[Federal Register Volume 71, Number 10 (Tuesday, January 17, 2006)]
[Notices]
[Pages 2586-2600]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-320]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 22, 2005 to January 5, 2006. The
last biweekly notice was published on January 3, 2006 (71 FR 145).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this
[[Page 2587]]
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-
[[Page 2588]]
mail to [email protected]. A copy of the request for hearing and
petition for leave to intervene should also be sent to the attorney for
the licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station (OCNGS), Ocean County, New Jersey
Date of amendment request: December 2, 2005.
Description of amendment request: The amendment would revise the
Technical Specifications to increase the allowable as-found main steam
safety valve code safety function lift setpoint tolerance from 1% to 3%.
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 the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
Response: No.
The proposed TS changes allow for an increase in the as-found
Main Steam Safety Valve (MSSV) setpoint tolerance from 1% to 3%. The proposed changes do not alter the
MSSV nominal lift setpoints or MSSV lift setpoint test frequency.
The proposed TS changes have been evaluated on both a generic
and plant specific basis. The NRC has approved the general approach
of this change; however, implementation is contingent on several
plant specific evaluations. The required plant specific analyses and
evaluations included transient analysis of the anticipated
operational transients (AOTs); analysis of the design basis
overpressurization event; evaluation of the performance of high
pressure systems, and evaluation of the containment response during
Loss-of-Coolant Accident (LOCA) and hydrodynamic loads on the MSSV
discharge lines and containment. These analyses and evaluations
demonstrate that there is adequate margin to the design core thermal
limits and reactor vessel pressure limits using the 3%
MSSV as-found setpoint tolerance. The analyses and evaluations also
demonstrate that the operation of high-pressure safety systems will
not be adversely affected and that the containment response during a
LOCA will be acceptable.
Evaluations of the impact of the proposed change on the
equipment important to safety have been performed and no adverse
conditions were identified. The reactor pressure vessel and attached
systems and piping have been evaluated for the impact of this
proposed TS change. A plant specific analysis has been performed
which indicates that the ASME Code upset limits for the reactor
pressure vessel will not be exceeded for the limiting event, i.e.,
Main Steam Isolation Valve (MSIV) closure with flux Scram. The
reactor pressure vessel and attached piping design values will not
be exceeded. Therefore, the probability of a malfunction of the
reactor pressure vessel and attached systems and piping is not
increased and the consequences of such an accident remain
acceptable.
The nuclear fuel has been evaluated for the impact of the
proposed change.
Plant specific analyses were performed which indicate that for
all abnormal operational transients adequate margin to the fuel
thermal limit parameters, i.e., Minimum Critical Power Ratio (MCPR)
and thermal-mechanical limits, is maintained. Emergency Core Cooling
System (ECCS)/LOCA performance is maintained adequate to meet the
requirements of 10 CFR 50.46. Therefore, the consequences of these
accidents remain acceptable and the probability of the malfunction
of the nuclear fuel is not increased.
The Containment response during a LOCA has been evaluated for
the impact of the proposed change. The major factor in the
Containment pressure response to a LOCA is the rate of reactor
vessel water inventory loss due to a DBA LOCA. The rate of reactor
vessel water inventory loss is mainly dependent on the initial
reactor pressure, which is not affected by the proposed setpoint
tolerance change. The major factor in the Containment temperature
response to a LOCA is the integrated steam inventory loss due to
Main Steamline Break. The rate of reactor vessel steam inventory
loss is mainly dependent on the reactor decay heat, which is not
affected by the proposed setpoint tolerance change. Therefore, the
consequences of these accidents remain acceptable and the
probability of the malfunction of Containment is not increased.
The Control Rod Drive (CRD) system has been evaluated for the
impact of the proposed change. The CRD system capability of
controlling reactor power during normal plant operation and rapidly
inserting control rod blades (Scram) during abnormal plant
conditions is not impacted by the proposed change. Therefore, the
probability of a malfunction of the CRD system is not increased.
The Reactor Vessel Instrumentation System has been evaluated for
the impact of the proposed change. The Reactor Vessel
Instrumentation System will continue to be operated within the
current design pressure/temperature requirements; therefore, the
probability of a malfunction of the Reactor Vessel Instrumentation
System is not increased.
An administrative change is also being proposed to correct the
reference to ``IWV-3510 of Section XI of the ASME Boiler and
Pressure Vessel Code'' in TS 4.3.E because the stated ASME section
no longer exists. The TS is being changed to reference specification
4.3.C for MSSV testing. This is an administrative change and does
not affect previously evaluated accidents.
Therefore, the proposed TS changes do not significantly increase
the probability or consequences of an accident previously evaluated.
2. Will operation of the facility in accordance of the proposed
amendment create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No.
The proposed TS changes allow for an increase in the as-found
MSSV setpoint tolerance from 1% to 3%.
Generic and plant specific analyses and evaluations indicate that
the plant response to any previously evaluated event will remain
acceptable. All plant systems, structures, and components will
continue to be capable of performing their required safety function
as required by event analysis guidance.
The proposed TS changes do not alter the MSSV nominal lift
setpoints or MSSV lift setpoint test frequency. The operation and
response of the affected equipment important to safety is unchanged.
All systems, structures, and components will continue to be operated
within acceptable operating and/or design parameters. No system,
structure, or component will be subjected to a condition that has
not been evaluated and determined to be acceptable using the
guidance required for specific event analysis.
The change to correct the reference to ``IWV-3510 of Section XI
of the ASME Boiler and Pressure Vessel Code'' in TS 4.3.E is an
administrative change and does not affect the possibility of a new
or different kind of accident.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
identified.
3. Will operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
Response: No.
The proposed TS changes allow for an increase in the as-found
MSSV setpoint tolerance from 1% to 3%. The
proposed TS changes do not alter the MSSV nominal lift setpoints or
MSSV lift setpoint test
[[Page 2589]]
frequency. The operation and response of the affected equipment
important to safety is unchanged. All systems, structures, and
components will continue to be operated within acceptable operating
and/or design parameters. While the calculated peak reactor vessel
pressure for the ASME overpressure event is higher than that
calculated without the increase in setpoint tolerance, it is still
within the respective licensing acceptance limits associated with
this event. These licensing acceptance limits have been determined
by the NRC to provide a sufficient margin of safety.
The increase in MSSV steam flow and reactor vessel pressure does
not reduce the margin of safety associated with the MSSVs and
associated components and structures since the increased MSSV steam
flow rate and reactor vessel pressure are bounded by the current
design analysis.
The margin of safety for fuel thermal limits and 10 CFR 50.46
limits are unaffected by the proposed change.
The margin of safety for the Containment is unaffected by the
proposed change.
The capability of the SLC system and the CRD system to perform
their safety functions during all required events, using the
required guidance for event analysis, is maintained. Therefore, the
proposed changes do not reduce the margin of safety provided by the
SLC and CRD systems.
The change to correct the reference to ``IWV-3510 of Section XI
of the ASME Boiler and Pressure Vessel Code'' in TS 4.3.E is an
administrative change and does not affect the margin of safety.
Therefore, these proposed TS 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: Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Darrell J. Roberts.
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 amendments request: November 3, 2005.
Description of amendments request: The proposed amendments would
revise the accident source term in the design-basis radiological
consequences analyses and the associated Technical Specifications
(TSs), pursuant to section 50.67 of part 50 of Title 10 of the Code of
Federal Regulations (10 CFR 50.67). The proposed amendments would
provide for the full implementation of the alternate source term (AST)
in accordance with the guidance in Regulatory Guide 1.183,
``Alternative Radiological Source Terms for Evaluating Design Basis
Accidents at Nuclear Power Reactors.'' The proposed amendments would
also increase the flow rate for the control room emergency ventilation
system (CREVS) from 2000 to 10000 cubic feet per minute in TS 5.5.11,
``Ventilation Filter Testing Program,'' by means of a modification to
the CREVS. In addition, automatic isolation dampers and radiation
monitors will also be installed at access control heating, ventilating,
and air conditioning (HVAC) unit no. RTU-1 and access control air
conditioning unit no. 13.
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 results of the applicable radiological design basis
accidents (DBAs) re-evaluation demonstrated that, with the requested
changes, the dose consequences of these limiting events are within
the regulatory limits and guidance provided by the Nuclear
Regulatory Commission in 10 CFR 50.67 and Regulatory Guide 1.183 for
AST methodology. The AST is an input to calculations used to
evaluate the consequences of an accident and does not by itself
affect the plant response or the actual pathway of the activity
released from the fuel. It does, however, better represent the
physical characteristics of the release such that appropriate
mitigation techniques may be applied.
The change from the original source term to the new proposed AST
is a change in the analysis method and assumptions and has no effect
on accident initiators or causal factors that contribute to the
probability of occurrence of previously analyzed accidents. Use of
an AST to analyze the dose effect of DBAs shows that regulatory
acceptance criteria for the new methodology continues to be met.
Changing the analysis methodology does not change the sequence or
progression of the accident scenario.
The proposed Technical Specification changes reflect the plant
configuration that will either support implementation of the AST
analyses or eliminate requirements that are no longer needed as a
result of the revised DBA analyses. The equipment affected by the
proposed changes is mitigative in nature and relied upon after an
accident has been initiated. The operation of various filtration
systems have been considered in the evaluations for these proposed
changes. While the operation of some systems does change with the
implementation of an AST, the affected systems are not accident
initiators; and application of the AST methodology, itself, is not
an initiator of a DBA.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
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 accident previously evaluated.
As described in Item 1 above, the changes proposed in this
license amendment request involve the use of a new analysis
methodology and related regulatory acceptance criteria. The proposed
Technical Specification changes reflect the plant configuration that
will either support implementation of the new methodology or
eliminate requirements that are no longer needed as a result of the
new methodology. No new or different accidents result from utilizing
the proposed changes. Although the proposed changes require
modification to the Control Room emergency ventilation system and
installation of automatic isolation dampers and radiation monitors
at Access Control HVAC Unit RTU-1 and Access Control Air
Conditioning Unit 13 on the Auxiliary Building roof, none of these
changes can initiate a new or different kind of accident since they
are only related to system capabilities that provide protection from
accidents that have already occurred. As a result, no new failure
modes are being introduced that could lead to different accidents.
These changes do not alter the nature of events postulated in the
Updated Final Safety Analysis Report nor do they introduce any
unique precursor mechanisms.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
As described in Item 1 above, the changes proposed in this
license amendment request involve the use of a new analysis
methodology and related regulatory acceptance criteria. The proposed
Technical Specification changes reflect the plant configuration that
will either support implementation of the new methodology or
eliminate requirements that are no longer needed as a result of the
new methodology. Safety margins and analytical conservatisms have
been evaluated and have been found acceptable. The analyzed events
have been carefully selected and, with plant modification, margin
has been retained to ensure that the analyses adequately bound
postulated event scenarios. The analyses have been performed using
conservative methodologies, as specified in Regulatory Guide 1.183.
The dose consequences of these DBAs remain within the acceptance
criteria presented in 10 CFR 50.67, ``Accident Source Term,'' and
Regulatory Guide 1.183. The proposed changes continue to ensure that
the doses at the exclusion area boundary and low population zone
boundary, as well as the Control Room, are within corresponding
regulatory limits.
[[Page 2590]]
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 amendments request involves no significant hazards
consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt
Street, 17th floor, Baltimore, MD 21202.
NRC Branch Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of amendment request: December 14, 2005.
Description of amendment request: The proposed amendment
modifies the Technical Specifications (TSs) to incorporate a revised
Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO
SLMCPR) due to the cycle-specific analysis.
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 derivation of the cycle specific Single Loop Operation
Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) for
incorporation into the Technical Specifications (TS), and its use to
determine cycle-specific thermal limits, has been performed using
the methodology discussed in ``General Electric Standard Application
for Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and U.S.
Supplement, NEDE-24011-P-A-15-US, September, 2005, which includes
Amendment 25. Amendment 25 was approved by the NRC in a March 11,
1999 safety evaluation report.
The basis of the SLO SLMCPR calculation is to ensure that
greater than 99.9% of all fuel rods in the core avoid transition
boiling if the limit is not violated. The new SLO SLMCPR preserves
the existing margin to transition boiling. The GE-14 fuel is in
compliance with Amendment 22 to ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and
U.S. Supplement, NEDE-24011-P-A-15-US, September 2005, which
provides the fuel licensing acceptance criteria. The probability of
fuel damage will not be increased as a result of this change.
Therefore, the proposed TS 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 SLO SLMCPR is a TS numerical value, calculated to ensure
that transition boiling does not occur in 99.9% of all fuel rods in
the core if the limit is not violated. The new SLO SLMCPR is
calculated using NRC approved methodology discussed in ``General
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-15
(GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September
2005, which includes Amendment 25. Additionally, the GE-14 fuel is
in compliance with Amendment 22 to ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and
U.S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which
provides the fuel licensing acceptance criteria. The SLO SLMCPR is
not an accident initiator, and its revision will 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.
There is no significant reduction in the margin of safety
previously approved by the NRC as a result of the proposed change to
the SLO SLMCPR, which includes the use of GE-14 fuel. The new SLO
SLMCPR is calculated using methodology discussed in ``General
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-15
(GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September,
2005, which includes Amendment 25. The SLO SLMCPR ensures that
greater than 99.9% of all fuel rods in the core will avoid
transition boiling if the limit is not violated when all
uncertainties are considered, thereby preserving the fuel cladding
integrity.
Therefore, the proposed TS change will not involve a significant
reduction in [a] margin of safety previously approved by the NRC.
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. Brad Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Darrell J. Roberts.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: December 21, 2005.
Description of amendment request: The proposed amendment revises
the Technical Specifications by relocating the Pressure Isolation
Valve (PIV) tables to the Technical Requirements Manual (TRM).
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 relocation of Technical Specification Table
3.4.3.2-1 does not alter the requirements for pressure isolation
valve operability or surveillance currently in the Technical
Specifications. The proposed change to remove the pressure isolation
valve table from TS and relocate the information to an
administratively controlled document, and to revise the wording in
TS to reflect this change, will have no impact on any safety related
structures, systems or components. The probability of occurrence of
a previously evaluated accident is not increased because this change
does not introduce any new potential accident initiating conditions.
The consequences of accidents previously evaluated in the UFSAR
[Updated Final Safety Analysis Report] are not affected because the
ability of the PIVs to limit leakage through these valves in amounts
that do not compromise safety is not affected. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are administrative in nature and do not
result in physical alterations or changes in the method by which any
safety related system performs its intended function(s). The
proposed changes do not impact any safety analysis assumptions. The
proposed changes do not create any new accident initiators or
involve an activity that could be an initiator of an accident of a
different type.
All PIVs and alarm instrumentation will continue to be tested to
the same rigorous requirements as defined in the Technical
Specification Surveillance Requirements. The proposed revision does
not make changes in any method of testing or how any safety related
system performs its safety functions. Therefore, the possibility of
an accident of a different type than any previously evaluated in the
UFSAR is not created.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The administrative change to relocate Technical Specification
Table 3.4.3.2-1 to the Technical Requirements Manual does not alter
the basic regulatory requirement for Reactor Coolant System pressure
isolation and will not affect the isolation capability for credible
accident scenarios. Future revisions to the Technical Requirements
Manual Table will be subject to evaluation pursuant to 10 CFR 50.59.
Additionally, the proposed relocation does not alter the
requirements for pressure isolation valve and alarm instrumentation
operability currently in the Technical Specifications. The LCO
[limiting condition for operation] and Surveillance Requirements
will be retained in the revised Technical Specifications. The
proposed change will not affect the meaning, application, and
function of the current Technical Specification requirements for the
valves in Table 3.4.3.2-1. Therefore, the proposed changes do not
result in a significant reduction in [a] margin of safety.
[[Page 2591]]
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. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Darrell J. Roberts.
Exelon Generation Company, LLC (EGC, licensee), Docket No. 50-265, Quad
Cities Nuclear Power Station (QCNPS), Unit 2, Rock Island County,
Illinois
Date of amendment request: December 15, 2005.
Description of amendment request: The proposed change revises the
values of the safety limit minimum critical power ratio (SLMCPR) in
Technical Specification (TS) section 2.1.1, ``Reactor Core SLs.''
Specifically, the proposed change would require that for Unit 2, the
minimum critical power ratio (MCPR) for Global Nuclear Fuel (GNF) fuel
shall be >=1.09 for two recirculation loop operation, or >=1.10 for
single recirculation loop operation. Additionally, the proposed change
would require that MCPR for Westinghouse fuel shall be >=1.11 for two
recirculation loop operation, or >=1.13 for single recirculation loop
operation.
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:
According to 10 CFR 50.92, ``Issuance of amendment,'' paragraph
(c), a proposed amendment to an operating license involves no
significant hazards consideration if 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.
EGC has evaluated the proposed change to the TS for QCNPS, Unit
2, using the criteria in 10 CFR 50.92, and has determined that the
proposed change does not involve a significant hazards
consideration. The following information is provided to support a
finding of no significant hazards consideration.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
change conservatively establishes the SLMCPR for QCNPS, Unit 2,
Cycle 19 such that the fuel is protected during normal operation and
during plant transients or anticipated operational occurrences
(AOOs).
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
The proposed change revises the SLMCPR to protect the fuel
during normal operation as well as during plant transients or AOOs.
Operational limits will be established based on the proposed SLMCPR
to ensure that the SLMCPR is not violated. This will ensure that the
fuel design safety criterion (i.e., that at least 99.9% of the fuel
rods do not experience transition boiling during normal operation
and AOOs) is met. Since the proposed change does not affect
operability of plant systems designed to mitigate any consequences
of accidents, the consequences of an accident previously evaluated
are not expected to increase.
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.
Creation of the possibility of a new or different kind of
accident would require creating one or more new accident precursors.
New accident precursors may be created by modifications of plant
configuration, including changes in allowable modes of operation.
The proposed change does not involve any plant configuration
modifications or changes to allowable modes of operation. The
proposed change to the SLMCPR assures that safety criteria are
maintained for QCNPS, Unit 2, Cycle 19.
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 SLMCPR provides a margin of safety by ensuring that at least
99.9% of the fuel rods do not experience transition boiling during
normal operation and AOOs if the MCPR limit is not violated. The
proposed change will ensure the appropriate level of fuel protection
by continuing to ensure that at least 99.9% of the fuel rods do not
experience transition boiling during normal operation and AOOs if
the MCPR limit is not violated. Additionally, operational limits
will be established based on the proposed SLMCPR to ensure that the
SLMCPR is not violated. This will ensure that the fuel design safety
criteria (i.e., that no more than 0.1% of the rods are expected to
be in boiling transition if the MCPR limit is not violated) are met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based upon the above, EGC concludes that the proposed amendment
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.
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. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Acting Branch Chief: Mindy S. Landau.
First Energy Nuclear Operating Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio
Date of amendment request: November 21, 2005.
Description of amendment request: The proposed amendment would
revise the acceptance criteria of Technical Specification (TS)
Surveillance Requirements (SRs) associated with TS 3.8.1, ``AC
Sources--Operating,'' to modify the Emergency Diesel Generator (EDG)
start tests to provide minimum voltage and frequency limits and clarify
other limits as steady state parameters. Specifically, the amendment
would revise SRs 3.8.1.2, 3.8.1.7, 3.8.1.12, 3.8.1.15 and 3.8.1.20.
This change is consistent with the approved Technical Specification
Task Force Traveler (TSTF) 163, Revision 2.
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 is a LAR (license amendment request) that
modifies the acceptance criteria for the PNPP TS SRs pertaining to
the EDGs. The EDGs mitigate the consequences of previously evaluated
[[Page 2592]]
accidents involving a loss of offsite power. The EDGs are used to
support mitigation of the consequences of an accident, but they are
not considered as the initiator of any previously analyzed accident.
The proposed LAR does not change the manner in which the EDGs
are operated and when implemented will continue to ensure the EDGs
perform their function when called upon. The proposed revision to
the TS SRs will continue to ensure that minimum frequency and
voltage are attained within the required time. The SRs will continue
to ensure that proper steady state voltage and frequency are
attained consistent with proper EDG governor and voltage regulator
performance.
The proposed LAR does not affect the design of the EDGs, the
operational characteristics of the EDGs, the interfaces between the
EDGs and other plant systems, the function, or reliability of the
EDGs. Thus, the EDGs will be capable of performing their accident
mitigation function and there is no impact to the radiological
consequences of any accident analysis.
As such, the proposed change continues to provide adequate
assurance of operable EDGs and does not involve any increase to the
probability or consequences of an accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed LAR introduces no new mode of plant operation and
it does not involve physical modification to the plant. New
equipment is not installed with the proposed LAR, nor does the
proposed LAR cause existing equipment to be operated in a new or
different manner.
Since the proposed changes do not involve a change to the plant
design or operation, no new system interactions are created by this
change. The proposed LAR does not produce any parameters or
conditions that could contribute to the initiation of accidents
different from those already evaluated in the Updated Safety
Analysis Report.
The changes to the affected TS SRs do not affect the assumed
accident performance of the EDGs, nor any plant structure, system or
component previously evaluated.
Therefore, the proposed LAR does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The proposed change is a LAR that does not impact EDG
performance, including the capability for each EDG to attain and
maintain required voltage and frequency for accepting and supporting
plant safety loads within the required time, as assumed in the plant
safety analysis.
The proposed LAR does not involve a significant reduction in a
margin of safety since the operability of the EDGs continues to be
determined as required to support the capability of the EDGs to
provide emergency power to plant equipment that mitigate the
consequences of an accident.
The proposed LAR does not introduce changes to setpoints or
limits established or assumed by the accident analysis. Therefore,
implementation of the proposed LAR 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 W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Mindy Landau, Acting.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: December 6, 2005.
Description of amendment request: The proposed amendment would
revise the Seabrook Station, Unit No. 1 Technical Specification
3.8.3.1, ``Onsite Power Distribution,'' to extend the allowed outage
time for balance-of-plant vital inverters 1-EDE-I-1E and 1-EDE-I-1F
from 24 hours to 7 days.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change extends the allowed outage time (AOT) for
the balance-of-plant (BOP) instrument bus inverters from 24 hours to
7 days. The BOP instrument bus inverters do not solely support any
risk-significant functions. The failure of an inverter is not an
initiator of any analyzed event and does not increase the frequency
of an initiating event. Consequently, extending the AOT will not
have an impact on the frequency of occurrence of any event
previously analyzed. The proposed change does not alter the design,
configuration, operation, or function of any plant system,
structure, or component. As a result, the outcomes of previously
evaluated accidents are unaffected. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed changes do 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 change does not challenge the performance or integrity
of any safety-related system. The proposed change neither installs
nor removes any plant equipment, not alters the design, physical
configuration, or mode of operation of any plant structure, system,
or component. Installed equipment will not be operated in a new or
different manner. No physical changes are being made to the plant,
so no new accident causal mechanisms are being introduced.
Procedures that ensure the unit operates within analyzed limits and
procedures that respond to off-normal and emergency conditions are
not altered with this proposed change. Therefore, the proposed
change does not create the possibility of a new or different
accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in [a] margin of safety.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change does not alter the
design, configuration, operation, or function of any plant system,
structure, or component. The ability of any operable structure,
system, or component to perform its designated safety function is
unaffected by this change. Operation with one instrument bus
inverter inoperable and the associated instrument bus aligned to its
maintenance supply does not result in a significant reduction in [a]
margin of safety. Surveillance testing of the emergency diesel
generators (EDGs) and the electrical distribution system provides
confidence that the EDGs will energize the emergency AC buses
following a loss of power. 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: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Darrell J. Roberts.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: November 12, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.7, ``Inservice Testing
Program,'' and TS 5.5.8, ``Steam Generator (SG) Tube Surveillance
Program,'' to update references to the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code (Code) and certain
associated periodicities for inservice testing activities consistent
with the requirements of Title 10 of the Code of Federal Regulations
(10 CFR) section 50.55a, ``Codes and standards.''
[[Page 2593]]
The proposed amendment would also correct a typographical error
contained in TS 5.5.8.b.2.
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 Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
The proposed change revises Technical Specifications for
consistency with the requirements of 10 CFR 50.55a(f)(4) and 10 CFR
50.55a(g)(4).
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing pumps and
valves.
The proposed change does not involve any hardware changes, nor
does it affect the probability of any event initiators. There will
be no change to normal plant operating parameters, engineered safety
feature actuation setpoints, accident mitigation capabilities, or
accident analysis assumptions or inputs.
Therefore, the probability or consequences of any accident
previously evaluated will not be significantly increased as a result
of the proposed change.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing. The
proposed change does not involve a modification to the physical
configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure.
Equipment important to safety will continue to operate as
designed. The changes do not result in any event previously deemed
incredible been made credible. The changes do not result in adverse
conditions or result in any increase in the challenges to safety
systems. Therefore, operation of the Point Beach Nuclear Plant in
accordance with the proposed amendment will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing. The safety
function of the affected components will be maintained.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences. The
proposed amendment will not otherwise affect the plant protective
boundaries, will not cause a release of fission products to the
public, nor will it degrade the performance of any other structures,
systems or components (SSCs) important to safety. Therefore, the
requested change will not result in 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: October 11, 2005.
Description of amendment request: The proposed amendment would
revise certain 18-month Technical Specification (TS) Surveillance
Requirements (SRs) to eliminate the condition that testing be conducted
during shutdown.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes permit PSEG to evaluate the conditions
required to safely perform a TS SR. These surveillance tests verify
that equipment will perform its intended safety function of
mitigating an accident. No analyzed accident scenario is being
revised. The initiating conditions and assumptions for accidents
described in the Hope Creek Generating Station Updated Final Safety
Analysis Report (UFSAR) remain as previously analyzed.
The proposed changes do not reduce the ability of the mitigating
equipment to perform its safety function. The TS will continue to
require the surveillance tests to be performed on an eighteen-month
periodicity to verify operability. As a result, the ability of the
mitigating equipment to perform its safety function is unaffected by
the proposed change.
The capitalization change is proposed to improve readability and
does not alter any requirement.
Based upon the above, the proposed changes will not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
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 do not create the possibility of a new or
different kind of accident from any accident previously evaluated in
the UFSAR. No new accident scenarios, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes. Specifically, no new hardware is being added to the plant
as part of the proposed change, no existing equipment is being
modified, and no significant changes in operations are being
introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes will not alter any assumptions, initial
conditions, or results of any accident analyses. The proposed
changes to remove the requirement to perform certain testing during
shutdown conditions allows PSEG to evaluate the conditions needed to
safely perform the required testing. There is no change to the
frequency of testing or in the testing that is required. There is no
change in the responsibility of PSEG to perform tests in a safe and
responsible manner. Any changes to procedures will have to be
individually evaluated to ensure that they do not reduce the margin
of safety. The changes do not affect the ability of systems,
structures or components to perform their safety related functions.
In addition, the proposed changes do not affect the ability of the
safety systems to ensure that the facility can be maintained in a
shutdown or refueling condition for extended periods of time.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Darrell J. Roberts.
[[Page 2594]]
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: August 31, 2005; as supplemented
December 8, 2005.
Description of amendment request: The proposed amendment would
relocate the containment high range accident monitors from the
radiation monitoring instrumentation technical specification (TS) to
the accident monitoring TS and correct a typographical error contained
in a previous amendment.
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 presents no change in the probability of a
previously evaluated accident.
The proposed change presents no change in the consequence of an
accident, since the containment high range accident monitors are
used post-accident to determine the amount of core damage and status
of the fission product barriers.
The containment high range accident monitors are used post
accident to assess the conditions inside containment. They have an
automatic function to switch the subcooling margin monitor (SCMM) to
``adverse'' mode (i.e., it displays a more conservative indication
of the amount of subcooling in the RCS) [reactor coolant system].
Additionally, the containment high range accident monitors provide
an indication that is used post accident in determining the status
of the fission product barriers. There will be no change in the
operation or use of the containment high range accident monitors.
The remaining change is editorial in nature and does not impact
the accident analysis in any manner.
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 is a minor change that is administrative in
nature. No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
No new hardware is added, existing hardware is not modified and no
significant changes in operations are implemented. Post accident
monitoring instrumentation is not associated with the initiation of
an accident.
3. Does the proposed change involve a significant reduction in
[a] margin of safety?
Response: No.
The proposed change does not alter the manner in which safety
limits, limiting safety systems settings or limiting conditions for
operation are determined. The proposed change will not alter any
assumptions, initial conditions or results specified in any accident
analysis.
There is no change in the containment high range accident
monitor high level alarm setpoint. The ECS [electronic check source]
is functionally equivalent to the TS definition of SOURCE CHECK.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Darrell J. Roberts.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey
Date of amendment request: September 21, 2005.
Description of amendment request: The amendment would change the
scope of steam generator (SG) tube inspections required in the SG
tubesheet region.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Of the various accidents previously evaluated, 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 Westinghouse 51 Series SGs has shown that axial
loading of the tubes is negligible during an SSE.
PSEG's amendment request takes credit for how the tubesheet
enhances the tube integrity in the Westinghouse Electric Company
explosive tube expansion (WEXTEX) region by precluding tube
deformation beyond its initial expanded outside diameter. For the
SGTR and SLB events, the required structural margins of the SG tubes
will be maintained due to the presence of the tubesheet. Tube
rupture is precluded for axial cracks in the WEXTEX region due to
the constraint provided by the tubesheet. Therefore, the normal
operating 3[Delta]P margin and the postulated accident
1.43[Delta]P margin against burst are maintained.
The W* length supplies the necessary resistive force to preclude
pullout loads under both normal operating and accident conditions.
The contact pressure results from the WEXTEX expansion process,
thermal expansion mismatch between the tube and tubesheet, and from
the differential pressure between the primary and secondary side.
Therefore, the proposed change results in no significant increase in
the probability or the occurrence of an SGTR or SLB accident.
The proposed changes do not affect other systems, structures,
components or operational features. Therefore, based on the above
evaluation, the proposed changes do not involve a significant
increase in the probability of an accident previously evaluated.
The consequences of an SGTR event are primarily affected by the
primary-to-secondary flow rate and the time duration of the primary-
to-secondary flow during the event. Primary-to-secondary flow rate
through a postulated ruptured tube (i.e., complete severance of a
single SG tube) is not affected by the proposed change since the
flow rate is based on the inside diameter of a[n] SG tube and the
pressure differential. PSEG's amendment request does not change
either of these. The duration of primary-to-secondary leakage is
based on the time required for an operator to determine that a[n]
SGTR has occurred, the time to identify and isolate the faulted SG,
and ensure termination of radioactive release to the atmosphere from
the faulted SG. PSEG's amendment request does not affect the
duration of the primary-to-secondary leakage because it does not
change the control room indicators with which an operator would
determine that an SGTR has occurred. The consequences of an SGTR are
secondarily affected by primary-to-secondary leakage, which could
occur due to axial cracks remaining in service in the WEXTEX region
in a non-faulted SG. During a[n] SGTR, the primary-to-secondary
differential pressure is less than or equal to the normal operating
differential pressure; therefore, the primary-to-secondary leakage
due to axial cracks in the WEXTEX region of a non-faulted SG during
a[n] SGTR would be less than or equal to the primary-to-secondary
leakage experienced during normal operation. Primary-to-secondary
leakage is considered in the calculation determining the
consequences of a[n] SGTR and the value is bounding.
The postulated SLB has the greatest primary-to-secondary
pressure differential, and therefore could experience the greatest
primary-to-secondary leakage. PSEG's amendment request requires the
aggregate leakage, (i.e., the combined leakage for the tubes with
service induced degradation inside the tubesheet) to remain below
the maximum allowable SLB primary-to-secondary leakage rate limit
such that the doses are maintained to less than the 10 CFR [Part]
100 limits and also less than the GDC-[General Design Criterion]19
limits.
[[Page 2595]]
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.
PSEG's amendment request does not introduce any physical changes
to the Salem Unit 2 SGs. PSEG's amendment request takes credit for
how the tubesheet enhances the SG tube integrity in the WEXTEX
region. Because degradation detected within the W* distance are
required to be plugged, it is highly unlikely that a tube would fail
as a result of a circumferential defect. Therefore a tube severance,
which would strike neighboring tubes and create a multiple tube
rupture, is not credible. The proposed change does not introduce any
new equipment or any change to existing equipment. No new effects on
existing equipment are created. Based on the above evaluation, the
proposed 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 amendment request maintains the structural margins of the SG
tubes for both normal and accident conditions that are required by
Regulatory Guide 1.121. For cracking located within the tubesheet,
tube burst is precluded due to the presence of the tubesheet. WCAP-
14797, Revision 2 defines a length W* of degradation free expanded
tubing, that provides the necessary resistance to tube pullout due
to the pressure induced forces (with applicable safety factor
applied). Application of the W* methodology 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 W*
criteria.
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 Updated Final Analysis Report or
Technical Specifications. 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Darrell J. Roberts.
PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric
Station, Unit 1 (SSES 1), Luzerne County, Pennsylvania
Date of amendment request: December 1, 2005.
Description of amendment request: The proposed amendment would
change the SSES-1 Technical Specifications (TSs) by revising the Unit 1
Cycle 15 (U1C15) minimum critical power ratio (MCPR) safety limit for
single loop operation in section 2.1.1.2 and references listed in TS
5.6.5.b.
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed change to the single-loop MCPR Safety Limit does
not directly or indirectly affect any plant system, equipment,
component, or change the processes used to operate the plant.
Further, the proposed U1C15 MCPR Safety Limit was generated using
NRC approved methodology and meets the applicable acceptance
criteria. Thus, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously evaluated.
Prior to the startup of U1C15, licensing analyses are performed
(using NRC approved methodology referenced in Technical
Specification Section 5.6.5.b) to determine changes in the critical
power ratio as a result of anticipated operational occurrences.
These results are added to the MCPR Safety Limit values to generate
the MCPR operating limits in the U1C15 COLR [core operating limits
report]. These limits could be different from those specified for
the current Unit 1 COLR. The COLR operating limits thus assure that
the MCPR Safety Limit will not be exceeded during normal operation
or anticipated operational occurrences. Postulated accidents are
also analyzed prior to the startup of U1C15 and the results shown to
be within the NRC approved criteria.
The changes to the references in Section 5.6.5.b were made to
properly reflect the NRC approved methodology used to generate the
U1C15 core operating limits. The use of this approved methodology
does not increase the probability of occurrence or consequences of
an accident previously evaluated.
Therefore, the proposed amendment does not involve a significant
increase in the probability of occurrence 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 change to the single-loop MCPR Safety Limit does not
directly or indirectly affect any plant system, equipment, or
component and therefore does not affect the failure modes of any of
these items. Thus, the proposed change does not create the
possibility of a previously unevaluated operator error or a new
single failure. The changes to the references in Section 5.6.5.b
were made to properly reflect the NRC approved methodology used to
generate the U1C15 core operating limits. The use of this approved
methodology does not create the possibility of a new or different
kind of accident.
Therefore, this proposed amendment 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.
Since the proposed changes do not alter any plant system,
equipment, component, or the processes used to operate the plant,
the proposed change will not jeopardize or degrade the function or
operation of any plant system or component governed by Technical
Specifications. The proposed single-loop MCPR Safety Limit does not
involve a significant reduction in the margin of safety as currently
defined in the Bases of the applicable Technical Specification
sections, because the MCPR Safety Limits calculated for U1C15
preserve the required margin of safety.
The changes to the references in section 5.6.5.b were made to
properly reflect the NRC approved methodology used to generate the
U1C15 core operating limits. This approved methodology is used to
demonstrate that all applicable criteria are met, thus,
demonstrating that there is no reduction in the margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer.
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: October 5, 2005.
Description of amendment request: The proposed amendment would
revise the SSES 1 and 2 Technical Specifications (TSs) 3.4.10, ``RCS
[reactor coolant system] Pressure and Temperature (P/T) Limits,'' to
remove valid P/T curve limit date and replacing
[[Page 2596]]
it with the effective full-power years (EFPY) of radiation exposure on
each of the P/T limit curves for SSES 1 and 2. The new P/T limit would
be 35.7 EFPY for SSES 1 and 30.2 EFPY for SSES 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed changes request that the P/T limits curves in
TS 3.4.10, ``RCS Pressure and Temperature (P/T) Limits'' be revised
by removing the valid date and replacing it with the Effective Full
Power Years of radiation exposure limit on each of the P/T curves
for SSES Units 1 and 2.
The P/T limits are prescribed during all operational conditions
to avoid encountering pressure, temperature, and temperature rate of
change conditions that might cause undetected flaws to propagate,
resulting in nonductile failure of the reactor coolant pressure
boundary, an unanalyzed condition. Therefore, the proposed changes
do not have any effect on the probability of an accident previously
evaluated.
The P/T curves are used as operational limits during heatup or
cooldown maneuvering, when pressure and temperature indications are
monitored and compared to the applicable curve to determine that
operation is within the allowable region. The P/T curves provide
assurance that station operation is consistent with previously
evaluated accidents. Thus, the radiological consequences of an
accident previously evaluated are not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed changes do not change the response of any plant
equipment to transient conditions. The proposed changes do not
introduce any new equipment, modes of system operation, or failure
mechanisms.
Therefore, there are no new types of failures or new or
different kinds of accidents or transients that could be created by
these changes. The proposed 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?
No. The consequences of a previously evaluated accident are not
increased by these proposed changes, since the Loss of Coolant
Accident analyzed in the FSAR [Final Safety Analysis Report] assumes
a complete break of the reactor coolant pressure boundary. The
changes to the P/T limits curves do not change this assumption.
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer.
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: November 9, 2004, as supplemented
December 15, 2005. This notice supersedes the original notice published
on April 26, 2005 (70 FR 21463), which was based upon the licensee's
application dated November 9, 2004.
Description of amendment request: The proposed amendments would
change the SSES 1 and 2 Technical Specifications (TSs) 3.8.4, ``DC
Sources-- Operating,'' 3.8.5, ``DC Sources--Shutdown,'' 3.8.6,
``Battery Cell Parameters,'' and add a new TS section, 5.5.13,
``Battery Monitoring and Maintenance Program.'' These changes are
consistent with Technical Specification Change Traveler (TSTF) 360,
Revision 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed changes restructure the Technical
Specifications (TSs) for the DC Electrical Power Systems. The
proposed changes consist of the relocation of several surveillance
requirements that perform preventive maintenance on the safety
related batteries, to a new license controlled program. The DC
electrical power systems, including associated battery chargers, are
not initiators to any accident sequence analyzed in the Final Safety
Analysis Report (FSAR). Operation in accordance with the proposed TS
ensures that the DC electrical power systems are capable of
performing functions as described in the FSAR. Therefore, the
mitigative functions supported by the DC Power Systems will continue
to provide the protection assumed by the analysis.
The relocation of preventive maintenance surveillance, and
certain operating limits and actions to a newly created, licensee-
controlled TS 5.5.13, ``Battery Monitoring and Maintenance
Program,'' will not challenge the ability of the DC electrical power
systems to perform their design functions. The maintenance and
monitoring required by current TS, which are based on industry
standards, will continue to be performed. In addition, the DC Power
Systems are within the scope of 10 CFR 50.65, ``Requirements for
Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants,'' which will ensure the control of maintenance activities
associated with the DC electrical power systems. The integrity of
fission product barriers, plant configuration, and operating
procedures as described in the FSAR will not be affected by the
proposed changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed changes involve restructuring the TS for the DC
electrical power systems. These changes will rely on a new license
controlled program to monitor battery parameters for operability.
The DC electrical power systems, which include the associated
battery chargers, are not initiators to any accident sequence
analyzed in the FSAR. Rather, the DC electrical power systems are
used to supply equipment used to mitigate an accident. These
mitigative functions, supported by the DC electrical power systems
are not affected by these changes and they will continue to provide
the protection assumed by the safety analysis described in the FSAR.
There are no new types of failures or new or different kinds of
accidents or transients that could be created by these changes.
Therefore, the proposed 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?
No. The margin of safety is established through equipment
design, operating parameters, and the setpoints at which automatic
actions are initiated. The proposed changes will not adversely
affect operation of plant equipment. These changes will not result
in a change to the setpoints at which protective actions are
initiated. Sufficient DC electrical system capacity is ensured to
support operation of mitigation equipment. The changes associated
with the new Battery Maintenance and Monitoring Program will ensure
that the station batteries are maintained in a highly reliable
state. The equipment fed by the DC electrical sources will continue
to provide adequate power to safety related loads in accordance with
analysis assumptions.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
[[Page 2597]]
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: October 26, 2005.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.6.6, ``Containment Spray and Cooling
Systems,'' to change Required Action D.1 that currently allows 72 hours
of operation with both containment cooling trains out of service as
long as both containment spray trains are operable. The required action
would be revised to impose the more stringent requirement of requiring
plant shutdown if both containment cooling trains are out of service
instead of allowing the 72 hours to restore an inoperable train. There
are also changes to other required actions in TS 3.6.6 to reflect the
revision to Required Action D.1. In addition, the required action for
two inoperable containment spray trains is being revised.
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 in the required action when two containment
cooling trains are inoperable to require plant shutdown is more
restrictive than the current required action that allows 72 hours of
operation [to restore one containment cooling train to operable
status]. Also the proposed change to the required action [F.1 for]
when two containment cooling trains are inoperable to be in MODE 3
within 6 hours and MODE 5 within 36 hours [are the same as in the
current Required Actions E.1 and E.2 for when the two containment
cooling trains are inoperable. The proposed change to the required
action for two containment spray trains being inoperable] is more
restrictive than the current required action to enter LCO [Limiting
Condition for Operation] 3.0.3 immediately [because] LCO 3.0.3
requires the plant to be in MODE 3 within 7 hours. The more
stringent requirements are imposed to ensure process variables,
structures, systems and components are maintained consistently with
the safety analysis and licensing basis [for Callaway].
All of these proposed changes have been reviewed to ensure no
previously evaluated accident has been adversely affected. [The
proposed changes are not accident initiators.] Therefore, this
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 does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in controlling [plant] parameters. The proposed change
does impose different requirements. However, these changes are
consistent with [the] assumptions made in the safety analysis and
licensing basis [for Callaway]. Thus, this change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No
The imposition of more stringent requirements has no impact on
or will increase the margin of safety. The change in the required
action when two containment cooling trains are out of service will
increase the margin of safety by decreasing the allowed restoration
time [to restore an inoperable containment cooling train to operable
status].
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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station (OCNGS), Ocean County, New Jersey
Date of application for amendment: December 17, 2004.
Brief description of amendment: The amendment revised Appendix B,
Environmental Technical Specifications, of the OCNGS Facility Operating
License, principally by deleting redundant reporting requirements,
aligning various requirements with regulations and accepted guidance
documents, and correcting administrative errors.
Date of Issuance: January 4, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
[[Page 2598]]
Amendment No.: 257.
Facility Operating License No. DPR-16: The amendment revised the
Environmental Technical Specifications.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19113).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated January 4, 2006.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: February 25, 2005, as
supplemented by letter dated August 4, 2005.
Brief description of amendment: The amendment revised the Millstone
Power Station, Unit No. 2, Technical Specifications Surveillance
Requirement for trisodium phosphate to remove the granularity term and
chemical detail. In addition, the proposed change will increase the
allowed outage time from 48 to 72 hours.
Date of issuance: January 3, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 290.
Facility Operating License No. DPR-65: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 2005 (70 FR
41444). The additional information provided in the supplemental letter
dated August 4, 2005, did not expand the scope of the application as
noticed and did not change the NRC staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 3, 2006.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: December 16, 2004, as
supplemented on October 5, 2005.
Brief description of amendment: The amendment revised the current
fuel rod average licensing basis burnup limit for one lead test
assembly containing advanced zirconium based alloys to a limit not
exceeding 71,000 megawatt-days per metric ton of uranium.
Date of issuance: December 30, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 228
Facility Operating License No. NPF-49: The amendment revised the
design basis.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5238). The October 5, 2005, supplement provided clarifying information
and did not 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 30, 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: July 20, 2005, as supplemented by letter
dated September 14, 2005.
Brief description of amendment: The amendment approves the transfer
of Facility Operating License and Materials License No. NPF-38, held by
Entergy Louisiana, Inc. (ELI) and Entergy Operatings, Inc. (EOI), for
the Waterford Steam Electric Station, Unit 3 (Waterford 3). The
transfer is associated with the restructuring of ELI from a Louisiana
corporation to a Texas limited liability company, Entergy Louisiana,
LLC (ELL). EOI will continue to operate Waterford 3, and the
restructuring will not affect the technical or financial qualifications
of ELL or EOI.
Date of issuance: December 31, 2005.
Effective date: At the time the transfer is completed.
Amendment No.: 203.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Materials License.
Date of initial notice in Federal Register: October 17, 2005 (70 FR
60374).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 2, 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: December 17, 2004.
Brief description of amendments: The amendments revised the
Appendix B, Environmental Technical Specifications.
Date of issuance: January 3, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 257 and 260.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the Environmental Technical Specifications.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19112).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 3, 2006.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 28, 2005, as supplemented
September 23, 2005.
Description of amendment request: The amendment extended the
expiration of Facility Operating License (FOL) NPF-86 for Seabrook
Station, Unit No. 1, by approximately 3.4 years. The extension sets the
date of expiration of the FOL to occur 40 years from the date of
issuance of the full-power operating license. Specifically, the FOL,
with a previous expiration date of October 17, 2026, now expires March
15, 2030. This change allows the recapture of zero-power and low-power
testing time in accordance with SECY-98-296, ``Agency Policy Regarding
Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear
Power Plants,'' dated December 21, 1998.
Date of issuance: December 28, 2005.
Effective date: As of its date of issuance, and shall be
implemented within 30 days.
Amendment No.: 105.
Facility Operating License No. NPF-86: The amendment revised the
License.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29797). The licensee's September 23, 2005 supplement provided
clarifying information that did not change the scope of the proposed
amendment as described in the original notice of proposed action
published in the Federal Register, and did not 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 28, 2005.
No significant hazards consideration comments received: No.
[[Page 2599]]
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: March 31, 2005, as supplemented
November 9, 2005.
Brief description of amendment: This amendment extended the date
for the next Appendix J, Type A test at St. Lucie Unit 2 until the end
of the SL2-17 refueling outage.
Date of Issuance: December 23, 2005.
Effective Date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 140.
Renewed Facility Operating License No. NPF-16: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33215). The November 9, 2005, supplement did not affect the original
proposed no significant hazards determination, or expand the scope of
the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 2005.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: October 29, 2004, as
supplemented by letters dated May 6 and October 31, 2005.
Brief description of amendments: The amendments revised the
Technical Specification (TS) requirements for the handling of
irradiated fuel in the containment and fuel building, and certain
specifications related to performing core alterations. These changes
are based on analysis of the postulated fuel handling and core
alteration accidents and transients for Diablo Canyon Nuclear Power
Plant, Units 1 and 2. The amendments are consistent with the NRC-
approved Industry/Technical Specification Task Force (TSTF) Standard
Technical Specifications Change Traveler, TSTF-51, Revision 2, ``Revise
containment requirements during handling irradiated fuel and core
alterations.'' In addition, the amendments made editorial corrections
to TS 3.1.7, ``Rod Position Indication,'' TS 3.3.1, ``Reactor Trip
System (RTS) Instrumentation,'' TS 3.4.16, ``RCS Specific Activity,''
TS 3.7.3, ``Main Feedwater Isolation Valve (MFIVs), Main Feedwater
Regulating Valves (MFRVs), MFRV Bypass Valves, and Main Feedwater Pump
(MFWP) Turbine Stop Valves,'' and TS 3.7.13, ``Fuel Handling Building
Ventilation System (FHBVS).''
Date of issuance: January 3, 2006.
Effective date: January 3, 2006, and shall be implemented within 90
days of issuance.
Amendment Nos.: Unit 1--184; Unit 2--86.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 4, 2005 (70 FR
403)
The supplements dated May 6 and October 31, 2005, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 3, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: November 12, 2004, as
supplemented by letters dated September 2 and September 16, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) 3.1.7, ``Standby Liquid Control (SLC)
System,'' for Hatch, Units 1 and 2. The amendments update Figure 3.1.7-
1 and 3.1.7-2 of the Units 1 and 2 TS to reflect the increased
concentration of Boron-10 in the solution. Conforming revisions to
Bases B3.1.7, are also included.
Date of issuance: January 5, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 247/191.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5249).
The supplemental letter dated September 2, 2005, contained
clarifying information only and did not change the initial proposed no
significant hazards consideration determination or expand the scope of
the original Federal Register notice. The supplemental letter dated
September 16, 2005, contained information that expanded the scope of
the original Federal Register notice. The proposed amendment was re-
noticed on October 25, 2005 (70 FR 61662).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 5, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: April 27, 2005, as supplemented
by letter dated November 17, 2005.
Brief description of amendments: The amendments relocate several
Technical Specification (TS) requirements to the Sequoyah Technical
Requirements Manual (TRM). Specifically, the amendments relocate the
provisions for TS 3.3.2 (Movable Incore Detectors), TS 3.3.3.4
(Meteorological Instrumentation), TS 3.4.7 (Reactor Coolant System
Chemistry), TS 3.4.11 (Reactor Coolant System Head Vents), TS 3.7.2
(Steam Generator Pressure and Temperature Limitations), TS 3.7.10
(Sealed Source Contamination), TS 3.9.5 (Refueling Operations
Communications), and TS 3.9.6 (Manipulator Crane) to the TRM. These
changes are consistent with the latest version of NUREG-1431, Revision
3, ``Standard Technical Specifications for Westinghouse Plants,'' and
do not diminish the level of safety found in the current TSs.
Date of issuance: December 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 305, 295.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: July 5, 2005 (70 FR
38723). The supplemental letter of November 17, 2005, provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 28, 2005.
No significant hazards consideration comments received: No.
[[Page 2600]]
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of application for amendments: December 17, 2004.
Brief Description of amendments: These amendments revised the
reactor coolant pressure and temperature limits, low-temperature
overpressure protection system (LTOPS) setpoint values, and LTOPS
enable temperatures that are valid for up to 47.6 effective full-power
years (EFPY) and 48.1 EFPY of operation at Surry Power Station, Unit
Nos. 1 and 2, respectively.
Date of issuance: January 3, 2006.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment Nos.: 245/244.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9999).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 3, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 9th day of January 2006.
For the Nuclear Regulatory Commission.
Edwin M. Hackett,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 06-320 Filed 1-13-06; 8:45 am]
BILLING CODE 7590-01-P