendment Nos.: 272 and 154.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 2004 (69 FR
43462).
The supplements dated February 11, May 12, October 31, and November
14, 2005, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 6, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: February 11, 2005, as
supplemented August 8, 2005.
Brief description of amendments: The amendments approved the
adoption of the Relaxed axial offset control (RAOC) and FQ
surveillance methodologies in accordance with NRC-approved Topical
Report WCAP-10216-P-A, ``Relaxation of Constant Axial Offset Control--
FQ Surveillance Technical Specification.'' TS 3.2.1, ``Axial
Flux Difference (AFD),'' and TS 3.2.2, ``Heat Flux Hot Channel Factor--
FQ(Z),'' were revised to adopt the RAOC calculational
procedure of NUREG-1431, ``Standard Westinghouse Technical
Specifications for Westinghouse Plants,'' Revision 3, June 2004.
Changes to TS 3.2.3, ``Nuclear Enthalpy Hot Channel Factor--
FN[Delta]H,'' TS 3.2.4, ``Quadrant Power Tilt
Ratio (QPTR),'' TS 3.3.1, ``Reactor Trip System Instrumentation (Table
4.3-1, Note 3),'' and TS 6.9.5, ``Core Operating Limits Report
(COLR),'' were made to provide consistency with the changes made to TSs
3.2.1 and 3.2.2.
Date of issuance: February 27, 2006.
Effective date: Prior to entry into Mode 4 upon restart from the
spring 2006 refueling outage which begins on or about February 10,
2006, for BVPS-1 and prior to entry into Mode 4 from startup following
the fall 2006 refueling outage which begins in October 2006, for BVPS-
2.
Amendment Nos.: 274 and 155.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21457). The supplement dated August 8, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 27, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: April 20, 2005.
Brief description of amendment: The changes revised the Technical
Specifications (TSs) to replace plant-specific position titles with
generic position titles. Also, the changes deleted TS 6.7, ``Safety
Limit Violations or Protective Limit Violation,'' and included a change
to TS 2.1.2, ``Reactor Core,'' associated with the deletion of TS 6.7.
Additionally, the changes relocated to the Davis-Besse Nuclear Power
Station Updated Safety Analysis Report the Process Control Program
[[Page 13187]]
requirements from TS 6.8, ``Procedures and Programs,'' and from TS
6.14, ``Process Control Program (PCP).'' Associated with this change,
TS Definition 1.30, ``Process Control Program,'' was deleted. Also, TS
6.15, ``Offsite Dose Calculation Manual (ODCM),'' was modified to
eliminate the requirement that changes to the ODCM be reviewed and
accepted by the Plant Operations Review Committee (PORC). These changes
to administrative requirements also eliminated the need to propose
additional changes in the future to plant-specific position/
organizational titles. The changes are consistent with NUREG-1430,
``Standard Technical Specifications--Babcock and Wilcox Plants,''
Revision 3, dated June 2004. Lastly, the changes revised in the TSs the
title ``Industrial Security Plan'' to ``Physical Security Plan.''
Date of issuance: February 7, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 272.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29795).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 7, 2006.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 28, 2005.
Description of amendment request: The amendment revised the
Seabrook Station, Unit No. 1, Technical Specifications (TSs)
Surveillance Requirement 4.1.1.3, ``Moderator Temperature
Coefficient,'' to allow the option of not measuring the moderator
temperature coefficient within 7 effective full-power days of reaching
an equilibrium boron concentration of 300 parts per million. This
option is available only if the conditions described in WCAP-13749-P-A,
``Safety Evaluation Supporting the Conditional Exemption of the Most
Negative Moderator Temperature Coefficient Measurement'' have been met.
Date of issuance: February 17, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 107.
Facility Operating License No. NPF-86: The amendment revised the
TSs.
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24652).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 17, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-388, Susquehanna Steam Electric
Station, Unit 2 (SSES-2), Luzerne County, Pennsylvania
Date of application for amendment: January 28, 2005.
Brief description of amendment: The amendment revises the SSES-2
Technical Specification (TS) Table 3.3.5.1-1, ``Emergency Core Cooling
System Instrumentation,'' Function 3.e, `` High Pressure Coolant
Injection (HPCI) System,'' to change Condition ``D'' to ``C'' as the
condition to reference from Required Action A.1. This is an editorial
revision to correct a typographical error that had been present since
the conversion to the Improved TSs in July 1998.
Date of issuance: February 6, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 206.
Facility Operating License No. NPF-22: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24654).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 2, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: February 7, 2005
Brief description of amendments: The amendments change the SSES 1
and 2 Technical Specifications (TSs) for ``Secondary Containment,''
limiting condition for operation 3.6.4.1, by revising the frequency
note applicable to Surveillance Requirements (SR) 3.6.4.1.4 and SR
3.6.41.5. The revised note requires each zone configuration be tested
at least once every 60 months.
Date of issuance: February 2, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 90 days.
Amendment Nos.: 229 and 205.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29799).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 2, 2006.
No significant hazards consideration comments received: No.
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 application for amendments: April 26, 2004, as supplemented
by letters dated September 16, 2004, September 23, 2004, February 25,
2005, and June 13, 2005.
Brief description of amendments: These amendments revised the
Technical Specifications to incorporate a full-scope application of an
alternate source term methodology in accordance with 10 CFR 50.67.
Date of issuance: February 17, 2006.
Effective date: As of the date of issuance, to be implemented with
90 days.
Amendment Nos.: 271 and 252.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34705). The supplements did not effect the scope of changes discussed
in the original no significant hazards determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 17, 2006.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: April 29, 2005, as supplemented
on September 19, 2005.
Brief description of amendment: The amendment revised the Technical
Specifications to incorporate the relaxed axial offset control and heat
flux hot channel (FQ) surveillance methodologies. These methodologies
are used to reduce operator action required to maintain conformance
with power distribution control requirements and to increase the
ability to return to power after a plant trip or transient. The changes
are consistent with Westinghouse Electric Company Report WCAP-10216-P-
A, ``Relaxation of Constant Axial Offset Control/FQ
Surveillance Technical Specification.''
[[Page 13188]]
Date of issuance: February 15, 2006.
Effective date: As of the date of issuance to be implemented prior
to startup following the fall 2006 refueling outage.
Amendment No.: 94.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33220).
The September 19, 2005, letter provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 15, 2006.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: July 15, 2005, and as
supplemented by letter dated January 20, 2006.
Brief description of amendments: The amendments are for the San
Onofre Nuclear Generating Station (SONGS), Units 2 and 3, operating
licenses, but they involved Unit 1, which is not an operating nuclear
plant and is in the process of being decommissioned. The amendments
revised License Condition 2.B.(6) for both SONGS, Units 2 and 3, by (1)
deleting the sentence ``Transshipment of Unit 1 fuel between Units 1
and [2 or 3] shall be in accordance with SCE [Southern California
Edison Company] letters to U.S. Nuclear Regulatory Commission dated
March 11, March 18 and March 23, 1988, and in accordance with the
Quality Assurance requirements of 10 CFR Part 71'' and (2) adding the
phrase ``and by the decommissioning of San Onofre Nuclear Generating
Station Unit 1'' to the remaining sentence in the license condition.
This change recognized that Unit 1 is now in the stage of
decommissioning and that in the future any radioactive waste water
produced in the further decommissioning of Unit 1 would be released
from the San Onofre site by transferring the waste water from Unit 1 to
Units 2 and 3. The processing (if required) and discharging of this
waste water would be using the Units 2 and 3 radioactive waste system
and ocean outfall discharge line.
Date of issuance: February 28, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2--202; Unit 3--193.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: September 13, 2005 (70
FR 54089).
The supplement dated January 20, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 28, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: November 2, 2005.
Brief Description of amendments: The amendments modify technical
specifications (TS) to adopt the provisions of Industry/TS Task Force
(TSTF) change TSTF-359, ``Increased Flexibility in Mode Restraints.''
Date of issuance: February 22, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 170 and 163.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75498).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 22, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 7th day of March, 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-2383 Filed 3-13-06; 8:45 am]
BILLING CODE 7590-01-P