[Federal Register Volume 71, Number 176 (Tuesday, September 12, 2006)]
[Notices]
[Pages 53715-53726]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-14938]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 18, 2006 to August 31, 2006. The last
biweekly notice was published on August 29, 2006 (71 FR 51222).
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
[[Page 53716]]
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, 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-
[[Page 53717]]
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 the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3 New London County, Connecticut
Date of amendment request: June 14, 2006.
Description of amendment request: The proposed amendment will
permit Millstone Power Station, Unit 3 a one-time, 5-year extension, to
Type A testing, of a surveillance requirement referenced in Technical
Specification (TS) 4.6.1, relevant to the containment structure. TS
4.6.1 specifies performance of an integrated leak rate test at a
frequency of up to 10 years with allowance for a 15-month extension.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1:
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed extension to Type A testing cannot increase the
probability of an accident previously evaluated since extension of
the containment Type A testing is not a physical plant modification
that could alter the probability of accident occurrence nor, is it
an activity or modification that by itself could lead to equipment
failure or accident initiation.
The proposed one-time, five-year extension to Type A testing
does not result in a significant increase in the consequences of an
accident as documented in NUREG-1493. The NUREG notes that very few
potential containment leakage paths are not identified by Type B and
C tests. It concludes that even reducing the Type A (ILRT) testing
frequency to once per twenty years leads to an imperceptible
increase in risk.
DNC provides a high degree of assurance through indirect testing
and inspection that the containment will not degrade in a manner
detectable only by Type A testing. The last two Type A tests
identified containment leakage within acceptance criteria,
indicating a very leak-tight containment. Inspections required by
the ASME Code [American Society of Mechanical Engineers Boiler and
Pressure Vessel Code] are also performed in order to identify
indications of containment degradation that could affect leak-
tightness. Separately, Type B and C testing required by Technical
Specifications, identifies any containment opening from design
penetrations, such as valves, that would otherwise be detected by a
Type A test. These factors establish that a one-time, five-year
extension to the Millstone Power Station Unit 3 Type A test interval
will not represent a significant increase in the consequences of an
accident.
Criterion 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 revision to Technical Specifications adds a one-
time extension to the current interval for Type A testing for
Millstone Power Station Unit 3. The current test interval of ten
years, based on past performance, would be extended on a one-time
basis to fifteen years from the last Type A test. The proposed
extension to Type A testing does not create the possibility of a new
or different type of accident since there are no physical changes
being made to the plant and there are no changes to the operation of
the plant that could introduce a new failure.
Criterion 3:
Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The proposed revision to Millstone Power Station Unit 3
Technical Specifications adds a one-time extension to the current
interval for Type A testing. The current test interval of ten years,
based on past performance, would be extended on a one-time basis to
fifteen years from the last Type A test for Millstone Power Station
Unit 3. RG [Regulatory Guide] 1.174 provides guidance for
determining the risk impact of plant-specific changes to the
licensing basis. RG 1.174 defines very small changes in risk as
resulting in increases of CDF [core damage frequency] below
10-6/yr and increases in LERF [large early release
frequency] below 10-7/yr. Since the ILRT [integrated leak
rate testing] does not impact CDF, the relevant criterion is LERF.
The increase in LERF resulting from a change in the Type A ILRT test
interval from a once-per-ten-years to a once-per-fifteen-years is
3.1 x 10-7/yr, based on internal events. RG 1.174 states
that when the calculated increase in LERF is in the range of
10-7/yr to 10-6/yr, applications will be
considered if it can be shown that the total [LERF] is less than
10-5/yr. Since the total LERF for the 15-year metric is
6.3 x 10-7/yr, then the change is considered acceptable.
Increasing the ILRT interval from ten to fifteen years is,
therefore, considered non-risk significant and will not
significantly reduce the margin of safety. The NUREG-1493 generic
study of the effects of extending containment leakage testing found
that a 20-year interval in Type A leakage testing resulted in an
imperceptible increase in risk to the public. NUREG-1493 generically
concludes that the design containment leakage rate contributes about
0.1 percent of the overall risk. Decreasing the Type A testing
frequency would have a minimal [e]ffect on this risk since 95% of
the Type A detectable leakage paths would already be detected by
Type B and C testing.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Acting Branch Chief: Brooke D. Poole.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: June 21, 2006.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.7.3, Action a, to extend the
allowed outage time (AOT) for one inoperable intake cooling water (ICW)
pump from 7 days to 14 days. The proposed amendments were prepared in
accordance with the guidance provided by the NRC in Regulatory Guide
1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk
Informed Decisions on Plant-Specific Changes to the Licensing Basis''
and Regulatory Guide 1.177, ``An Approach for Plant-Specific, Risk-
Informed Decisionmaking; Technical Specifications.''
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
[[Page 53718]]
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 affects the AOT for TS 3.7.3,
Action a. The proposed change allows an extension of the current AOT
for an inoperable ICW pump from 7 days to 14 days. The proposed
change does not affect the design of the ICW System, the operational
characteristics or function of the ICW System, the interfaces
between the ICW System and other plant systems, or significantly
affect the reliability of the ICW System. Limiting conditions for
operation and their associated allowed outage times are not
considered initiating conditions for any accident previously
evaluated, nor is the ICW System considered an initiator for any
accident previously evaluated. The ICW System provides the cooling
water to the safety related CCW [component cooling water] heat
exchangers. The ICW System also provides cooling water to the TPCW
[turbine plant cooling water] heat exchangers and supplies water to
the Lube Water System. During accident conditions, the ICW System
performs the accident mitigation function of removing the heat load
from the CCW System to support both reactor heat removal and
containment heat removal requirements. The consequences of accidents
previously evaluated are not affected by the proposed change in AOT.
To fully evaluate the effect of the proposed ICW AOT extension, PRA
[probabilistic risk assessment] methods and a deterministic analysis
were utilized. The results of the analysis show no significant
increase in Core Damage Frequency or Large Early Release Frequency
based upon the guidance provided in Regulatory Guides 1.174 and
1.177.
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 probability of a new or
different accident from any accident previously evaluated?
Response: No. The proposed change does not involve a change in
the design, configuration, or method of operation of the plant. The
proposed change will not alter the manner in which equipment
operation is initiated, nor will the functional demands on credited
equipment be changed. The proposed change allows operation of a
Turkey Point unit to continue while an ICW pump is repaired and
tested. The proposed extension does not affect the interaction of an
ICW pump with any system whose failure or malfunction can initiate
an accident. As such, no new failure modes are being introduced.
Therefore, the proposed action 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 does not alter the plant
design, nor does it affect the assumptions contained in the safety
analyses. Specifically, there are no changes being made to the ICW
design, including instrument setpoints. The proposed change has been
evaluated both deterministically, and using risk-informed methods.
Based upon these evaluations, margins of safety ascribed to ICW
availability and to plant risk have been determined to not be
significantly reduced. The evaluation has concluded the following
with respect to the proposed change:
Applicable regulatory requirements will continue to be met,
adequate defense-in-depth will be maintained, sufficient safety
margins will be maintained, and any increases in CDF [core damage
frequency] and LERF [large early release frequency] are small and
consistent with the NRC Safety Goal Policy Statement (Federal
Register, Vol. 5.1, P. 30028 (51 FR 30028), August 4, 1986) as
interpreted by NRC Regulatory Guides 1.174 and 1.177. Furthermore,
increases in risk posed by potential combinations of equipment out
of service during the proposed extended ICW pump AOT will be managed
under a configuration risk management program consistent with 10 CFR
50.65, ``Requirements for monitoring the effectiveness of
maintenance at nuclear power plants,'' paragraph (a)(4).
The availability of the other ICW pumps and the use of on-line
risk assessment tools provide adequate compensation for the
potential small incremental increase in plant risk associated with
the extended ICW pump AOT.
Therefore, the proposed change does not involve a significant
reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: July 6, 2006.
Description of amendment request: The proposed amendments would
incorporate new large-break loss-of-coolant accident (LBLOCA) analyses
using the realistic LBLOCA methodology in the NRC-approved WCAP-16009-
P-A, ``Realistic Large Break LOCA [loss-of-coolant-accident] Evaluation
Methodology Using Automated Statistical Treatment of Uncertainty Method
(ASTRUM),'' and would revise Technical Specification (TS) 5.6.5.b to
include reference to WCAP-16009-P-A.
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.
This license amendment request proposes to incorporate large
break loss of coolant accident analyses using the ASTRUM
methodology, documented in WCAP-16009-P-A, ``Realistic Large Break
LOCA Evaluation Methodology Using the Automated Statistical
Treatment of Uncertainty Method (ASTRUM)'', in the Prairie Island
Nuclear Generating Plant licensing basis and add reference to WCAP-
16009-P-A in the Technical Specification's list of approved
methodologies for establishing core operating limits.
Accident analyses are not accident initiators, therefore, this
proposed licensing basis change does not involve a significant
increase in the probability of an accident. The analyses using
ASTRUM demonstrated that the acceptance criteria in 10 CFR 50.46,
``Acceptance criteria for emergency core cooling systems for light-
water nuclear power reactors'', were met. The NRC has approved WCAP-
16009-P-A for application to two-loop Westinghouse plants with upper
plenum injection. Since the Prairie Island Nuclear Generating Plant
is a two-loop Westinghouse plants with upper head injection and the
analysis results meet the 10 CFR 50.46 acceptance criteria, this
change does not involve a significant increase in the consequences
of an accident.
Addition of the reference to WCAP-16009-P-A in the Technical
Specifications is an administrative change that does not affect the
probability or consequences of an accident previously evaluated.
The changes proposed in this license amendment 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.
This license amendment request proposes to incorporate large
break loss of coolant accident analyses using the ASTRUM
methodology, documented in WCAP-16009-P-A, ``Realistic Large Break
LOCA Evaluation Methodology Using the Automated Statistical
Treatment of Uncertainty Method (ASTRUM)'', in the Prairie Island
Nuclear Generating Plant licensing basis and add reference to WCAP-
16009-P-A in the Technical Specification's list of approved
methodologies for establishing core operating limits.
There are no physical changes being made to the plant as a
result of using the Westinghouse ASTRUM analysis methodology in
WCAP-16009-P-A for performance of the large break loss of coolant
[[Page 53719]]
accident analyses. No new modes of plant operation are being
introduced. The configuration, operation and accident response of
the structures or components are unchanged by utilization of the new
analysis methodology. Analyses of transient events have confirmed
that no transient event results in a new sequence of events that
could lead to a new accident scenario. The parameters assumed in the
analysis are within the design limits of existing plant equipment.
In addition, employing the Westinghouse ASTRUM large break loss
of coolant accident analysis methodology does not create any new
failure modes that could lead to a different kind of accident. The
design of all systems remains unchanged and no new equipment or
systems have been installed which could potentially introduce new
failure modes or accident sequences. No changes have been made to
any reactor protection system or emergency safeguards features
instrumentation actuation setpoints.
Based on this review, it is concluded that no new accident
scenarios, failure mechanisms or limiting single failures are
introduced as a result of the proposed methodology changes.
Addition of the reference to WCAP-16009-P-A in the Technical
Specifications is an administrative change that does not create the
possibility of a new or different kind of accident.
The licensing basis and Technical Specification changes proposed
in this license amendment do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
This license amendment request proposes to incorporate large
break loss of coolant accident analyses using the ASTRUM
methodology, documented in WCAP-16009-P-A, ``Realistic Large Break
LOCA Evaluation Methodology Using the Automated Statistical
Treatment of Uncertainty Method (ASTRUM)'', in the Prairie Island
Nuclear Generating Plant licensing basis and add reference to WCAP-
16009-P-A in the Technical Specification's list of approved
methodologies for establishing core operating limits.
The analyses using ASTRUM demonstrated that the applicable
acceptance criteria in 10 CFR 50.46, ``Acceptance criteria for
emergency core cooling systems for light-water nuclear power
reactors'' are met. Margins of safety for large break loss of
coolant accidents include quantitative limits for fuel performance
established in 10 CFR 50.46. These acceptance criteria and the
associated margins of safety are not being changed by this proposed
new methodology. The NRC has approved WCAP-16009-P-A for application
to two-loop Westinghouse plants with upper head injection. Since the
Prairie Island Nuclear Generating Plant is a two-loop Westinghouse
plants with upper plenum injection and the analysis results meet the
10 CFR 50.46 acceptance criteria, this change does not involve a
significant reduction in a margin of safety.
Addition of the reference to WCAP-16009-P-A in the Technical
Specifications is an administrative change that does not involve a
significant reduction in a margin of safety.
Addition of the reference to WCAP-16009-P-A in the Technical
Specifications is an administrative change that does not involve a
significant reduction in a margin of safety.
The licensing basis and Technical Specification changes proposed
in this license amendment 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Martin Murphy.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: August 11, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to relocate response time
limit tables for the reactor trip system and engineered safety features
actuation system to the Updated Final Safety Analysis Report.
The August 11, 2006, application supersedes the previous
application related to relocation of response time limits, dated August
19, 2005, which was noticed in the Federal Register on December 20,
2005 (70 FR 75496). Instead of changing the response time definitions
in TSs 1.12 and 1.26, as proposed in the August 19, 2005, application,
the licensee proposes to revise certain TS Bases to clarify that
Nuclear Regulatory Commission approval would be required to use a means
other than testing to verify that response times are within limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment relocates the instrument response time
limits for the reactor trip system (RTS) and engineered safety
feature actuation system (ESFAS) from the technical specifications
to the Updated Final Safety Analysis Report (UFSAR). The proposed
amendment conforms to the guidance given in Enclosures 1 and 2 of
Generic Letter 93-08. Neither the response time limits nor the
surveillance requirements for performing response time testing will
be altered by this submittal. The overall RTS and ESFAS functional
capabilities will not be changed and assurance that action
requirements of the reactor trip and engineered safety features
systems are completed within the time limits assumed in the accident
analyses is unaffected by the proposed amendment.
Therefore, operation of the facility in accordance with the
proposed amendment will 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 amendment will not change the physical plant or the
modes of plant operation defined in the operating license. The
change does not involve the addition or modification of equipment
nor does it alter the design or operation of plant systems.
Therefore, operation of the facility in accordance with the
proposed amendment will 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 measurement of instrumentation response times at the
frequencies specified in the technical specification provides
assurance that actions associated with the reactor trip and
engineered safety features are accomplished within the time limits
assumed in the accident analyses. The response time limits and the
measurement frequencies remain unchanged by the proposed amendment.
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.
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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Acting Branch Chief: Brooke D. Poole.
[[Page 53720]]
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: July 14, 2006.
Description of amendment requests: This amendment application
proposes to delete duplicative notifications, reporting, and restart
requirements if a safety limit is violated; replace plant-specific
position titles with generic position titles; and make several
additional administrative changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
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 remove the duplicative safety limit
reporting requirements from the TSs [Technical Specifications] does
not affect the plant or operation of the plant. The change simply
removes duplicative information from the TSs that is covered in the
NRC regulations. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes to make plant-specific position/
organizational titles more generic do not affect any plant
structures, systems, and components, and have no effect on plant
operations. The proposed changes are administrative and do not
affect any existing limits. Accident initial conditions,
probability, and assumptions remain as previously analyzed. The
proposed changes will have no effect on accident initiation
frequency. The proposed changes do not invalidate the assumptions
used in evaluating the radiological consequences of any accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The remaining changes are administrative and do not modify the
qualifications, responsibilities, or requirements for the positions.
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?
Response: No.
The proposed change to remove the duplicative safety limit
reporting requirements from the TSs does not introduce any new
accident scenarios, failure mechanisms, or limiting single failures.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed change has no adverse effect
on any safety-related system or component and does not challenge the
performance or integrity of any safety related system. This change
is considered an administrative action to remove duplicative
reporting requirements. Therefore, this proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
The proposed changes to make plant-specific position/
organizational titles more generic are administrative and do not
introduce any new or different accident initiators. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
The remaining proposed changes are administrative and do not
modify the qualifications, responsibilities, or requirements for the
positions. 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?
Response: No.
The proposed changes are administrative and do not involve any
reduction in a margin of safety. Removal of duplicative information,
replacing plant-specific position titles with generic position
titles, and the other proposed administrative changes do not affect
compliance with the regulations. 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: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: David Terao.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: July 14, 2006.
Description of amendment requests: The proposed change incorporates
a description of the parent tube inspection limitation adjacent to the
nickel band portion of the lower sleeve joint and provides the basis
for the structural and leakage integrity of the joint being ensured
with the existing inspection of the parent tube adjacent to the nickel
band region.
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.
This proposed change revises the San Onofre Units 2 and 3
Technical Specifications (TS) Section 5.5.2.11.f.1.j to provide a
description of the parent tube inspection limitation adjacent to the
nickel band and to provide the basis for the structural and leakage
integrity. This is supported by Westinghouse Topical Report SG-SGDA-
05-48-P Revision 1, ``WOG [Westinghouse Owners Group] PA-MSC-0190,
Revision 1: Test Results Related to TIG [tungsten inert gas] and
Alloy 800 Sleeve Installation in \3/4\ Inch and \7/8\ Inch OD SG
[steam generator] Tubing In-Service Inspection Requirements.''
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.
Steam generator tube leakage and structural integrity will be
maintained during all plant conditions upon implementation of the
proposed inspection scope and repair limit changes to the San Onofre
Units 2 and 3 Technical Specifications. This change does not
introduce any new mechanisms that might result in a different kind
of accident from those previously evaluated.
Therefore, 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.
Structural and leakage integrity of the steam generator sleeve
joint is ensured with the existing inspection of the parent tube
adjacent to the nickel band region.
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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: David Terao.
[[Page 53721]]
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: February 28, 2006.
Brief description of amendments: This request proposes changes to
Technical Specification (TS) 3/4.8.2.1, ``DC Sources--Operating,'' and
3/4.8.2.2, ``DC Sources--Shutdown,'' and the addition of a new TS 3/
4.8.2.3, ``Battery Parameters.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[1.] The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change rearranges the Technical Specifications for
the direct current electrical power system, and adds new Conditions
and required actions with revised completion times to allow for
battery charger inoperability. Neither the direct current electrical
power subsystem nor associated battery chargers are initiators of an
accident sequence previously evaluated. Performance of plant
operational activities in accordance with the proposed Technical
Specification changes ensures that the direct current electrical
power subsystem is capable of performing its function as previously
described. Therefore, the mitigating functions supported by the
subject subsystem will continue to provide the protection assumed by
the safety analysis.
Relocation of preventive maintenance surveillances and certain
operating limits and actions to a ``Battery Monitoring and
Maintenance Program'' will not challenge the ability of the subject
subsystem to perform its design function. Maintenance and monitoring
currently required will continue to be performed. In addition, the
direct current electrical power subsystem is within the scope of 10
CFR 50.65, ``Requirements for monitoring the effectiveness of
maintenance at nuclear power plants,'' which will ensure continued
control of maintenance activities associated with the subject
subsystem.
Revision of battery performance test interval to 12 months from
18 months in 4.8.2.1.f (now 4.8.2.3.f.1) is a conservative change
that is intended to ensure continued battery operability. In
addition, a surveillance requirement will be added as 4.8.2.3.f.2 to
require performance discharge tests at least once per 24 months for
any battery reaching 85% of the service life expected for the
application and capacity is equal to or greater than 100% of the
manufacturer's rating. Surveillance requirement 4.8.2.3.f.2 is an
additional criterion that supplements 4.8.2.3.f.1. Modified
performance tests of batteries that have reached 85% of their
service life are to be performed at 12-month intervals with capacity
less [than] 100% of the manufacturer's rating, and at 24-month
intervals if the capacity is 100% or greater. These surveillance
requirements are consistent with the requirements of IEEE-450.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[2.] The proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
The proposed change does not involve any physical alteration of
the units. No new equipment is introduced, and installed equipment
is not operated in a new or different manner. The proposed changes
do not affect setpoints for initiation of protective or mitigating
actions.
Revision of battery performance test interval to 12 months from
18 months in 4.8.2.1.f (now 4.8.2.3.f.1) is a conservative change
that is intended to ensure continued battery operability. In
addition, a surveillance requirement will be added as 4.8.2.3.f.2 to
require performance discharge tests at least once per 24 months for
any battery reaching 85% of the service life expected for the
application and capacity is equal to or greater than 100% of the
manufacturer's rating. Surveillance requirement 4.8.2.3.f.2 is an
additional criterion that supplements 4.8.2.3.f.1. Modified
performance tests of batteries that have reached 85% of their
service life are to be performed at 12-month intervals with capacity
less [than] 100% of the manufacturer's rating, and at 24-month
intervals if the capacity is 100% or greater. These surveillance
requirements are consistent with the requirements of IEEE-450.
Operability of the DC [direct currrent] electrical power
subsystems in accordance with the proposed technical specifications
is consistent with the initial assumptions of the accident analyses
and is based upon meeting the design basis of the plant.
The proposed changes will not alter the manner in which
equipment operation is initiated, nor will the functional demands on
credited equipment be changed. No alteration in the operating
procedures is proposed, and no change is being made to procedures
relied upon in response to an off-normal event. No new failure modes
are being introduced, and the proposed change does not alter
assumptions made in the safety analyses.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
[3.] The proposed change does not involve a significant
reduction in the margin of safety.
The proposed change will not adversely affect operation of plant
equipment and will not result in a change to the setpoints at which
protective actions are initiated. Sufficient DC capacity to support
operation of mitigation equipment is ensured. The provisions of the
Battery Monitoring and Maintenance Program will ensure that the
station batteries are maintained in a highly reliable manner.
Revision of battery performance test interval to 12 months from
18 months in 4.8.2.1.f (now 4.8.2.3.f.1) is a conservative change
that is intended to ensure continued battery operability. In
addition, a surveillance requirement will be added as 4.8.2.3.f.2 to
require performance discharge tests at least once per 24 months for
any battery reaching 85% of the service life expected for the
application and capacity is equal to or greater than 100% of the
manufacturer's rating. Surveillance requirement 4.8.2.3.f.2 is an
additional criterion that supplements 4.8.2.3.f.1. Modified
performance tests of batteries that have reached 85% of their
service life are to be performed at 12-month intervals with capacity
less [than] 100% of the manufacturer's rating, and at 24-month
intervals if the capacity is 100% or greater. These surveillance
requirements are consistent with the requirements of IEEE-450.
The equipment fed by the DC electrical system will continue to
provide adequate power to safety-related loads in accordance with
analysis assumptions.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: David Terao.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: June 7, 2006.
Brief description of amendments: The proposed change would revise
the Spent Fuel Pool (SFP) and In-Containment Storage Area Criticality
Analysis as described in Section 5.6 of the Technical Specifications
(TSs).
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.
There is no increase in the probability of an accident. The
proposed change does allow a greater number of fuel storage
configurations in SFP. While this could increase the probability of
a fuel misloading, the presence of sufficient soluble boron in
[[Page 53722]]
the SFP precludes criticality as a result of the misloading. Fuel
assembly placement will continue to be controlled pursuant to
approved fuel handling procedures and will be in accordance with the
TS and the spent fuel rack storage configuration limitations of
UFSAR [updated final safety analysis report] Chapter 9.1.2.
Reactivity changes due to SFP temperature changes have been
evaluated. The base case criticality analysis covers a ``normal''
SFP temperature range of 50 [deg]F to 160 [deg]F. Spent fuel pool
temperature accidents are considered outside the normal temperature
range extending from 50 [deg]F to 240 [deg]F. In all SFP temperature
accident cases, sufficient reactivity margin is available to the
0.95 keff limit without requiring additional soluble
boron above the base case level. Because adequate soluble boron will
be maintained in the SFP water to maintain keff < 0.95,
the consequences of a loss of normal cooling to the SFP will not be
increased.
There is no increase in the consequences of the accidental
misloading of spent fuel assemblies into the SFP racks. The
criticality analysis demonstrates that the pool keff will
remain <= 0.95 following an accidental misloading due to the boron
concentration of the pool. The current TS limitation will ensure
that an adequate SFP boron concentration is maintained.
The criticality analysis shows the consequences of a fuel
assembly drop accident in the SFP are not affected when considering
the presence of soluble boron. The rack keff remains <=
0.95.
The editorial changes proposed in this license amendment request
do not impact the probability or consequences of an accident.
Therefore, based on the conclusions of the above evaluation, 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?
Response: No.
Spent fuel handling accidents are not new or different types of
accidents; they have been analyzed in Section 15.7.4 of the UFSAR.
Criticality accidents in the SFP are not new or different types
of accidents. They have been analyzed in the UFSAR and in
Criticality Analysis Reports associated with specific licensing
amendments for fuel enrichments that are assumed for the proposed
change. Because the proposed SFP storage configuration limitations
will be similar to the current ones, the new limitations will not
have any significant effect on normal SFP operations and
maintenance, and will not create any possibility of a new or
different kind of accident. Verifications will continue to be
performed to ensure that the SFP loading configuration meets
specified requirements.
The misloading of a fuel assembly in the required storage
configuration has been evaluated. In all cases, the rack
keff remains <= 0.95. Removal of an RCCA [rod cluster
control assembly] from a checkerboard storage configuration has been
analyzed and found to be bounded by the misloading of a fuel
assembly.
As discussed above, the proposed changes will not create the
possibility of a new or different kind of accident. There is no
significant change in plant configuration, equipment design, or
equipment.
The editorial changes proposed in this license amendment request
do not impact the design basis accidents of STP [South Texas
Project].
Under the proposed amendment, no changes are being made to the
racks themselves, to any other systems, or to the physical
structures of the Fuel Handling Building.
Therefore, 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 TS changes and the resulting spent fuel storage
operation limits will provide [an] adequate safety margin to ensure
that the stored fuel assembly array always remains subcritical.
Those limits are based on a plant-specific criticality analysis
performed in accordance with Westinghouse spent fuel rack
criticality analysis methodology.
While the criticality analysis utilized credit for soluble
boron, storage configurations have been defined using 95/95
keff calculations to ensure that the spent fuel rack
keff is < 1.0 with no soluble boron. Soluble boron credit
is used to offset uncertainties, tolerances, and off-normal
conditions, and to provide subcritical margin such that the SFP
keff is maintained <= 0.95.
The loss of substantial amounts of soluble boron from the SFP
that could lead to keff exceeding 0.95 has been
previously evaluated and approved (Ref. 4 and 5) and shown to be not
credible. A safety evaluation has been performed which shows that
dilution of the SFP boron concentration from 2500 ppm [part per
million] to 700 ppm is not credible. Also, the spent fuel rack
keff will remain < 1.0 (with a 95/95 confidence level)
with the SFP flooded with unborated water. These safety analyses
demonstrate a level of safety comparable to the conservative
criticality analysis methodology required by Westinghouse WCAP-
14416-P-A.
The editorial changes proposed in this license amendment request
do not affect the margin of safety.
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: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station (CPSES), Units 1 and 2, Somervell County, Texas
Date of amendment request: March 31, 2006.
Brief description of amendments: The amendments requested would
revise Technical Specifications (TS) requirement 5.0, ``ADMINISTRATIVE
CONTROLS.''
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 change involves organizational changes at the
executive level and does not impact nor effect accident analysis
assumptions. The method and tools used to maintain, and produce
proposed changes to, the Technical Specifications has no bearing on
any accident analysis assumptions. Therefore, these assumptions are
preserved and there is no change in the probability or consequences
of any previously evaluated accident.
Therefore, the proposed change does 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 change involves an organizational change due to a
change in title. There are no changes in existing reporting
relationships or assigned responsibilities for safe operation of
CPSES. The proposed re-issuance of the entire Technical
Specifications stems from a change in the software utilized by TXU
Power to produce and maintain the Technical Specifications. This
software is not used to operate the plant nor is it used to
establish any operational limits.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. The proposed change will not effect the normal method of
plant operation. No performance requirements will be affected or
eliminated. The proposed change will not result in physical
alteration to any plant system nor will there be any change in the
method by which any safety-related plant system performs its safety
function.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change. There will be no adverse effect or challenges
imposed as a result of this
[[Page 53723]]
change. There will be no adverse effect or challenges imposed on any
safety-related system as a result of these changes.
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 do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). There will be no effect on the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions.
Therefore, the proposed change does not involve a reduction in a
margin of safety. q
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.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
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: July 20, 2006.
Brief description of amendment request: The proposed amendment
would revise the Vogtle Electric Generating Plant (VEGP), Units 1 and
2, Technical Specifications (TS) 5.5.9, ``Steam Generator (SG) Tube
Surveillance Program,'' to incorporate changes in the SG inspection
scope for VEGP, Unit 1 during Refueling Outage 13 and the subsequent
operating cycle, and VEGP Unit 2 during Refueling Outage 12 and the
subsequent operating cycle. The proposed changes modify the inspection
requirements for portions of SG tubes within the tubesheet region of
the SGs.
Date of publication of individual notice in Federal Register: July
31, 2006 (71 FR 43225).
Expiration date of individual notice: 30-day, August 30, 2006; 60-
day, September 29, 2006.
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, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: February 2, 2005, as
supplemented by letters dated April 19, April 21, and June 13, 2006.
Brief description of amendment: The amendment revised the Oyster
Creek Nuclear Generating Station Technical Specifications (TSs) to
incorporate the isolation trip setting and the instrumentation
surveillance requirements of the reactor water clean-up system high
energy line break detection and isolation equipment.
Date of Issuance: August 25, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 259.
Facility Operating License No. DPR-16: The amendment revised the
TSs.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12744). The April 19, April 21, and June 13, 2006, letters provided
clarifying information within the scope of the original application and
did not change the staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated August 25, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: December 14, 2004.
Brief description of amendment: The amendment deleted redundant
administrative responsibilities, changed certain administrative titles
and included editorial corrections and clarifications.
Date of issuance: August 9, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 223.
Facility Operating License No. DPR-35: The amendment revised the
Facility
[[Page 53724]]
Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9990)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 9, 2006.
No significant hazards consideration comments received: No
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 15, 2005, as supplemented by
letters dated March 22, June 2, and July 12, 2006.
Brief description of amendment: The amendment modified Technical
Specification (TS) 6.5.9, ``Steam Generator (SG) Program,'' and TS
6.9.1.5, ``Steam Generator Tube Inspection Report,'' to eliminate the
need to inspect a portion of the tube within the SG tubesheet region,
thereby potentially allowing flaws to remain in the uninspected region.
Date of issuance: August 29, 2006.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 207.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications and the Facility Operating License.
Date of initial notice in Federal Register: June 21, 2005 (70 FR
35737). The March 22, June 2, and July 12, 2006, supplemental letters
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 August 29, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 25, 2005.
Brief description of amendment: The amendment modified the
Surveillance Requirements related to Waterford 3 Technical
Specification 3.1.1.3, ``Moderator Temperature Coefficient,'' to permit
use of the Startup Test Activity Reduction Program (WCAP-16011-P-A).
Date of issuance: August 29, 2006.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 206.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications and the Facility Operating License.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72673). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 29, 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 27, 2004, as
supplemented by letters dated October 25, 2004, October 10, 2005, April
27, May 30, June 16, and August 4, 2006.
Brief description of amendments: This amendment incorporated a
revision to the Technical Specifications (TSs) and licensing and design
bases that supports a full-scope application of an Alternative Source
Term methodology.
Date of issuance: August 23, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos. 185, 146.
Facility Operating License Nos. NPF-39 and NPF-85. This amendment
revised the TSs.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34700). The supplements provided clarifying information that did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as originally published in the Federal
Register. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 23, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania
Date of application for amendment: October 14, 2005, as
supplemented March 31, 2006.
Brief description of amendment: The amendment revised Technical
Specifications 3/4 8.2.3 and 3/4 8.2.4 to permit implementation of
design changes associated with a battery charger upgrade during the
fall 2006 refueling outage.
Date of issuance: August 28, 2006.
Effective date: As of the date of its issuance and shall be
implemented within 90 days.
Amendment No: 157.
Facility Operating License No. NPF-73. Amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: November 22, 2005 (70
FR 70642). The supplement dated March 31, 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 August 28, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: February 16, 2006, as
supplemented by letters dated May 11 and July 13, 2006.
Brief description of amendments: The amendments revised the
Technical Specification (TS) requirements related to steam generator
tube integrity consistent with NRC-approved Revision 4 to TS Task Force
(TSTF) Standard Technical Specification Change Traveler TSTF-449,
``Steam Generator Tube Integrity.''
Date of issuance: August 22, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 223 and 229.
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 11, 2006 (71 FR
18374). The supplements dated May 11 and July 13, 2006, 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 August 22, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: August 11, 2005, as revised by letter
dated November 8, 2005, as supplemented by letter dated April 12, 2006.
[[Page 53725]]
Brief description of amendment: The amendment revised TS 4.2.1,
``Fuel Assemblies,'' to permit the use of AREVA (Framatome ANP) M5
advanced alloy for fuel rod cladding and structural components such as
guide tubes, intermediate spacer grids, end plugs, and guide thimble
tubes at the Fort Calhoun Station, Unit 1 (FCS). M5 will be used
beginning with Refueling Cycle 24. The M5 cladding is a proprietary
zirconium-based alloy that is chemically different from that of
zircaloy and ZIRLO, the fuel cladding materials currently approved for
use in the FCS TS. In addition, TS 5.9, ``Reporting Requirements,'' was
revised to include the Framatome ANP topical report evaluating the
impact of M5 material properties on NRC-approved methodologies used at
the FCS.
Date of issuance: August 30, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 241.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72675). The April 12, 2006, supplemental 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.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated August 30, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: February 23, 2006.
Brief description of amendment: The amendment revised Paragraph
2.C.(6) of the facility operating license to clarify that the license
condition that limits the number of fuel assemblies that can be outside
of approved shipping containers, fuel storage racks, or the reactor
does not apply to fuel assemblies stored in approved dry spent fuel
storage systems.
Date of issuance: August 28, 2006.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 169.
Facility Operating License No. NPF-57: This amendment revised
Paragraph 2.C.(6).
Date of initial notice in Federal Register: May 9, 2006 (71 FR
27003).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 28, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Dates of application for amendments: March 29, 2006, as
supplemented on June 5, 2006.
Brief description of amendments: The amendments revised the
Technical Specification (TS) requirements related to steam generator
tube integrity. The changes are consistent with Nuclear Regulatory
Commission (NRC)-approved Revision 4 to Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
449, ``Steam Generator Tube Integrity.'' The availability of this TS
improvement was announced in the Federal Register on May 6, 2005 (70 FR
24126) as part of the consolidated line item improvement process.
Date of issuance: August 28, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 144 and 124.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23961). The supplement dated June 5, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the Nuclear
Regulatory Commission (NRC) staff's original proposed no significant
hazards consideration determination as published in the Federal
Register on April 25, 2006 (71 FR 23961).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 28, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 30, 2005 (TS-05-02).
Brief description of amendments: The amendment revises Technical
Specification (TS) Section 5.0 ``Design Features,'' to conform with
NUREG-1431, Revision 3, ``Standard Technical Specifications for
Westinghouse Plants.'' The changes include elimination of the exclusion
area, low population zone, and effluent subsections and associated
figures referred to in Section 5.1 ``Site;'' elimination of Section 5.2
``Containment;'' elimination of Section 5.4 ``Reactor Coolant System,''
as well as Section 5.5 ``Meteorological Tower Location,'' and its
figure. Lastly, a change has been made to TS Section 6.0,
Administrative Control,'' to acquire the component cyclic or transient
limits currently located in the ``Design Features'' section.
Date of issuance: August 2, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 309 and 298.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67752).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 2, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: December 14, 2005 (TS-05-07), as
supplemented by letter dated March 31, 2006.
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 5.7.2.19, ``Containment Leakage Rate Testing
Program,'' to allow a one time, 5-year extension to the current 10 year
test interval for the performance-based leakage rate test program for
10 CFR 50, Appendix J, Type A tests.
Date of issuance: August 22, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 63.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10078). The supplemental letter provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a
[[Page 53726]]
Safety Evaluation dated August 22, 2006.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: April 14, 2005, as supplemented
by letter dated December 21, 2005.
Brief description of amendment: The amendment added a new Technical
Specification (TS) 3.1.9, ``RCS [Reactor Coolant System] Boron
Limitations < 500 [deg]F,'' and revised TS 3.3.1, ``Reactor Trip System
(RTS) Instrumentation,'' for the power range neutron flux--low reactor
trip function.
Date of issuance: August 21, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 174.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 23, 2006 (71 FR
29682). The supplemental letter dated December 21, 2005, provided
clarifying information that 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 published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 21, 2006.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 7, 2006, as supplemented by
letter dated July 25, 2006.
Brief description of amendment: The amendment revised Technical
Specification Table 3.3.1-1, ``Reactor Trip System Instrumentation,''
by adding the existing Surveillance Requirement 3.3.1.16 to Function
3.a of the table.
Date of issuance: August 29, 2006.
Effective date: As of its date of issuance and shall be implemented
within 90 days of the date of issuance.
Amendment No.: 165.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10080). The supplemental letter dated July 25, 2006, provided
additional clarifying information, 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
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 29, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 1st day of September 2006.
For the Nuclear Regulatory Commission.
Timothy McGinty,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E6-14938 Filed 9-11-06; 8:45 am]
BILLING CODE 7590-01-P