[Federal Register Volume 70, Number 108 (Tuesday, June 7, 2005)]
[Notices]
[Pages 33210-33225]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E5-2848]
-----------------------------------------------------------------------
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 May 13, 2005 to May 25, 2005. The last
biweekly notice was published on May 24, 2005 (70 FR 29785).
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
[[Page 33211]]
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-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://
[[Page 33212]]
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].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: May 18, 2005.
Description of amendment request: The proposed amendment would
revise Fermi 2 Technical Specifications (TSs) to add Actions to
Limiting Condition for Operation (LCO) 3.8.1, ``AC Sources--
Operating,'' for one offsite circuit inoperable, for two offsite
circuits inoperable, and for one offsite circuit and one or both
emergency diesel generators (EDGs) in one Division inoperable, in
accordance with Regulatory Guide 1.93, ``Availability of Electric Power
Sources.'' The current Fermi 2 TSs contain only a single Action for one
or two offsite circuits inoperable.
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 to replace the existing LCO 3.8.1 Action C
for one or two offsite circuits inoperable with a required
Completion Time of 12 hours to be in MODE 3, and 36 hours to be in
MODE 4, with new Actions C, D, and E to allow a single offsite
circuit to be inoperable for up to 72 hours, two offsite circuits to
be inoperable for up to 24 hours, and one offsite circuit and one or
both EDGs in one Division to be inoperable for up to 12 hours,
provided other Required Actions are taken is consistent with the
NUREG 1433, ``Standard Technical Specifications General Electric
Plants, BWR/4,'' criteria, and with the guidelines in Regulatory
Guide 1.93. There is no change in plant design, and [Title 10 of the
Code of Federal Regulations (10 CFR)] 10 CFR 50, Appendix A, General
Design Criteria 17, ``Electric Power Systems'' will continue to be
met. Increasing the Completion Times for inoperable offsite circuits
will not significantly increase the potential for a loss of offsite
power. This is due to the redundancy and diversity of the offsite
electrical configuration at Fermi 2. Inoperability of an offsite
circuit does slightly increase the potential for a loss of
divisional power. The probability of losing the opposite division of
offsite power in this condition is extremely small due to the
physical separation of the offsite power sources that feed Fermi 2.
Furthermore, the 10 CFR 50.65(a)(4) program monitors the condition
of the offsite electrical system and switchyard configuration for
each entry into the extended completion time to ensure that there is
no significant increase in the probability or consequences of an
accident.
The proposed change does not alter the operation of any plant
equipment assumed to function in response to an analyzed event or
otherwise increase its failure probability. Therefore, this change
does not involve a significant increase in the probability or the
consequences of any accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not alter the design, configuration, or
method of operation of the plant. It simply provides longer
Completion Times for inoperable offsite circuits. No physical or
operational changes to the components of the A. C. power systems are
being made by this change; therefore, no new system interactions are
being created. The proposed change does not produce any parameters
or conditions that could contribute to the initiation of accidents
different from those already evaluated. Therefore, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed change will replace the existing LCO 3.8.1 Action C
for one or two offsite circuits inoperable with a required
Completion Time of 12 hours to be in MODE 3, and 36 hours to be in
MODE 4, with new Actions C, D, and E to allow a single offsite
circuit to be inoperable for up to 72 hours, two offsite circuits to
be inoperable for up to 24 hours, and one offsite circuit and one or
both EDGs in one Division to be inoperable for up to 12 hours,
provided other Required Actions are taken. This change is consistent
with NUREG 1433, ``Standard Technical Specifications General
Electric Plants, BWR/4,'' and with the guidelines in Regulatory
Guide 1.93. The proposed change does not affect any analysis that is
used to establish safety margins, nor does it alter the design,
configuration, or method of operation of the plant. 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 Section Chief: L. Raghavan.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: April 19, 2005.
Description of amendment request: The proposed amendment would
revise technical specifications (TS) testing frequency for the
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.''
Specifically, the proposed change would revise the frequency for SR
3.1.4.2, Control Rod Scram Time Testing, from ``120 days cumulative
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in licensing amendment applications in the Federal Register on August
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the
model NSHC determination in its application dated April 19, 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The frequency
of surveillance testing is not an initiator of any accident
previously evaluated. The frequency of surveillance testing does not
affect the ability to mitigate any accident previously evaluated, as
the tested component is still required to be operable. 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 extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change does not result in any new or different modes of plant
operation. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change extends the frequency for testing control
rod scram time
[[Page 33213]]
testing from every 120 days of cumulative Mode 1 operation to 200
days of cumulative Mode 1 operation. The proposed change continues
to test the control rod scram time to ensure the assumptions in the
safety analysis are protected. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
Based on the above, the proposed change presents no significant
hazards consideration under the standards set forth in 10 CFR 50.92(c),
and accordingly, a finding of ``no significant hazards consideration''
is justified.
Attorney for licensee: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point
Nuclear Generating Unit Nos. 2 and 3 (IP2 and 3), Westchester County,
New York
Date of amendment request: April 22, 2005.
Description of amendment request: The amendments would revise the
surveillance requirements (SRs) for Technical Specification (TS) 3.3.5,
``Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.''
Specifically, a note would be added to IP2 TS SR 3.3.5.2 to indicate
that the verification of the setpoint is not required for the 480 volt
(V) bus degraded voltage function when performing the trip actuating
device operational test (TADOT). A similar note would be added to IP3
TS SR 3.3.5.1 for the 480V degraded voltage and undervoltage 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 change adds a note to indicate that the IP2 and IP3
degraded voltage relays and the IP3 undervoltage relays do not
require setpoint verification when the TADOT required by TS
surveillances is performed on a monthly basis. Setpoint verification
of these relays occurs as part of the channel calibration that is
performed at either an 18 month or a 24 month frequency. These
relays are used to sense either degraded voltage or undervoltage on
the 480 volt safety related buses and to initiate the start of the
EDG [emergency diesel generator] for all events where the loss of
offsite power is postulated. This function has no effect on the
probability of an accident previously evaluated since it is not
associated with the initiation of any accident. The relay setpoint
verification frequency of 18 or 24 months has no significant effect
on the consequences of an accident because the relays are intended
to be calibrated on this frequency. This frequency of calibration is
based on operating experience, and is consistent with industry
practice. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change adds a note to indicate that the IP2 and IP3
degraded voltage relays and the IP3 undervoltage relays do not
require setpoint verification when the TADOT required by TS
surveillances is performed on a monthly basis. This effectively
changes the frequency required by the surveillance requirement from
31 days to either 18 months or 24 months. The change does not affect
the function of the relays or otherwise affect the design and
operation of plant systems and components and therefore no new
accident scenarios would be created. The change does not affect the
manner is which equipment is operated but does affect the manner in
which it is maintained by extending the frequency for setpoint
verification. The frequency change continues to provide adequate
verification of the operability of equipment and limits the time
which the relay function is inoperable or degraded while performing
verification. Therefore, no new failure modes are being introduced
that could lead to different accidents.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change adds a note to indicate that the IP2 and IP3
degraded voltage relays and the IP3 undervoltage relays do not
require setpoint verification when the TADOT required by TS
surveillances is performed on a monthly basis. Setpoint verification
of these relays occurs as part of the channel calibration that is
performed at either an 18 month or a 24 month frequency. The margin
associated with these relays is the assurance that these relays will
properly sense either degraded voltage or undervoltage on the 480
volt safety related buses and to initiate the start of the EDG for
all events where the loss of offsite power is postulated. The
proposed frequency of calibration is based on operating experience,
and is consistent with industry practice. These indicate that
setpoint verification at 18 month or 24 month [frequency] is
adequate to assure performance of the function. Verification of
setpoints on a monthly basis either degrades the reliability of the
function or makes it inoperable. 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 Section Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 13, 2005.
Description of amendment request: The proposed amendments would
extend the completion time (CT) for required Action A.1, ``Restore
Residual Heat Removal Service Water (RHRSW) subsystem to OPERABLE
status,'' associated with Technical Specification (TS) Section 3.7.1
from 7 days to 10 days. This proposed change would only be used during
the upcoming Unit 1 2006 refueling outage. The establishment of 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 (DGCW) subsystem(s) are
inoperable. This proposed change will only be used during each of the
upcoming Unit 1 2006, and Unit 2 2007, refueling outages, and during
the subsequent Unit 1 2008, refueling outage. An extension of the CT
for required Action C.4, ``Restore required Diesel Generator (DG) to
OPERABLE status,'' associated with TS Section 3.8.1 from 72 hours to 6
days. This proposed change will only be used during the upcoming Unit 2
2007 refueling outage, and during subsequent Unit 1, 2008, refueling
outage. An extension of the CT for required Action F.1, ``Restore one
required Diesel Generator (DG) to OPERABLE status,'' associated with TS
Section 3.8.1 from 2 hours to 6 days. This proposed change will only be
used during the upcoming Unit 2, 2007, refueling outage, and during
subsequent Unit 1, 2008, refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 33214]]
The proposed changes have been evaluated using the risk-informed
processes described in RG [Regulatory Guide] 1.174, ``An Approach
for Using Probabilistic Risk Assessment in Risk-Informed Decisions
on Plant-Specific Changes to the Licensing Basis,'' dated July 1998,
and RG 1.177, ``An Approach for Plant-Specific, Risk-Informed
Decision Making: Technical Specifications,'' dated August 1998. The
risk associated with the proposed change was found to be acceptable.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed change
does not have a detrimental impact on the integrity of any plant
structure, system, or component that initiates an analyzed event. No
active or passive failure mechanisms that could lead to an accident
are affected. Non-code line stops required to isolate the Unit 1
portion of the common discharge header from the Unit 2 portion of
the header during the specified CSCS maintenance will maintain the
availability of the online unit's Division 2 CSCS system. The non-
code line stops being used to isolate the system during the
specified refueling outages are being designed to the same pressure
rating and seismic requirements as the CSCS piping.
Redundancy is provided by designing the CSCS system as multiple
independent subsystems. Separation between subsystems assures that
no single failure can affect more than one subsystem. Therefore,
assuming a single failure in any subsystem including the subsystem
shared between units, two subsystems in each unit will remain
unaffected. These two subsystems can supply the minimum required
cooling water for safe shutdown of a unit or mitigate the
consequences of an accident.
The proposed limited use of increased CT's of the operating
unit's CSCS system maintains the design basis assumptions;
therefore, the proposed change does not involve a significant
increase in the consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change involves the temporary installation of new
equipment (mechanical line stops) that will be designed and
installed to the same pressure rating and seismic design as the CSCS
piping. The currently installed equipment will not be operated in a
new or different manner. No new or different system interactions are
created and no new processes are introduced. The proposed changes
will not introduce any new failure mechanisms, malfunctions, or
accident initiators not already considered in the design and
licensing bases. 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. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed change does not alter any existing setpoints at
which protective actions are initiated and no new setpoints or
protective actions are introduced. The design and operation of the
CSCS system remains unchanged. The risk assessment with the proposed
increase in the CTs for TS 3.7.1, TS 3.7.2, and TS 3.8.1 were
evaluated using the risk-informed processes described in RG 1.174,
``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing
Basis,'' dated July 1998, and RG 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decision Making: Technical Specifications,''
dated August 1998. The risk was shown to be acceptable. 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. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief : Gene Y. Suh.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit No. 2 (BVPS-2), Beaver County,
Pennsylvania
Date of amendment request: April 11, 2005.
Description of amendment request: The proposed amendment would
revise the BVPS-2 Technical Specification (TS) 3.4.5 to change the
scope of the steam generator (SG) tubesheet examinations required in
the SG tubesheet region by using the F* inspection methodology.
Specifically, the proposed amendment would alter the tube inspection to
exclude the portion of the SG tube within the tubesheet below the F*
distance and to exclude the tube-to-tubesheet weld, by crediting the
methodology described in Westinghouse Topical Report, WCAP-16385,
Revision 1. The F* distance is the distance from the top of the
tubesheet to the bottom of the F* length (the maximum length of tubing
below the bottom of the roll transition (BRT) which must be
demonstrated to be non-degraded and which is defined as 1.97 inches on
the hot leg side) plus the distance to the BRT and non-destructive
examination uncertainties. The licensee's proposed amendment also would
revise the TS requirements to require tubes with service-induced
degradation identified in the F* distance or less than or equal to 3.0
inches below the top of the tubesheet, whichever is greater, to be
repaired or removed from service upon detection. The TS Index, affected
TS pages and Bases would also be revised and repaginated as necessary
to reflect the proposed TS change.
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 modifies the BVPS Unit 2 TSs to
incorporate steam generator tube inspection scope based on WCAP-
16385, Revision 1. Of the various accidents previously evaluated in
the BVPS Unit 2 Updated Final Safety Analysis Report (UFSAR), the
proposed changes only affect the steam generator tube rupture (SGTR)
event evaluation and the postulated steam line break (SLB) accident
evaluation. Loss-of-coolant accident (LOCA) conditions cause a
compressive axial load to act on the tube. Therefore, since the LOCA
tends to force the tube into the tubesheet rather than pull it out,
it is not a factor in this amendment request. Another faulted load
consideration is a safe shutdown earthquake (SSE); however, the
seismic analysis of Model 51M SGs has shown that axial loading of
the tubes is negligible during an SSE.
For the SGTR event, the required structural margins of the steam
generator tubes will be maintained by the presence of the tubesheet.
Tube rupture is precluded for cracks in the tube expansion region
due to the constraint provided by the tubesheet. Therefore,
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR
[pressurized-water reactor] Steam Generator Tubes,'' margins against
burst are maintained for both normal and postulated accident
conditions.
The F* length supplies the necessary resistive force to preclude
pullout loads under both normal operating and accident conditions.
The contact pressure results from the tube expansion process used
during manufacturing and from the differential pressure between the
primary and secondary side. The proposed changes do not affect other
systems, structures, components or operational features. Therefore,
the proposed change results in no significant increase in the
probability of the occurrence of an SGTR or SLB accident.
The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary
leakage flow through a postulated broken tube is not affected by the
proposed change since the tubesheet enhances the tube integrity in
the region of the expansion by precluding tube deformation beyond
its initial expanded outside diameter. The resistance to both tube
rupture and collapse is strengthened by the tubesheet in that
region. At normal operating pressures, leakage from primary water
stress corrosion cracking (PWSCC) below the F* length is limited by
both the tube-to-tubesheet crevice and the limited crack opening
permitted by
[[Page 33215]]
the tubesheet constraint. Consequently, negligible normal operating
leakage is expected from cracks within the tubesheet region.
SLB leakage is limited by leakage flow restrictions resulting
from the crack and tube-to-tubesheet contact pressures that provide
a restricted leakage path above the indications and also limit the
degree of crack face opening compared to free span indications. The
total leakage (i.e., the combined leakage for all such tubes) meets
the industry performance criterion, plus the combined leakage
developed by any other alternate repair criteria, and will be
maintained below the maximum allowable SLB leak rate limit, such
that off-site doses are maintained less than 10 CFR [Part] 100
guideline values and the limits evaluated in the BVPS Unit 2 UFSAR.
Therefore, based on 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?
No. The proposed changes do not introduce any changes or
mechanisms that create the possibility of a new or different kind of
accident. Tube bundle integrity will continue to be maintained for
all plant conditions upon implementation of the F* methodology.
The proposed changes do not introduce any new equipment or any
change to existing equipment. No new effects on existing equipment
are created nor are any new malfunctions introduced.
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?
No. The proposed changes maintain the required structural
margins of the steam generator tubes for both normal and accident
conditions, including the planned uprated power level of 2910 Mwt.
NRC [Nuclear Regulatory Commission] Regulatory Guide (RG) 1.121 is
used as the basis in the development of the F* methodology for
determining that steam generator tube integrity considerations are
maintained within acceptable limits. RG 1.121 describes a method
acceptable to the NRC staff for meeting General Design Criteria 14,
15, 31, and 32 by reducing the probability and consequences of an
SGTR. RG 1.121 concludes that by determining the limiting safe
conditions of tube wall degradation beyond which tubes with
unacceptable cracking, as established by inservice inspection,
should be removed from service or repaired, the probability and
consequences of an SGTR are reduced. This RG uses safety factors on
loads for tube burst that are consistent with the requirements of
Section III of the American Society of Mechanical Engineers (ASME)
Code.
For primarily axially oriented cracking located within the
tubesheet, tube burst is precluded due to the presence of the
tubesheet. WCAP-16385, Revision 1, defines a length, F*, of
degradation-free expanded tubing that provides the necessary
resistance to tube pullout due to the pressure-induced forces (with
applicable safety factors applied). Application of the F* criteria
will preclude unacceptable primary-to-secondary leakage during all
plant conditions. The methodology for determining leakage provides
for large margins between calculated and actual leakage values in
the F* criteria.
Plugging of the steam generator tubes reduces the reactor
coolant flow margin for core cooling. Implementation of F*
methodology at Beaver Valley Unit 2 will result in maintaining the
margin of flow that may have otherwise been reduced by tube
plugging.
Based on the above, it is concluded that the proposed changes do
not result in a significant reduction of margin with respect to
plant safety as defined in the Final Safety Analysis Report Update
or bases of the plant Technical Specifications.
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: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant,
Unit No. 2 (SL2), St. Lucie County, Florida
Date of amendment request: March 31, 2005.
Description of amendment request: The proposed amendment would
revise Administrative Technical Specification Section 6.8.4.h,
``Containment Leakage Rate Testing Program,'' to allow a one-time
extension of the currently approved 15-year test interval to
approximately 15.5 years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment of the Technical Specifications adds a one-
time extension to the current surveillance interval for Type A
testing (ILRT [integrated leak rate testing]). The current test
interval of 15 years from the last Type A test would be extended to
end prior to startup from the SL2-17 refueling. This is anticipated
to be an approximately six-month addition to the 15 year interval.
The proposed extension to the Type A testing interval does not
significantly increase the probability of an accident previously
evaluated since the containment Type A test is not a modification,
nor a change in the way that plant systems, structures or components
(SSC) are operated, and is not an activity that could lead to
equipment failure or accident initiation. The proposed extension of
the test interval does not involve a significant increase in the
consequences of an accident since research documented in NUREG-1493
has found that generically, very few potential leak paths are not
identified with Type B and C tests (LLRT [local leak-rate test]).
The Type B and C testing are unaffected by this proposed change. The
NUREG concluded that an increase in the Type A test interval to
twenty years resulted in an imperceptible increase in risk. St.
Lucie Unit 2 provides a high degree of assurance through testing and
inspection that the containment will not degrade in a manner only
detectable by Type A testing. Inspections required by the ASME
[American Society of Mechanical Engineers] Code, the containment
leakage rate testing program, the plant protective coatings program,
and Maintenance Rule are performed in order to identify indications
of containment degradation that could affect leak tightness. Type B
and C testing required by 10 CFR 50, Appendix J, are not affected by
this proposed extension to the Type A test interval and will
identify openings in containment penetrations that would otherwise
require a Type A test.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated.
The proposed change does not result in facility operation that
would create the possibility of a new or different kind of accident
from any accident previously evaluated. The proposed extension to
Type A testing does not create a new or different type of accident
for St. Lucie because no physical plant changes are made and no
compensatory measures are being imposed that could potentially lead
to a failure. There are no operational changes that could introduce
a new failure mode or create a new or different kind of accident.
The proposed change only adds an extension to the current interval
for Type A testing and does not change implementation aspects of the
test.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed change would not result in operation of the
facility involving a significant reduction in a margin of safety.
The proposed license amendment adds a one-time extension to the
current interval for Type A testing (ILRT). The current one-time
test interval of 15 years from the last Type A test would be
extended to end prior to startup from the SL2-17 refueling outage.
This is anticipated to be an approximately six month addition to the
15 year interval.
[[Page 33216]]
The NUREG-1493 generic study of the effects of extending the
Type A test interval out to 20 years concluded that there is an
imperceptible increase in plant risk. A plant specific risk
calculation obtained results consistent with the generic conclusions
regarding risk which show a slight but negligible increase in risk.
Inspections required by the ASME code and maintenance rule are
performed to ensure that the containment will not degrade in a
manner that is only detectable by Type A testing (ILRT).
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 Section Chief: Michael L. Marshall, Jr.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: April 13, 2005.
Description of amendment request: The proposed amendment would
incorporate several Technical Specification Task Force (TSTF) changes
to the licensee's Technical Specifications (TSs). The specific TSTF
changes that would be incorporated are:
1. TSTF-222-A, Revision 1, ``Control Rod Scram Time Testing''--This
change modifies TS Section 3.1.4, ``Control Rod Scram Times,'' to
clarify that control rod scram time testing is required only for core
cells in which work on the control rod or drive has been performed or
fuel has been moved or replaced.
2. TSTF-275-A, Revision 0, ``Clarify Requirement for EDG [emergency
diesel generator] start signal on RPV [reactor pressure vessel] Level--
Low, Low, Low during RPV cavity flood-up''--This change modifies the TS
Section 3.3.5.1, ``ECCS [emergency core cooling system]
Instrumentation,'' to clarify that the ECCS initiation instrumentation,
identified as being required in modes 4 and 5, is required to be
operable only when the associated ECCS subsystems are required to be
operable as defined in limiting condition of operation (LCO) 3.5.2,
``ECCS--Shutdown.''
3. TSTF-300-A, Revision 0, ``Eliminate DG [diesel generator] LOCA
[loss-of-coolant accident]--Start SRs [surveillance requirements] while
in S/D [shutdown] when no ECCS is Required''--This change modifies the
TS Section 3.8.2, ``AC [alternating current] Sources--Shutdown,'' to
add an additional note to the surveillance that verifies automatic
start of the emergency diesel generators and automatic load shedding
from the emergency buses, is considered to be met without the ECCS
initiation signals operable when ECCS initiation signals are not
required to be operable per Table 3.3.5.1-1, ECCS Instrumentation.
4. TSTF-225, Revision 2, ``Fuel movement with inoperable refueling
equipment interlocks''--This change modifies TS Section 3.9.1,
``Refueling Equipment Interlocks,'' to add required actions to allow
insertion of a control rod withdrawal block and verification that all
control rods are fully inserted as alternate actions to suspending in-
vessel fuel movement in the event that one or more required refueling
equipment interlocks are inoperable.
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.
1. Revision of CNS [Cooper Nuclear Station] TS SR 3.1.4.1 and SR
3.1.4.4. The frequency at which control rod scram time is verified
is not a precursor of an accident. A scram time slower than required
might result in an increase in the consequences of an accident.
However, revising the frequency for verifying the scram time of the
control rods does not impact the scram time. Verifying that the
scram time is acceptable will continue to be required prior to plant
startup following fuel movement or work on the control rods or
control rod drive system. Therefore, revising the frequency for
verifying insertion time to clarify when it is required does not
involve a significant increase in the probability of an accident or
an increase in the consequences of an accident.
2. Revision of TS Table 3.3.5.1-1. Clarifying when certain ECCS
instrumentation must be operable with the plant shut down will not
increase either the probability of an accident or the consequences
of the accident. The ECCS instrumentation is required to be operable
only when the associated ECCS subsystems are required to be
operable. This continues to ensure that the instrumentation will be
operable when it is required.
3. Revision of TS SR 3.8.2.1. The frequency of verifying certain
actions by surveillances is not a precursor to accidents. Clarifying
that the actions required in response to an ECCS initiation signal
are not required when the ECCS initiation signals are not required
to be operable does not result in increased probability of an
accident or increased consequences of an accident. Not requiring
that a DG automatically start in response to the ECCS initiation
signal when the ECCS subsystems that are supported by the DG are not
required to be operable does not reduce the required ECCS
protection.
4. Revision of TS 3.9.1., Condition A Required Action. The
actions taken when a refueling equipment interlock is inoperable are
not initiators of any accident previously evaluated. The level of
protection against withdrawing a control rod during the insertion of
a fuel assembly or loading a fuel assembly into the vessel with a
control rod withdrawn, provided by the proposed alternate Required
Actions, is equivalent to that provided by the current Required
Action. The radiological consequences of an accident described in
the Updated Safety Analysis Report (USAR) while taking the proposed
alternate Required Actions are not different from the consequences
of an accident under the current Required Actions.
Based on the above NPPD [Nebraska Public Power District]
concludes that 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 the CNS operating license involve
revisions to the requirements for when certain surveillances are to
be performed (change no. 1 and no. 3), clarification of when ECCS
instrumentation is required to be operable (change no. 2), and
addition of alternative Required Actions if certain plant components
are inoperable (change no. 4). These changes will not result in
revision of plant design, physical alteration of a plant structure,
system, or component (SSC), or installation of a new or different
type of equipment. The changes do not involve any revision of how
the plant, an SSC, or a refueling equipment interlock, are operated.
Based on this, the proposed changes do not create the possibility of
a new or different kind of accident.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
1. Revision of CNS TS SR 3.1.4.1 and SR 3.1.4.4. Sufficiently
rapid insertion of control rods following certain accidents (scram
time) will prevent fuel damage, and thereby maintain a margin of
safety to fuel damage. No change is being made to the required
insertion rate specified in plant technical specifications.
Clarifying when control rod insertion times must be verified
following movement of fuel assemblies, without actually changing the
requirement (verification of insertion times will continue to be
required whenever work that might impact the rod insertion time is
done), does not reduce the margin of safety related to fuel damage.
2. Revision of TS Table 3.3.5.1-1. Clarifying when certain ECCS
instrumentation is required to be operable when CNS is in a shutdown
mode does not change the requirement. Not requiring ECCS signals
that initiate a DG to be operable when the ECCS subsystems that are
supported by
[[Page 33217]]
the DG are not required to be operable does not result in a
reduction of a margin of safety for the safety related equipment
that is required to be operable.
3. Revision of TS SR 3.8.2.1. Clarifying that automatic start of
the DGs in response to the ECCS initiation signal is not required
when the ECCS subsystems that are supported by the DG are not
required to be operable does not result in a reduction in a margin
of safety.
4. Revision of TS 3.9.1, Condition A Required Action. The
proposed alternate Required Actions to be taken when a refueling
interlock is inoperable provide a level of protection against
inadvertent criticality while inserting or moving fuel in the
reactor vessel that is equivalent to the level provided by the
current Required Action. As a result, the proposed alternate
Required Actions do not result in a significant reduction in a
margin of safety related to protection against inadvertent
criticality when inserting or moving fuel assemblies.
Based on the above NPPD concludes that 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: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: David Terao.
PSEG Nuclear, LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: February 25, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.1.3.1, ``Control Rod
Operability,'' such that scram discharge volume (SDV) vent or drain
lines with inoperable valves would be isolated instead of requiring
that the valve be restored to Operable status or the unit be placed in
Hot Shutdown within 12 hours.
The NRC staff issued a Notice of Opportunity for Comment in the
Federal Register on February 24, 2003 (68 FR 8637), on possible
amendments to revise the action for one or more SDV vent or drain lines
with an inoperable valve, 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 license
amendment applications in the Federal Register on April 15, 2003 (68 FR
18294). The licensee affirmed the applicability of the model NSHC
determination (modified slightly to address plant-specific TS format)
in its application dated February 25, 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.
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with inoperable valves instead or requiring the valves to be
restored to operable status or the unit be in hot shutdown within 12
hours. With SDV vent or drain valves inoperable in one or more
lines, the isolation function would be maintained since the
redundant valve in the affected line would perform its safety
function of isolating the SDV. Following the completion of the
required action, the isolation function is fulfilled since the
associated line is isolated. The ability to vent and drain the SDV
is maintained and controlled through administrative controls. This
requirement assures the reactor protection system is not adversely
affected by the inoperable valves. With the safety functions of the
valves being maintained, the probability or consequences of an
accident previously evaluated are not significantly increased.
Criterion 2--The proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in [a] margin of safety.
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDV is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of the SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin of safety.
Based on the reasoning presented above, 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 Section Chief: Darrell J. Roberts.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: March 10, 2005.
Description of amendment request: The amendment would revise
Technical Specification Section 5.5.15, ``Containment Leakage Rate
Testing Program,'' to allow a one-time extension of the interval
between the Type A, integrated leakage rate tests (ILRTs), from 10
years to no more than 15 years.
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 to Technical Specification 5.5.15,
Containment Leakage Rate Testing Program, involves a one-time
extension to the current interval for Type A containment testing.
The current test interval of ten (10) years would be extended on a
one-time basis to no longer than fifteen (15) years from the last
Type A test.
The proposed Technical Specification change does not involve a
physical change to the plant or a change in the manner which the
plant is operated or controlled. The reactor containment is designed
to provide an essentially leak tight barrier against the
uncontrolled release of radioactivity to the environment for
postulated accidents. As such the reactor containment itself and the
testing requirements invoked to periodically demonstrate the
integrity of the reactor containment exist to ensure the plant's
ability to mitigate the consequences of an accident, and do not
involve the prevention or identification of any precursors of an
accident.
The proposed change involves only the extension of the interval
between Type A containment leakage tests. Type B and C containment
leakage tests will continue to be performed at the frequency
currently required by plant Technical Specifications. Industry
experience has shown, as documented in NUREG-1493, that Type B and C
containment leakage tests have identified a very large percentage of
containment leakage paths and that the percentage of containment
leakage paths that are detected only by Type A testing is very
small. The Ginna ILRT test history supports this conclusion. In
NUREG-1493 Section 10, Summary of Technical Findings, it is
concluded, in part, that reducing the frequency of Type A
containment leak tests to once per twenty (20) years leads to an
imperceptible increase in risk.
[[Page 33218]]
The proposed change does not result in an increase in core
damage frequency since the containment system is used for mitigation
purposes only. Containment Leakage Rate Testing Program local leak
rate test requirements and administrative controls such as design
change control, ASME [American Society of Mechanical Engineers]
Section XI Inservice Inspection (ISI) Program Containment Repair and
Replacement Program and procedural requirements for system
restoration ensure that containment integrity is not degraded by
plant modifications or maintenance activities. The design and
construction requirements of the reactor containment itself combined
with the containment inspections performed in accordance with the
ASME Section XI Inservice Inspection (ISI) Program Containment
Program, Boric Acid Corrosion Program, inspections in accordance
with Regulatory Guide 1.163 position C.3 and the Maintenance Rule
serve to provide a high degree of assurance that the containment
will not degrade in a manner that is detectable only by Type A
testing.
Therefore, the proposed Technical Specification change does not
involve a significant increase in the consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change to Technical Specification 5.5.15 involves a
one-time extension to the current interval for Type A containment
testing. The reactor containment and the testing requirements
invoked to periodically demonstrate the integrity of the reactor
containment exist to ensure the plant's ability to mitigate the
consequences of an accident and do not involve the prevention or
identification of any precursors of an accident. The proposed
Technical Specification change does not involve a physical change to
the plant (i.e., no new or different type of equipment will be
installed) or changes in the methods in which the plant is operated
or controlled.
Therefore, the proposed Technical Specification change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change to Technical Specifications involves a one-
time extension to the current interval for Type A containment
testing. The proposed Technical Specification change does not alter
the manner in which safety limits, limiting safety system set
points, or limiting conditions for operation are determined. The
specific requirements and conditions of the Primary Containment
Leakage Rate Testing Program, as defined in Technical
Specifications, exist to ensure that the degree of reactor
containment structural integrity and leak-tightness that is
considered in the plant safety analysis is maintained. The overall
containment leakage rate limit specified by Technical Specifications
is maintained. The proposed change involves only the extension of
the interval between Type A containment leakage tests. Type B and C
containment leakage tests will continue to be performed at the
frequency currently required by plant Technical Specifications.
Ginna and industry experience strongly supports the conclusion
that Type B and C testing detects a large percentage of containment
leakage paths and that the percentage of containment leakage paths
that are detected only by Type A testing is small. The containment
inspections performed in accordance with the ASME Section XI
Inservice Inspection (ISI) Program Containment Program, Boric Acid
Corrosion Program, inspections in accordance with Regulatory Guide
1.163 position C.3 and the Maintenance Rule serve to provide a high
degree of assurance that the containment will not degrade in a
manner that is detectable only by Type A testing. The combination of
these factors ensures that the margin of safety that is inherent in
plant safety analysis is maintained.
Therefore, the proposed Technical Specification change does not
involve a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: April 29, 2005.
Description of amendment request: The amendment would revise
Technical Specification Section 3.7.3, ``Main Feedwater Regulating
Valves (MFRVs), Associated Bypass Valves, and Main Feedwater Pump
Discharge Valves (MFPDVs),'' to allow the use of the main feedwater
isolation valves in lieu of the main feedwater pump discharge valves to
provide isolation capability to the steam generators in the event of a
steam line break.
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:
1. Does the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes involve a modification to the plant
configuration to ensure the acceptability of containment response
for Steam Line Breaks (SLB) inside containment.
The changes have also been evaluated to ensure the core response
for steam system piping breaks remains acceptable. The changes to
the Technical Specifications (TS) are necessary to properly
accommodate the changes in plant configuration and ensure proper
testing of the modified components.
The proposed changes do not adversely affect accident initiators
or precursors nor significantly alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes do not
adversely 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 and amounts of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposures. The proposed changes cannot
affect the probability of an accident occurring since they reflect a
change in plant design consistent with current design which is not
an accident initiator. The proposed changes cannot increase the
consequences of postulated accidents since they reflect a change in
plant design that will continue to mitigate the effects of feedwater
addition to a faulted steam generator for a main steam line break
inside containment.
Therefore, the 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.
The proposed changes involve a modification to the plant
configuration to ensure the acceptability of containment response
for Steam Line Breaks (SLB) inside containment. The changes have
also been evaluated to ensure the core response for steam system
piping breaks remains acceptable. The changes to the Technical
Specifications (TS) are necessary to properly accommodate the
changes in plant configuration and ensure proper testing of the
modified components.
The change in plant configuration significantly reduces the
available water volume and therefore the mass and energy released to
the containment in the event of an SLB with failure of a feedwater
regulating valve. Existing feedwater flow paths or piping are not
significantly altered. An existing manual valve in the flow path to
each steam generator is utilized as the main feedwater isolation
valve by the addition of an air actuator to provide automatic
isolation capability. The changes do not involve a significant
change in the methods governing normal plant operation. The TS
changes modify the limiting condition for operation, required action
statements, associated completion times and surveillance
requirements to those that are consistent with those previously
approved for Westinghouse
[[Page 33219]]
plants in the Standard Technical Specifications found in NUREG-1431.
The proposed TS changes do not create the possibility of a new or
different [kind] of accident from those previously evaluated since
they reflect a design change that will accomplish the same feedwater
isolation function as previously performed by the main feedwater
pump discharge isolation valves with no significant change to the
manner in which the feedwater system operates.
Therefore, the changes do 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 changes involve a modification to the plant
configuration to ensure the acceptability of containment response
for Steam Line Breaks (SLB) inside containment. The changes have
also been evaluated to ensure the core response for steam system
piping breaks remains acceptable. The changes to the Technical
Specifications (TS) are necessary to properly accommodate the
changes in plant configuration and ensure proper testing of the
modified components.
The level of safety of facility operation is unaffected by the
proposed changes since there is no change in the intent of the TS
requirements of assuring proper main feedwater isolation in the
event of a steam line break inside containment. The response of the
plant systems to accidents and transients reported in the Updated
Final Safety Analysis Report (UFSAR) is not adversely affected by
this change. Therefore, the capability to satisfy accident analysis
acceptance criteria is not adversely affected. The TS changes modify
the limiting condition for operation, required action statements,
associated completion times and surveillance requirements to those
that are consistent with those previously approved for Westinghouse
plants in the Standard Technical Specifications found in NUREG-1431.
The proposed TS changes do not involve a significant reduction in
[a] margin of safety since they are based upon a modification that
will maintain [a] margin of safety with respect to feedwater
addition for a main steam line break inside containment to the
previously analyzed condition. Therefore, the 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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: April 29, 2005.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.5.1, ``Accumulators,'' and TS 3.5.4,
``Refueling Water Storage Tank (RWST),'' to reflect the results of
revised analyses performed to accommodate a planned power uprate for
the facility and revise TS 5.6.5, ``Core Operating Limits Report
(COLR),'' to permit the use of NRC-approved methodology for large-break
and small-break loss-of-coolant accidents (LOCAs).
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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes include revising accumulator volume and
boron concentration requirements and Refueling Water Storage Tank
(RWST) boron concentration requirements that are necessary to
accommodate expected changes in the nuclear fuel (e.g., higher
enrichment) that are associated with the planned power uprate.
Additionally, the change would allow Ginna to utilize analysis
methodologies that have been previously approved for use at
Westinghouse nuclear plants. The changes to the TS are necessary to
ensure the acceptability of these systems to perform their intended
function in the event of an accident.
The proposed changes do not adversely affect accident initiators
or precursors nor significantly alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes do not
adversely 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 and amounts of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposures. The proposed changes cannot
affect the probability of an accident occurring since they reflect a
necessary change in plant design consistent with current design
which is not an accident initiator. The proposed changes cannot
increase the consequences of postulated accidents since they reflect
a change in plant design that will continue to mitigate the effects
of potential accidents. Therefore, the 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.
The proposed changes include revising accumulator volume and
boron concentration requirements and RWST boron concentration
requirements that are necessary to accommodate expected changes in
the nuclear fuel (e.g., higher enrichment) that are associated with
the planned power uprate. Additionally, the change would allow Ginna
to utilize analysis methodologies that have been previously approved
for use at Westinghouse nuclear plants. The changes to the TS are
necessary to ensure the acceptability of these systems to perform
their intended function in the event of an accident.
The proposed changes involve changes to accumulator volume and
boron concentration requirements and RWST boron concentration
requirements to ensure the continued acceptability of LOCA and post
LOCA analysis results. The changes to the Technical Specifications
(TS) are necessary to properly accommodate the changes in plant
design. The changes ensure applicable acceptance criteria will
continue to be met. The changes do not involve a significant change
in the methods governing normal plant operation. The proposed TS
changes do not create the possibility of a new or different [kind]
of accident from those previously evaluated since they reflect a
change that will ensure the accumulators and RWST will continue to
perform their intended function in the event of an accident.
Therefore, the changes do 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 changes include revising accumulator volume and
boron concentration requirements and RWST boron concentration
requirements that are necessary to accommodate expected changes in
the nuclear fuel (e.g., higher enrichment) that are associated with
the planned power uprate. Additionally, the change would allow Ginna
to utilize analysis methodologies that have been previously approved
for use at Westinghouse nuclear plants. The changes to the TS are
necessary to ensure the acceptability of these systems to perform
their intended function in the event of an accident.
The level of safety of facility operation is not significantly
affected by the proposed changes since there is no change in the
intent of the TS requirements of assuring proper plant response in
the event of an accident. The response of the plant systems to
accidents and transients reported in the Updated Final Safety
Analysis Report (UFSAR) is not adversely affected by this
[[Page 33220]]
change. Therefore, the capability to satisfy accident analysis
acceptance criteria is not adversely affected. The proposed TS
change cannot involve a significant reduction in [a] margin of
safety since it is based upon changes that will maintain a
substantial margin of safety with respect to accumulators and RWST
functions. Therefore, the 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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: April 29, 2005.
Description of amendment request: The amendment would revise
Technical Specifications (TSs) to allow the use of Relaxed Axial Offset
Control (RAOC) methodology in reducing operator action required to
maintain conformance with power distribution control TS and increasing
the ability to return to power after a plant trip or transient while
still maintaining margin to safety limits under all operating
conditions.
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:
1. Does the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes do not initiate an accident. Evaluations
and analyses of accidents, which are potentially affected by the
parameters and assumptions, associated with the RAOC and
FQ(Z) methodologies have shown that design standards and
applicable safety criteria will continue to be met. The
consideration of these changes does not result in a situation where
the design, material, or construction standards that were applicable
prior to the change are altered. Therefore, the proposed changes
will not result in any additional challenges to plant equipment that
could increase the probability of any previously evaluated accident.
The proposed changes associated with the RAOC and
FQ(Z) methodologies do not affect plant systems such that
their function in the control of radiological consequences is
adversely affected. The actual plant configurations, performance of
systems, or initiating event mechanisms are not being changed as a
result of the proposed changes. The design standards and applicable
safety criteria limits will continue to be met; therefore, fission
barrier integrity is not challenged. The proposed changes associated
with the RAOC and FQ(Z) methodologies have been shown not
to adversely affect the plant response to postulated accident
scenarios. The proposed changes will therefore not affect the
mitigation of the radiological consequences of any accident
described in the Updated Final Safety Analysis Report (UFSAR).
Therefore, the proposed changes do not involve a significant
increase in the 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 will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed changes do not challenge the performance or integrity
of any safety-related system. The possibility for a new or different
type of accident from any accident previously evaluated is not
created since the proposed changes do not result in a change to the
design basis of any plant structure, system or component. Evaluation
of the effects of the proposed changes has shown that design
standards and applicable safety criteria continue to be met.
Equipment important to safety will continue to operate as
designed and component integrity will not be challenged. The
proposed changes do not result in any event previously deemed
incredible being made credible. The proposed changes will not result
in conditions that are more adverse and will not result in any
increase in the challenges to safety systems.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes will not involve a significant reduction in
a margin of safety.
The proposed changes will assure continued compliance within the
acceptance limits previously reviewed and approved by the NRC for
RAOC and FQ(Z) methodologies. The appropriate acceptance
criteria for the various analyses and evaluations will continue to
be met.
The projected impact associated with the implementation of RAOC
on peak cladding temperature (PCT) has been incorporated into the
LOCA [loss-of-coolant accident] analyses for the planned extended
power uprate. It has [been] determined that implementation of RAOC
at the extended power uprate power level does not result in a
significant reduction in a margin of safety. The analysis performed
for EPU [extended power uprate] bounds operation at the current
power level.
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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
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.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: May 5, 2005.
Brief description of amendment request: The proposed amendment
would change the Technical Specifications to modify the auxiliary
feedwater (AFW) pump suction protection requirements and change the
design basis as described in the Updated Safety Analysis Report to
revise the functionality of the discharge pressure switches to provide
pump runout protection, which requires operator actions to restore the
AFW pumps for specific post-accident recovery activities.
[[Page 33221]]
Date of publication of individual notice in Federal Register: May
13, 2005 (70 FR 25619).
Expiration date of individual notice: June 13, 2005.
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 amendment request: April 27, 2005, as supplemented May 4,
2005.
Description of amendment request: The proposed amendment would
revise the SSES 1 and 2, Technical Specification 3.8.4, ``DC Sources-
Operating,'' to address new required actions for the condition in which
a 125 volt direct current (VDC) charger is taken out of service for the
purposes of a special inspection and related activities. The proposed
changes would be in effect until the special inspection and related
activities are completed on each of the 125 VDC Class 1E battery
chargers but no later than 60 days following the issuance of the Unit 1
and 2 amendments. Specifically, required Action A.2.1 would require
that surveillance requirement 3.8.6.1 be performed within 2 hours and
once-per-12 hours thereafter; and, required Action A.2.2 would restrict
the restoration time for the inoperable electrical power subsystem to
36 hours.
Date of publication of individual notice in Federal Register: May
12, 2005 (70 FR 25122).
Expiration date of individual notice: Comments, May 27, 2005;
Hearing, July 11, 2005.
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. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: October 21, 2004, as
supplemented January 4, 2005.
Brief description of amendment: The amendment deleted the Technical
Specification (TS) requirements to submit monthly operating reports and
annual occupational radiation exposure reports. The change is
consistent with Revision 1 of NRC-approved Industry/Technical
Specifications Task Force (TSTF) Standard TS Change Traveler, TSTF-369,
``Removal of Monthly Operating Report and Occupational Radiation
Exposure Report.'' This TS improvement was announced in the Federal
Register (69 FR 35067) on June 23, 2004, as part of the Consolidated
Line Item Improvement Process (CLIIP).
Date of issuance: May 20, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 165.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19114).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 20, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: February 14, 2005.
Brief description of amendments: The amendments revised the
Technical Specification Surveillance Requirement 3.3.7.1 to extend the
frequency of the channel functional test for the Engineered Safeguards
Protective System digital actuation logic channels from once every 31
days to once every 92 days.
Date of Issuance: May 19, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 345, 347 and 346.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12745).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 19, 2005.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: December 20, 2004, as supplemented by
letter dated April 12, 2005.
Brief description of amendment: The amendment deletes TS 6.6.1,
``Occupational Radiation Exposure Report'' and TS 6.6.4, ``Monthly
Operating Reports,'' as described in the Notice of Availability
published in the Federal Register on June 23, 2004 (69 FR 35067).
Date of issuance: May 13, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
[[Page 33222]]
Amendment No.: 259.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2890). The supplement dated April 12, 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 13, 2005.
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: December 22, 2004.
Brief description of amendment: The requested change deletes
Technical Specification (TS) 6.9.1.5, ``Occupational Radiation Exposure
Report,'' and 6.9.1.6, ``Monthly Operating Reports,'' as described in
the Notice of Availability published in the Federal Register on June
23, 2004 (69 FR 35067).
Date of issuance: May 25, 2005.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 202.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12746).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 25, 2005.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana
Date of amendment request: April 27, 2005, as supplemented by
letter dated May 12, 2005.
Brief description of amendment: The amendment removed the license
condition on instrument uncertainty that was imposed on the Waterford 3
license with the issuance of License Amendment 199 for the extended
power uprate.
Date of issuance: May 23, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 201.
Facility Operating License No. NPF-38: The amendment revised the
Operating License.
Date of initial notice in Federal Register: May 5, 2005 (70 FR
23892). The May 12, 2005, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 23, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: September 15, 2004.
Brief description of amendments: The amendments deleted the
Technical Specification (TS) requirements related to hydrogen
recombiners. The TS changes support implementation of the revisions to
Title 10 of the Code of Federal Regulations (10 CFR) section 50.44,
``Standards for Combustible Gas Control System in Light-Water-Cooled
Power Reactors,'' that became effective on October 16, 2003. The
changes are consistent with Revision 1 of the NRC-approved Industry/
Technical Specifications Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-447, ``Elimination of Hydrogen
Recombiners and Change to Hydrogen and Oxygen Monitors.''
Date of issuance: May 19, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 137, 137, 143, 143.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5243).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 19, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3,York
and Lancaster Counties, Pennsylvania
Date of application for amendments: June 15, 2004, as supplemented
January 12, 2005.
Brief description of amendments: These amendments changed
Surveillance Requirement (SR) 3.8.1.3, monthly diesel surveillance
test; SR 3.8.1.10, diesel full load rejection test; SR 3.8.1.14.3.b,
diesel 24-hour run test; and, SR 3.8.1.15, diesel hot restart test, to
permit these tests to be run at a higher load up to 2800 kW.
Date of issuance: May 20, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendments Nos.: 253 and 256.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 2004, (69 FR
43461). The January 12, 2005, supplement 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 on July 20, 2004 (69 FR 43461).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 20, 2005.
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: May 21, 2004, as supplemented by letters
dated September 16, and December 14, 2004.
Brief description of amendment: The amendment revised the Technical
Specification Bases Section to allow the containment spray pumps to be
secured during a loss-of-coolant accident, when certain conditions are
met, to minimize the potential for containment sump clogging.
Date of issuance: May 20, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 120 days of issuance.
Amendment No.: 235.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications Bases.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34703). The September 16, and December 14,
[[Page 33223]]
2004, supplemental letters provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated May 20, 2005.
No significant hazards consideration comments received: No.
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 application for amendment: May 21, 2004.
Brief description of amendment: The amendment revises Technical
Specifications related to the reactor coolant pump flywheel inspection
program by relocating the requirements from the limiting conditions for
operation to the administrative controls section and increasing the
inspection interval to 20 years.
Date of issuance: May 9, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 172.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9995).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 9, 2005.
No significant hazards consideration comments received: No.
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: October 21, 2004, as supplemented
December 13 and 22, 2004, and February 23 and March 1, 2005.
Brief description of amendments: Conforming license amendments to
remove AEP Texas Central Company as an ``Owner'' in the facility
operating licenses.
Date of issuance: May 19, 2005.
Effective date: As of the date of issuance and shall be implemented
within 365 days of issuance.
Amendment Nos.: Unit 1-172; Unit 2-160
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the licenses.
Date of initial notice in Federal Register: December 14, 2004 (69
FR 76019). The supplements dated December 13 and 22, 2004, and February
23 and March 1, 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 May 19, 2005.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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
[[Page 33224]]
System's (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].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, 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 identify 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
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
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.\1\ 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 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.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
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 or 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).
[[Page 33225]]
Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama
Date of amendment request: April 26, 2005, as supplemented on April
29 and on May 3, 2005.
Description of amendment request: Revises the Completion Time for
the Action associated with an inoperable low pressure Emergency Core
Cooling System injection/spray system to 14 days on a one-time basis.
Date of issuance: May 9, 2005.
Effective date: As of date of issuance and shall be implemented
within 7 days.
Amendment No.: 294.
Facility Operating License No. DPR-52: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, and final determination of NSHC determination
are contained in a Safety Evaluation dated May 9, 2005.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Dated in Rockville, Maryland, this 27th day of May 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. E5-2848 Filed 6-6-05; 8:45 am]
BILLING CODE 7590-01-P