[Federal Register Volume 68, Number 246 (Tuesday, December 23, 2003)]
[Notices]
[Pages 74262-74273]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-31314]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision 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 November 27 through December 11, 2003. The
last biweekly notice was published on December 9, 2003 (68 FR 68654).
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.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing 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.
[[Page 74263]]
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 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
By January 22, 2004, 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 2.714, 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 by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of 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. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 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
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, 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 filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
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 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/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
[[Page 74264]]
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: June 11, 2003, as supplemented by letter
dated August 20, 2003, and October 13, 2003.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 5.5.16, ``Containment Leakage Rate
Testing Program'' to allow a one-time extension of the containment Type
A leak rate test interval from once in 10 years to once in 15 years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed change to TS 5.5.16 provides a one-time extension
of the containment Type A test interval to 15 years for HBRSEP (H.
B. Robinson Steam Electric Plant), Unit No. 2. The proposed TS
change does not involve a physical change to the plant or a change
in the manner in which the plant is operated or controlled. The
containment vessel is designed to provide a leak-tight barrier
against the uncontrolled release of radioactivity to the environment
in the unlikely event of postulated accidents. As such, the
containment vessel is not considered as the initiator of an
accident. Therefore, the proposed TS change does not involve a
significant increase in the probability of an accident previously
evaluated.
The proposed change involves only a one-time change to the
interval between containment Type A tests. Types B and C leakage
testing will continue to be performed at the intervals specified in
10 CFR part 50, Appendix J, Option A, as required by the HBRSEP,
Unit No. 2, TS. As documented in NUREG-1493, ``Performance-Based
Containment Leakage-Test Program,'' industry experience has shown
that Types B and C containment leak rate tests have identified a
very large percentage of containment leak paths, and that the
percentage of containment leak paths that are detected only by Type
A testing is very small. In fact, an analysis of 144 integrated leak
rate tests, including 23 failures, found that none of the failures
involved a containment liner breach. NUREG-1493 also concluded, in
part, that reducing the frequency of containment Type A testing to
once per 20 years results in an imperceptible increase in risk. The
HBRSEP, Unit No. 2, test history and risk-based evaluation of the
proposed extension to the Type A test interval supports this
conclusion. The design and construction requirements of the
containment vessel, combined with the containment inspections
performed in accordance with the American Society of Mechanical
Engineers (ASME) Code, Section XI, and the Maintenance Rule (10 CFR
50.65) provide a high degree of assurance that the containment
vessel will not degrade in a manner that is detectable only by Type
A testing. Therefore, the proposed TS change does not involve a
significant increase in the consequences of an accident previously
evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident From Any Previously Evaluated.
The proposed change to TS 5.5.16 provides a one-time extension
of the containment Type A test interval to 15 years for HBRSEP, Unit
No. 2. The proposed change to the Type A test interval does not
result in any physical changes to HBRSEP, Unit No. 2. In addition,
the proposed test interval extension does not change the operation
of HBRSEP, Unit No. 2, such that a failure mode involving the
possibility of a new or different kind of accident from any accident
previously evaluated is created.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety.
The proposed change to TS 5.5.16 provides a one-time extension
of the containment Type A test interval to 15 years for HBRSEP, Unit
No. 2. The NUREG-1493 study of the effects of extending containment
leak rate testing found that a 20 year extension for Type A testing
resulted in an imperceptible increase in risk to the public. NUREG-
1493 found that, generically, the design containment leak rate
contributes a very small amount to the individual risk, and that the
decrease in Type A testing frequency would have a minimal affect on
this risk, since most potential leak paths are detected by Type B
and C testing.
The proposed change only involves a one-time extension of the
interval for containment Type A testing; the overall containment
leak rate specified by the HBRSEP, Unit No. 2, TS is being
maintained. Type B and C testing will continue to be performed at
the frequency required by the HBRSEP, Unit No. 2, TS. The regular
containment inspections being performed in accordance with the ASME
Code, Section XI, and the Maintenance Rule (10 CFR 50.65) provide a
high degree of assurance that the containment will not degrade in a
manner that is only detectable by Type A testing. In addition, a
plant-specific risk evaluation demonstrates that the extension of
the Type A test interval from 10 years to 15 years results in a
``very small'' increase in risk for those accident sequences
influenced by Type A testing and a ``small'' increase in risk when
compared to the test frequency of 3 tests per 10 years.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based on the above discussion, Progress Energy Carolinas, Inc.,
has determined that the requested change does not involve a
significant hazards consideration.
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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen G. Howe.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: March 20, 2003.
Description of amendment request: The amendments would revise the
Technical Specifications to update the heatup, cooldown, criticality,
and inservice test pressure and temperature limits for the reactor
coolant system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
First Standard
Does operation of the facility in accordance with the proposed
amendment involve a significant increase in the probability or
consequences of an accident previously evaluated.
Response: No.
The proposed changes to the reactor coolant system (RCS)
pressure-temperature (P/T) limits are developed utilizing the
methodology of ASME (American Society of Mechanical Engineers) XI,
10 CFR (part) 50 Appendix G, in conjunction with the methodology of
Code Case N-640. Usage of these methodologies provides compliance
with the underlying intent of 10 CFR (part) 50 Appendix G and
provides operational limits that ensure failure of the reactor
vessel will not occur. The proposed changes to allow operation with
two pumps capable of injecting into the RCS and utilization of the
residual heat removal (RHR) suction relief valves has been evaluated
and determined to provide adequate protection of the RCS from the
worst case pressure transient.
The probability of any design basis accident (DBA) is not
affected by these changes, nor are the consequences of any DBA
affected by these changes. The P/T limits, and low temperature
overpressure protection (LTOP) setpoints, and Tenable
value are not considered to be initiators or contributors to any
accident analysis addressed in the Catawba UFSAR (updated final
safety analysis report).
[[Page 74265]]
The proposed changes do not adversely affect the integrity of
the RCS such that its function in the control of radiological
consequences is affected. The changes do not alter any assumption
previously made in the radiological consequence evaluations nor
affect the mitigation of the radiological consequences of an
accident previously evaluated. The proposed changes to the TS are
consistent with the intent of the flexibility currently provided in
NUREG-1431, Standard Technical Specifications for Westinghouse
Plants, Revision 2.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated in the updated final safety analysis report
(UFSAR) because the accident analysis assumptions and initial
conditions will continue to be maintained.
Second Standard
Does operation of the facility in accordance with the proposed
amendment create the possibility of a new or different kind of
accident from any accident previously evaluated.
Response: No.
The proposed change does not involve any physical alteration of
plant systems, structures, or components. The requirements for the
P/T limit curves and LTOP setpoints remain in place. The fundamental
approach follows approved ASME and Westinghouse report methodology.
The proposed curves and change to the enable temperature for LTOP
system reflect changes in material properties acknowledged and
managed by regulation and an upgrade in technology, which has been
approved by ASME.
The proposed changes to allow operation with two pumps capable
of injecting into the RCS and utilization of the RHR suction relief
valves has been evaluated. The evaluation has shown that both the
PORVs (power-operated relief valves) and RHR suction relief valves
provide adequate relief protection of the RCS from the worst case
pressure transient and provide equivalent protection to that already
allowed by the current TS (technical specification).
The proposed changes do not introduce new failure mechanisms for
system structures, or components not already considered in the
UFSAR. Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated is not created
because no new failure mechanisms or initiating events have been
introduced.
Third Standard
Does operation of the facility in accordance with the proposed
amendment involve a significant reduction in the margin of safety.
Response: No.
The proposed changes are developed utilizing the methodology of
ASME XI, 10 CFR (part) 50 Appendix G, in conjunction with Code Case
N-640 and Code Case N-641 methodology. Usage of these methodologies
provides compliance with the underlying intent of 10 CFR (part) 50
Appendix G and provides operational limits that ensure failure of
the reactor vessel will not occur. Although the Code Cases
constitute relaxation from the current requirements of 10 CFR (part)
50 Appendix G, the alternative methodology allowed by the Code is
based on industry experience gained since the inception of the 10
CFR (part) 50 Appendix G requirements for which some of the
requirements have now been determined to be excessively
conservative. The more appropriate assumptions and provisions
allowed by the Code Cases maintain a margin of safety that is
consistent with the intent of 10 CFR (part) 50 Appendix G, i.e.,
with regard to the margin originally contemplated by 10 CFR (part)
50 Appendix G for determination of RCS P/T limits.
The analyses completed for this proposed TS amendment
demonstrate that established acceptance criteria continue to be met.
Specifically, the P/T limit curves, LTOP setpoints, allowances for
operating two pumps, utilization of RHR suction relief valves and
LTOP Tenable values provide acceptable margin to vessel
fracture under both normal operation and LTOPs design basis (mass
addition and heat addition) accident conditions. The proposed
changes to the TS are consistent with the intent of the flexibility
currently provided in NUREG-1431, Standard Technical Specifications
for Westinghouse Plants, Revision 2. Therefore, there will be no
significant reduction in a margin of safety as a result of the
proposed changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendment request: March 28, 2003, as
supplemented by letter dated October 23, 2003.
Description of amendment request: The proposed amendments would
revise the technical specifications to reduce the main steam line low
pressure primary containment isolation allowable value.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Current licensing bases events remain bounding for ATWS,
transient, and accident analyses. For the bounding events, a
reduction in the allowable value for the MSL LPIS produces no
significant change in the limiting results with respect to the
acceptance criteria. The proposed change does not alter the response
of plant equipment to transient conditions, nor does it introduce
any new equipment, modes of system operation or failure mechanisms.
The proposed change does not adversely impact structures, systems,
or components.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
ECCS-LOCA Performance
In the analyses used to evaluate the ECCS-LOCA performance, the
MSIVs are assumed to close at the start of the accident for all
break locations. Therefore, the low pressure isolation trip is not
used in the LOCA analyses and the LOCA analysis results are not
affected by the reduction in the LPIS.
For large breaks in the MSL (both inside and outside
containment), the MSIV closure is initiated by a high steam line
flow signal at the beginning of the event, well before the LPIS is
reached. For these cases, the ECCS performance is not affected by
the reduction in the LPIS.
If the steam line break is too small to result in a high flow
isolation signal, MSIV closure may be initiated by another signal
(e.g., high steam line tunnel temperature or low reactor water
level) or it may occur due to the LPIS trip. In either case, steam
line breaks of any size are not the limiting events with respect to
ECCS performance, and a 40 psi reduction in the LPIS will not affect
compliance with the acceptance criteria of 10 CFR 50.46,
``Acceptance criteria for emergency core cooling systems for light-
water nuclear power reactors.''
Based on the above discussions, the reduction of the MSIV LPIS
has no adverse impact on the plant response to a LOCA or on
compliance with the acceptance criteria of 10 CFR 50.46.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of a previously
evaluated ECCS-LOCA accident.
Containment System Response
In evaluating containment response to pipe breaks inside
containment, the MSIVs are assumed to close at the start of the
accident for all break locations in the containment system response
analyses. Therefore, the low pressure isolation trip is not assumed
and the analysis results are not affected by the reduction in the
LPIS.
In the event that MSIV closure does not occur at the beginning
of the accident, MSL isolation is effectively achieved as the
pressure regulator closes the turbine control and bypass valves in
an attempt to maintain turbine throttle pressure at the regulator
setpoint of approximately 925 psig. Thus, for events other than
breaks in the main steam
[[Page 74266]]
line, isolation occurs before the LPIS is reached.
For large breaks in the MSL (both inside and outside
containment), the MSIV closure is initiated by a high steam line
flow signal at the beginning of the event, well before the LPIS is
reached. For these cases, the containment system response is not
affected by the reduction in the LPIS. For a steam line break too
small to result in a high flow isolation signal, MSIV closure may be
initiated by another signal (e.g., low reactor water level) or it
may occur due to the LPIS trip. Small breaks do not determine the
peak drywell shell temperature and equipment qualification (EQ)
envelope. Large breaks, as characterized in Section 3.3.2 of
Attachment 4, are large enough to depressurize the reactor
irrespective of the MSIV closure. Hence, a 40-psi reduction in the
LPIS will not affect the peak drywell shell temperature or the
drywell temperature EQ envelope.
Based on the above discussions, the reduction of the MSIV LPIS
has no adverse impact on the containment system response.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated for containment system response.
Subcompartment Pressurization
The MSL break mass and energy release used in the evaluation are
based on steady-state reactor operating conditions. Therefore, the
low pressure isolation trip is not used in the subcompartment
pressurization analysis. In addition, the peak annulus
pressurization loads occur at the beginning of the event, well
before MSIV closure can occur.
The subcompartment pressurization results are not affected by
the reduction in the MSL LPIS.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated for subcompartment pressurization.
Appendix R Fire Protection
The reactor system response for the Appendix R fire protection
analysis was performed during the Extended Power Uprate (EPU)
project. The sequence of events for the analysis shows that closure
of the MSIVs is initiated on low-low reactor water level. However,
before the LPIS setpoint is reached, the turbine control valves
closing on low inlet pressure effectively isolate steam flow
following a scram. The revised LPIS has no adverse impact on the
reactor system response to an Appendix R fire protection event.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated for Appendix R fire protection.
Station Blackout
The initiating event for a station blackout, a loss of off-site
power, results in MSIV closure at the beginning of the event. The
reduction of the MSL LPIS has no adverse impact on the reactor
system response during a station blackout.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of a previously
evaluated station blackout event.
High Energy Line Break
The steam line break analysis assumes closure of the MSIVs due
to high steam line flow at the beginning of the event. Thus, the low
pressure isolation trip is not used in the analyses and the results
are not adversely affected by the reduced LPIS.
The steam line break case determines the short-term peak steam
tunnel temperature. However, the range of break sizes for which the
low pressure isolation trip initiates MSIV closure is limited. Such
a break must be large enough to depressurize the vessel below the
pressure regulator setpoint, approximately 925 psig, but small
enough such that high steam line flow trip does not result. Although
such cases could result in an increase in the mass and energy
released, similar to a larger line break, isolation will still occur
before the LPIS is reached. The isolation will occur as a result of
Main Steam Line Tunnel Temperature--High for any leak greater than
1% rated steam flow. Thus, a 40 psig reduction in the LPIS will not
adversely affect the peak temperature in the steam tunnel. In
addition, the dynamic effects (e.g., pipe whip and jet impingement)
on other structures, systems and components are unaffected by the
reduced LPIS.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of a high energy line
break accident previously evaluated.
Radiological Consequences
The MSIVs are assumed to close due to high steam line flow at
the start of an accident in the analysis. The low pressure isolation
trip is not used in the mass release analysis and the radiological
consequences are not affected by the reduction of the LPIS.
If the steam line break is too small to cause a high flow
isolation signal, MSIV closure may be initiated by another signal
(e.g., high steam tunnel temperature or low reactor water level) or
it may result from the low pressure isolation trip. Thus, a 40 psig
reduction in the LPIS will have no adverse impact on the
radiological consequences. The radiological consequences of a
reduction in the MSL LPIS are addressed further in Section 6 of this
attachment.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated for radiological consequences.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
General Electric Company (GE) evaluated the impact of reducing
the LPIS analytical limit from 825 to 785 psig, including analysis
of transient and safety related licensing bases for DNPS, Units 2
and 3, and QCNPS, Units 1 and 2. Current licensing bases events
remain bounding for ATWS, transient, and accident analyses. The
proposed change revises the allowable value of TS Table 3.3.6.1-1,
Function 1.b, but does not alter the instrumentation or control
logic of the Primary Containment Isolation System.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The revised LPIS does not change the current licensing bases
events, which remain bounding for ATWS, transient and accident
analyses. The conclusion that a reduction in the MSIV LPIS will not
have an adverse impact on plant accident analyses is valid. The LPIS
was analyzed by GE during the EPU project for impact on safety
limits and safety margins and was determined to be a non-impacted
item. 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. Edward J. Cullen, Vice President,
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
Kennett Square, PA 19348.
NRC Section Chief: Anthony J. Mendiola.
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: August 19, 2003.
Description of amendment request: The proposed amendments would
modify Technical Specification (TS) 5.5.13, ``Primary Containment
Leakage Rate Testing Program,'' by identifying a specific exception to
the testing guidance contained in Regulatory Guide (RG) 1.163,
``Performance-Based Containment Leak-Test Program.''
LaSalle County Station (LSCS) Units 1 and 2 conduct their leakage
rate testing of the primary containments to the requirements of 10 CFR
50.54(o) and 10 CFR part 50, Appendix J, Primary Reactor Containment
Leakage Testing for Water-Cooled Power Reactors,'' Option B as modified
by approved exemptions. Additionally, the program is in accordance with
the guidelines contained in RG 1.163. The proposed TS change would take
exception to RG 1.163 guidance by allowing the testing of potential
valve atmospheric leakage paths (e.g., valve stem packing), that are
not exposed to reverse direction Type B or C leakage test pressure
during the regularly scheduled Type A test. A list of the potential
valve atmospheric leakage paths, the leakage rate measurement method
and the acceptance criteria will be contained in the program. This
exception will be
[[Page 74267]]
applicable only to valves that are not isolable from the primary
containment free air space.
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 probability or consequences of an accident previously evaluated.
The proposed change will revise LaSalle County Station, Units I
and 2, Technical Specification (TS) 5.5.13, ``Primary Containment
Leakage Rate Testing Program'' by identifying a specific exception
to the testing guidance contained in Regulatory Guide (RG) 1.163,
``Performance-Based Containment Leak-Test Program.''
The function of the primary containment is to isolate and
contain fission products released from the reactor Primary Coolant
System (PCS) following a design basis Loss of Coolant Accident
(LOCA) and to confine the postulated release of radioactive material
to within limits. The probability of an accident previously
evaluated is not dependent on the test-frequency of the primary
containment Type A, B or C testing. The test interval associated
with primary containment testing is not a precursor of any accident
previously evaluated. The proposed specific exception to the testing
guidance contained in RG 1.163 will continue to test all potential
valve atmospheric leakage paths and will not be a precursor to a
Design Basis Accident (DBA). Containment testing does provide
assurance that the LaSalle County Station primary containments will
not exceed allowable leakage rate values specified in the Technical
Specifications and will continue to perform their design function
following an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed change does not affect the control
parameters governing unit operation or the response of plant
equipment to transient conditions. The proposed change does not
introduce any new equipment, modes of system operation or failure
mechanisms.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The integrity of the primary containment is verified through
Type B and Type C local leak rate tests (LLRTs) and the overall leak
tight integrity of the primary containment is verified by a Type A
integrated leak rate test (ILRT) as required by 10 CFR part 50,
Appendix J, ``Primary Reactor Containment Leakage Testing for Water-
Cooled Power Reactors.'' These tests are performed to verify the
essentially leak tight characteristics of the primary containment at
the design basis accident pressure. The proposed change for a
specific exception to the testing guidance contained in Regulatory
Guide (RG) 1.163 will continue to test all potential valve
atmospheric leakage paths and does not effect the test acceptance
criteria for Type A, B or C testing. Therefore, LSCS has determined
that the proposed change provides an equivalent level of protection
as that currently provided.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief : Anthony J. Mendiola.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station Unit No. 2, Oswego County, New York
Date of amendment request: November 20, 2003.
Description of amendment request: The licensee proposes to revise
the safety limit minimum critical power ratio (SLMCPR) values in
section 2.1.1.2 of the Technical Specifications (TSs). The SLMCPR
values are based on cycle-specific calculations done for the next fuel
cycle, Cycle 10, using methodology previously approved by the Nuclear
Regulatory Commission (NRC).
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. The NRC staff has reviewed the licensee's analysis
against the three standards of 10 CFR 50.92(c). The NRC staff's
analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The proposed SLMCPR values, calculated using an NRC-approved
methodology, will be made in a manner such that conservatism is
maintained through compliance with applicable NRC regulations and
guidance. No hardware design change is involved with the proposed
amendment, thus there will be no adverse effect on the functional
performance of any plant structure, system, or component (SSC). All
SSCs will continue to perform their design functions with no decrease
in their capabilities to mitigate the consequences of postulated
accidents. SLMCPR values were not previously factored into the
probability of accidents, nor were they factored into scenarios of
previously analyzed accidents. Accordingly, the revised SLMCPR values
will lead to no increase in the consequences of an accident previously
evaluated, and no increase of the probability of an accident previously
evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment is not the result of a hardware
design change, nor does it lead to the need for a hardware design
change. There is no change in the methods the unit is operated. As a
result, all SSCs will continue to perform as previously analyzed by the
licensee, and previously evaluated and accepted by the NRC staff.
Therefore, the proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the licensee did not propose to
exceed or alter a design basis or safety limit, the proposed amendment
will not affect in any way the performance characteristics and intended
functions of any SSC. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: November 21, 2003.
Description of amendment request: The proposed amendment would
allow the position of a rod to be monitored by
[[Page 74268]]
a means other than the movable incore detectors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. The proposed change provides an alternative method for the
monitoring of the position of a rod once the position of the rod is
verified using the moveable incore detector system. The proposed
monitoring of stationary gripper coil parameters provides a
reasonably similar approach to rod position monitoring as that
provided by the movable incore detector system. In particular, the
ability to immediately detect a rod drop or misalignment is not
directly provided by the movable incore detector system or by the
monitoring of stationary gripper coil parameters. Additionally,
neither the movable incore detector system, nor the monitoring of
stationary gripper coil parameters, provides the capability to
verify rod position following a reactor trip or shutdown. Therefore,
the monitoring of stationary gripper coil parameters, in lieu of the
use of the movable incore detector system, provides an equivalent
and acceptable method of monitoring rod position while a position
indicator is inoperable.
Therefore, 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.
(2) Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
No. As described above, the proposed change provides only an
alternative method of monitoring the position of a rod. No new
accident initiators are introduced by the proposed alternative
manner of performing rod position monitoring. The proposed change
does not affect the reactor protection system or the reactor control
system. Hence, no new failure modes are created that would cause a
new or different kind of accident from any accident previously
evaluated.
Therefore, 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.
(3) Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
No. The bases for TS (Technical Specification) 3.1.8 state that
the operability of the rod position indicators is required to
determine control rod positions and thereby ensure compliance with
the control rod alignment and insertion limits. The proposed change
does not alter the requirement to determine rod position but
provides an alternative method for monitoring the position of the
affected rod after the position of the rod is verified using the
moveable incore detector system. As a result, the initial conditions
of the accident analysis are preserved and the consequences of
previously analyzed accidents are unaffected.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
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, 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 NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: December 13, 2002, as
supplemented September 25, 2003.
Brief description of amendments: These amendments changed the
Technical Specifications (TSs) by removing the requirement to have the
charging pumps operable when thermal power is greater than 80% of rated
thermal power. The change also removes Surveillance Requirement 3.5.2.4
for verifying the required charging pump flow rate. The change to TS
3.5.2 does not modify any other charging pump requirements in the
Technical Requirements Manual (e.g., requirements of charging pump
availability for boration and cooldown remain in effect).
Date of issuance: December 3, 2003.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 260 and 237.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7812).
The September 25, 2003, supplemental letter provided clarifying
information that did not enlarge the scope of the amendment as noticed
in the original Federal Register notice or change the initial proposed
no significant hazards consideration determination.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated December 3, 2003.
No significant hazards consideration comments received: No.
[[Page 74269]]
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: May 28, 2003, as supplemented
November 25, 2003.
Brief description of amendments: These amendments changed the
reactor pressure vessel pressure-temperature limit cooldown curves in
the Calvert Cliffs 1 and 2 Technical Specifications by incorporating a
different range of temperatures for which a maximum cooldown rate of
100[deg]F/hr is acceptable.
Date of issuance: December 9, 2003.
Effective date: As of the date of issuance to be implemented within
120 days.
Amendment Nos.: 261 and 238.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40701).
The November 25, 2003, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated December 9, 2003.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No.
50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana.
Date of amendment request: March 14, 2003, as supplemented by
letter dated June 24, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.8.1, ``AC Sources--Operating,'' Surveillance
Requirements (SRs) pertaining to the testing of the Division 1 and 2
standby diesel generators (DGs). Specifically, the proposed changes
eliminate mode restrictions that previously prevented performance of
SRs during Modes 1 and 2 for the Division 1 and 2 DGs. The changes
allow the performance of SR 3.8.1.9 and SR 3.8.1.10 for the Division 1
and 2 DGs during any plant operating mode.
Date of issuance: November 7, 2003.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 137.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: (68 FR 18275). The June
24, 2003, 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 November 7, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: May 28, 2003, as supplemented on
June 24, 2003.
Brief description of amendment: The amendment revised Technical
Specification (TS) Section 3.4.3, ``RCS Pressure and Temperature (P/T)
Limits,'' and section 3.4.12, ``Low Temperature Overpressure Protection
(LTOP),'' to incorporate revised reactor pressure vessel P/T limits and
overpressure protection system limits to allow operation up to 20
effective full-power years. Specifically, the amendment changed TS
Figures 3.4.3-1 to 3.4.3-3 and TS Figures 3.4.12-1 to 3.4.12-4.
Date of issuance: December 3, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 220.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43389).
The June 24 letter provided clarifying information that did not
enlarge the scope of the amendment request or change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 3, 2003.
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 application for amendment: May 1, 2003, as supplemented by
letter dated September 30, 2003.
Brief description of amendment: The amendment modifies the
surveillance testing requirements for the containment spray system by
deleting the requirement to verify the position of valves that are
locked, sealed, or otherwise secured in their correct position (and by
deleting wording regarding the verified valves being positioned to take
suction from the refueling water tank), and replacing the quantitative
allowable pump degradation value with a requirement to verify the pumps
perform in accordance with the Inservice Testing Program.
Date of issuance: December 4, 2003.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 252.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28851).
The September 30, 2003, 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 December 4, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois
Date of application for amendments: February 27, 2003, as
supplemented on July 17, July 31, September 11, and November 25, 2003.
Brief description of amendments: The amendments revise Technical
Specification Section 3.4.9, ``Reactor Coolant System Pressure and
Temperature (P/T) Limits,'' incorporating revisions to the P/T limit
curves. The amendment also deletes the license conditions specified in
DNPS Unit 2 Facility Operating License Section 2.C(8) and DNPS Unit 3
Facility Operating License Section 3.P, ``Pressure-Temperature Limit
Curves.''
Date of issuance: November 26, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days of the date of issuance.
Amendment Nos.: 205/197.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Facility Technical Specifications and license conditions
specified in the Facility Operating Licenses.
Date of initial notice in Federal Register: August 5, 2003 (68 FR
46242).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 26, 2003.
No significant hazards consideration comments received: No.
[[Page 74270]]
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: November 26, 2002.
Brief description of amendments: These amendments revised TS
3.1.3.1, ``Control Rod Operability,'' by adding new Limiting Condition
for Operation criteria and applicable ACTION requirements for scram
discharge volume (SDV) vent and drain valves. The changes also modified
TS 3.6.3, ``Primary Containment Isolation Valves,'' to clarify the
relationship between TS 3.1.3.1 and TS 3.6.3 regarding SDV vent and
drain valves.
Date of issuance: November 26, 2003.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 168 and 131.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
803).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 26, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of application for amendments: March 26, 2003.
Brief description of amendments: These amendments modify Technical
Specifications (TSs) 4.0.1 and 4.0.3 to be consistent with the Improved
Standard Technical Specifications. The amendments also modify the TS
requirements for missed surveillances in TS 4.0.3 to be consistent with
the Nuclear Regulatory Commission-approved Technical Specification Task
Force (TSTF), Standard Technical Specification Change TSTF-358,
Revision 6.
Date of issuance: November 25, 2003.
Effective date: As of the date of its issuance and shall be
implemented within 60 days.
Amendment Nos.: 258 and 140.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37577).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 25, 2003.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: December 9, 2002, as
supplemented by letter dated August 28, 2003.
Brief description of amendments: The changes would revise Technical
Specification (TS) 3.75, ``Auxiliary Feedwater System,'' Surveillance
Requirement (SR) 3.7.5.2 Frequency. Specifically, the wording of the
Frequency of SR 3.7.5.2 would change from ``31 days on a Staggered Test
Basis'' to ``In accordance with the Inservice Testing Program.'' This
change is requested to implement recommendations of the Standard
Technical Specifications for Combustion Engineering Plants, NUREG-1432,
Revision 2.
Date of issuance: November 25, 2003.
Effective date: November 25, 2003, to be implemented within 60 days
of issuance.
Amendment Nos.: Unit 2--191; Unit 3--182.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
812).
The August 28, 2003, supplemental letter provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 25, 2003.
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: November 14, 2002, as supplemented by
letters dated October 30, and November 6, 2003.
Brief description of amendments: The amendments revise the Updated
Final Safety Analysis Report (UFSAR) to eliminate the turbine missile
design basis.
Date of issuance: December 2, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance. The UFSAR changes shall be implemented in
the next periodic update to the UFSAR in accordance with 10 CFR
50.71(e).
Amendment Nos.: Unit 1--158; Unit 2--146.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the UFSAR.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7821).
The October 30, and November 6, 2003, 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 NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register on February 18, 2003
(68 FR 7821).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 2, 2003.
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: May 22, 2003, as supplemented by letters
dated September 10 and September 30, 2003.
Brief description of amendments: The amendments change the
pressurizer safety valve lift tolerance, as specified in Technical
Specification (TS) 3.4.2.2, ``Reactor Coolant System,'' from plus/minus
(+/-) 2 percent (%) to +2% and -3%.
Date of issuance: December 2, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: Unit 1--159; Unit 2--147.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the TSs.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37583).
The September 10 and September 30, 2003, supplemental letters
provided clarifying information that was within the scope of the
original Federal Register notice (68 FR 37583) and did not change the
initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 2, 2003.
No significant hazards consideration comments received: No.
[[Page 74271]]
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: August 7, 2003.
Description of amendment request: The amendments modified Technical
Specification (TS) requirements for mode change limitations to adopt
Industry/TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility
in Mode Restraints.''
Date of issuance: December 1, 2003.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment Nos.: 249, 286 & 244.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68.
Amendments revised the TSs.
Date of initial notice in Federal Register: October 14, 2003 (68 FR
59221).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 1, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant
(SQN), Units 1 and 2, Hamilton County, Tennessee
Date of application for amendment: March 13, 2003, as supplemented
July 30, 2003.
Description of amendment: The amendment revises the boron
concentration requirements in Technical Specifications (TSs) 3.5.2,
``Cold Leg Accumulators,'' and 3.5.5, ``Refueling Water Storage Tank.''
The revised boron concentration requirement is a function of the number
of tritium producing burnable absorber rods (TPBARs) in the core.
Date of issuance: December 1, 2003.
Effective date: As of the date of issuance to be implemented no
later than startup from an outage in which TPBARs are loaded into the
reactor.
Amendment Nos.: 289 & 279.
Facility Operating License Nos. DPR-77 and DPR-79: Amendment
revised the TSs.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18286). The supplemental letter provided clarifying information only
and did not change the scope of the original amendment request or the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 1, 2003.
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 30-day 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, 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 Assess and
Management 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 NRC Public
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or
by email to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By January 22, 2004, the
licensee may file
[[Page 74272]]
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 2.714, 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 an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of 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. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses. 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of the continuing disruptions in delivery of mail to United
States government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the petition for
leave to intervene and request for hearing should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States government offices, 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 filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Southern Nuclear Operating Company, Inc., et al., Docket No. 50-425,
Vogtle Electric Generating Plant, Unit 2, Burke County, Georgia
Date of amendment request: November 5, 2003.
Description of amendment request: The proposed amendment would
extend the surveillance interval for the Memories Test portion of the
Actuation Logic Test for: (1) Power Range Block (Switch position 1),
(2) Intermediate Range Block (Switch position 2), (3) Source Range
Block (Switch positions 3 and 4), (3) Safety Injection (SI) Block,
Pressurizer (Switch positions 5 and 6), (4) SI Block, High Steam
Pressure Rate (Switch positions 7 and 8), (5) Auto SI Block (Switch
position 9), and (6) Feedwater Isolation on P14 or SI (Switch positions
10 and 11). In addition to the functions listed above, the licensee is
requesting an extension of the surveillance interval for the portions
of the Actuation Logic Test for Feedwater Isolation on P14 or SI that
pass through the memories circuits and the Power Range block of the
Source Range Trip test for the Unit 2 Train B Solid State Protection
System to the next refueling outage at the end of Cycle 10 or the next
Unit 2 shutdown to MODE 5, whichever comes first.
Date of issuance: December 3, 2003.
Effective date: December 3, 2003.
Amendment No.: 108.
Facility Operating License No. NPF-81: Amendment revises the
technical specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. November 18, 2003 (68 FR 65092). The notice
provided an opportunity to submit
[[Page 74273]]
comments on the Commission's proposed NSHC determination. No comments
have been received. The notice also provided an opportunity to request
a hearing by December 18, 2003, but indicated that if the Commission
makes a final NSHC determination, any such hearing would take place
after issuance of the amendment. The Commission's related evaluation of
the amendment, finding of exigent circumstances, state consultation,
and final NSHC determination are contained in a safety evaluation dated
December 3, 2003.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: John A. Nakoski.
Dated in Rockville, Maryland, this 15th day of December, 2003.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-31314 Filed 12-22-03; 8:45 am]
BILLING CODE 7590-01-P