[Federal Register Volume 70, Number 89 (Tuesday, May 10, 2005)]
[Notices]
[Pages 24645-24662]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E5-2207]
=======================================================================
-----------------------------------------------------------------------
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 April 15, 2005 to April 28, 2005. The last
biweekly notice was published on April 26, 2005 (70 FR 21449).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, 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
[[Page 24646]]
Atomic Safety and Licensing Board Panel, will rule on the request and/
or petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact 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: March 28, 2005.
Description of amendment request: The licensee proposed to revise
the licensing bases of OCNGS in the area of radiological dose analyses
for the design-basis accidents (DBAs). Specifically, the licensee
proposed to use the alternative source terms (AST) depicted in
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for
Evaluating Design Basis Accidents at Nuclear Power Reactors,'' instead
of the source terms used in the current licensing basis and depicted in
Technical Information Document 14844, ``Calculation of Distance Factors
for Power and Test Reactor Sites.'' The acceptance criteria for the
postulated consequences using AST are set forth in 10 CFR 50.67 and
General Design Criterion 19, ``Control Room.'' The licensee has
performed radiological consequence analysis for the most limiting DBAs
that result in offsite and control room operator exposure to support a
full-scope implementation of the AST. If approved, the amendment would:
(1) Revise Section 3.2.A, ``Standby Liquid Control System,'' of the
Technical Specifications (TSs) to add a specification to require that
the subject system is operable when the reactor is at or greater than
212 degrees Fahrenheit; (2) revise various pages of the TS Bases to
reflect use of the AST methodology. The issuance of the requested
amendment would also signify the NRC staff's approval to revise the
OCNGS Updated Final Safety Analysis Report to reflect implementation of
the AST in the OCNGS licensing basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 24647]]
consideration. The NRC staff's analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The 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 radiation release. It does,
however, better represent the physical characteristics of the release,
so that appropriate mitigation techniques may be applied. The proposed
amendment does not affect the design of plant systems, structures, or
components (SSCs), or their operational characteristics or function. As
a result, implementing the AST would not have any increase on the
frequency of occurrence for previously analyzed accidents. It may be
argued that the calculated radiological consequences are different
because a different set of assumptions, with accompanying acceptance
criteria, are used. However, since there is no design or operational
change associated with the proposed amendment, the actual consequences
of the same accident would not be changed regardless of what
methodology was used before the accident to arrive at postulated
consequences. As a result, implementing the AST would not increase the
consequences of any previously evaluated accident.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment does not alter the design,
configuration, or method of operation of any SSC. Therefore, no new
initiators or precursors of a new or different kind of accident are
created that could result in a new or different kind of accident.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Margins of safety are established in
the design of components, the configuration of components to meet
certain performance parameters, and in the establishment of setpoints
to initiate alarms or actions. These are principally documented in the
OCNGS licensing basis documents such as the Updated Final Safety
Analysis Report, and none of these would be changed by the amendment.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the proposed amendment 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 Section Chief: Richard J. Laufer.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: March 4, 2005.
Description of amendments request: The proposed amendments would
delete Section 2.F (2.G in Unit 3) of the Operating License which
requires reporting violations of the requirements in Section 2.C of the
Operating License. The amendments will also make administrative and
editorial changes to the Technical Specifications (TSs). Changes to TS
1.4, ``Frequency,'' and TS 3.4.3, ``RCS Pressure and Temperature (P/T)
Limits,'' will correct editorial errors. The changes to TS 2.1.1,
``Reactor Core SLs,'' and TS 3.3.1, ``Reactor Protective System (RPS)
Instrumentation--Operating,'' will remove the reference to departure
from nucleate boiling ratios (DNBR) based on operating cycle, since
only one of the listed DNBR values is now valid. TS 3.1.10, ``Special
Test Exceptions (STE)--MODES 1 and 2,'' will be changed to correct an
inconsistency between the limiting condition for operation and the TS
Bases. The changes to TS 3.7.2, ``Main Steam Isolation Valves (MSIVs)''
and TS 3.7.3, ``Main Feedwater Isolation Valves (MFIVs)'' will correct
the applicability for these specifications. The change to TS 3.8.1,
``AC Sources--Operating'' will add a note to a surveillance
requirement. Changes to TS 3.8.4, ``DC Sources--Operating'' and TS
3.8.6, ``Battery Cell Parameter'' will remove the reference to AT&T
batteries. The changes to TS 5.5.9, ``Steam Generator (SG) Tube
Surveillance Program'' will correct the reference for NRC notification.
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 amendment includes [the] following changes that are
considered to be administrative and/or editorial changes:
The reporting requirement in License Condition 2.F (2.G in Unit
3) is adequately addressed by the requirements identified in 10 CFR
50.72, ``Immediate notification requirements for operating nuclear
power reactors'' and 10 CFR 50.73, ``Licensee event report system.''
Since Condition 2.F (2.G in Unit 3) is adequately addressed by the
requirements in 10 CFR 50.72 and 10 CFR 50.73, the Condition is not
required. Therefore, this is considered an administrative change
that eliminates regulatory requirements that are adequately
addressed by the requirements in 10 CFR 50.72 and 10 CFR 50.73.
The changes to Technical Specifications (TS) 1.4 and 3.4.3 are
editorial changes only. These changes maintain the format of the
Technical Specifications and correct editorial errors in the
Technical Specifications.
The changes to Technical Specifications 2.1.1 and 3.3.1 remove
requirements that are no longer applicable to the Palo Verde Nuclear
Generating Station (PVNGS) units. As part of Amendment 133 to the
PVNGS Operating License, the minimum DNBR was revised based on Unit
operating cycle, >=1.30 (through operating cycle 10)'' and >=1.34
(operating cycle 11 and later).'' All three PVNGS units have
completed operating cycle 10. Therefore, the reference to the
minimum d[e]parture from nucleate boiling ratio (DNBR) through
operating cycle 10 (>=1.30) is no longer required.
The changes to Technical Specification 3.1.10 correct an
inconsistency between the Technical Specification limiting condition
for operation (LCO) and Bases. The Bases for this specification
states that ``Even if an accident occurs during PHYSICS TESTS with
one or more LCOs suspended, fuel damage criteria are preserved
because the limits on power distribution and shutdown capability are
maintained during PHYSICS TESTS.'' The limits on power distribution
are maintained by TSs 3.2.1, ``Linear Heat Rate (LHR)'' and 3.2.4
``Departure from Nucleate Boiling Ratio (DNBR).'' These changes
ensure that shutdown capability is maintained during physics tests.
The changes to Technical Specifications Section 3.7.2, ``Main
Steam Isolation Valves (MSIVs)'' and Section 3.7.3, ``Main Feedwater
Isolation Valves (MFIVs)'' correct an inconsistency between the
applicability and the required actions. The changes are consistent
with the guidance in NUREG-1432, ``Standard Technical
Specifications, Combustion Engineering Plants.'' Therefore, this is
considered an administrative change that corrects an inconsistency
in the Technical Specifications.
The changes to Technical Specifications Section 3.8.1, ``AC
Sources--Operating,'' correct an inconsistency in the surveillance
requirements that were revised in Amendment 129 to the PVNGS
Operating License. A note was not included with the change to one of
the surveillance requirements. This change adds the note to
[[Page 24648]]
the surveillance requirement. Therefore, this is considered an
administrative change that corrects an inconsistency in the
Technical Specifications.
The changes to Technical Specifications Section 3.8.4, ``DC
Sources--Operating'' and Section 3.8.6, ``Battery Cell Parameters''
removes the requirements and references to the AT&T batteries. APS
has replaced the AT&T batteries with low specific gravity batteries
in all three units. Therefore, this is considered an administrative
change that removes unnecessary requirements and references.
The changes to Technical Specifications Section 5.5.9, ``Steam
Generator (SG) Tube Surveillance Program,'' updates the requirement
to notify the NRC based on the January 23, 2001 rule change to 10
CFR 50.72. Therefore, this change corrects NRC notification
requirements in Technical Specifications, based on the January 23,
2001 rule change to 10 CFR 50.72 (65 FR 63786, 10/25/00).
As discussed above the proposed amendment involves
administrative and/or editorial changes only. The proposed amendment
does not impact any accident initiators, analyzed events, or assumed
mitigation of accident or transient events. The proposed changes do
not involve the addition or removal of any equipment or any design
changes to the facility. The proposed changes do not affect plant
operations, any design function or an analysis that verifies the
capability of structures, systems, and components (SSCs) of the
plant. The proposed changes do not change any of the previously
evaluated accidents in the updated final safety analysis report
(UFSAR). The proposed changes do not affect SSCs, operating
procedures, and administrative controls that have the function of
preventing or mitigating any of these accidents.
Therefore, the proposed changes do not represent 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.
As discussed in standard 1, the proposed amendment only involves
administrative and/or editorial changes. No actual plant equipment
or accident analysis will be affected by the proposed changes. The
proposed changes will not change the design function or operation of
any SSCs. The proposed changes will not result in any new failure
mechanisms, malfunctions, or accident initiators not considered in
the design and licensing bases. The proposed amendment does not
impact any accident initiators, analyzed events, or assumed
mitigation of accident or transient events.
Therefore, this proposed change does not create the possibility
of an accident of a different kind than previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety.
Response: No.
As discussed in standard 1, the proposed amendment only involves
administrative and/or editorial changes. Margin of safety is
associated with confidence in the ability of the fission product
barriers (i.e., fuel and fuel cladding, reactor coolant system
pressure boundary, and containment structure) to limit the level of
radiation dose to the public. This request involves administrative
and/or editorial changes only. No actual plant equipment or accident
analysis will be affected by the proposed changes. Additionally, the
proposed changes will not relax any criteria used to establish
safety limits, will not relax any safety system settings, or will
not relax the bases for any limiting conditions for operation.
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: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Section Chief: Robert A. Gramm.
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: January 27, 2005.
Description of amendments request: The proposed amendment would
allow entry into a mode or other specified condition in the
applicability of a Technical Specification (TS), while in a condition
statement and the associated required actions of the TSs, provided the
licensee performs a risk assessment and manages risk consistent with
the program in place for complying with the requirements of Title 10 of
the Code of Federal Regulations (10 CFR), Part 50, Section 50.65(a)(4).
Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual
TSs would be eliminated, several notes or specific exceptions would be
revised to reflect the related changes to LCO 3.0.4, and Surveillance
Requirement (SR) 3.0.4 would be revised to reflect the LCO 3.0.4
allowance.
This change was proposed by the industry's TS Task Force (TSTF) and
is designated TSTF-359. The NRC staff issued a notice of opportunity
for comment in the Federal Register on August 2, 2002 (67 FR 50475), on
possible amendments concerning TSTF-359, including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on April 4, 2003 (68 FR 16579). The licensee affirmed the
applicability of the following NSHC determination in its application
dated January 27, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the
[[Page 24649]]
plant without the full complement of equipment through the
conditions for not meeting the TS LCO. The risk associated with this
allowance is managed by the imposition of required actions that must
be performed within the prescribed completion times. The net effect
of being in a TS condition on the margin of safety is not considered
significant. The proposed change does not alter the required actions
or completion times of the TS. The proposed change allows TS
conditions to be entered, and the associated required actions and
completion times to be used in new circumstances. This use is
predicated upon the licensee's performance of a risk assessment and
the management of plant risk. The change also eliminates current
allowances for utilizing required actions and completion times in
similar circumstances, without assessing and managing risk. The net
change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Esquire, Counsel,
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor,
Baltimore, MD 21202.
NRC Section Chief: Richard J. Laufer.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: March 14, 2005.
Description of amendment request: The proposed amendments would
delete Technical Specification (TS) Section 5.5.4, ``Post Accident
Sampling,'' requirements to maintain a Post Accident Sampling System
(PASS). Licensees were generally required to implement PASS upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97, Revision 3,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Access
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the NRC's lessons
learned from the accident that occurred at TMI Unit 2. Requirements
related to PASS were imposed by Order for many facilities and were
added to or included in the TS for nuclear power reactors currently
licensed to operate. Lessons learned and improvements implemented over
the last 20 years have shown that the information obtained from PASS
can be readily obtained through other means or is of little use in the
assessment and mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 3, 2003 (68 FR 10052) on possible amendments
to eliminate PASS, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in a
license amendment application in the Federal Register on May 13, 2003
(68 FR 25664). The licensee affirmed the applicability of the following
NSHC determination in its application dated March 14, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radioisotopes
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
[[Page 24650]]
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: September 23, 2004, as supplemented by
letter dated April 19, 2005.
Description of amendment request: The amendment would revise the
reactor operational limits, as specified in the River Bend Station Core
Operating Limits Report (COLR), to compensate for the inoperability of
the End of Cycle Recirculation Pump Trip (EOC-RPT) instrumentation.
This will provide an alternative to the existing Limiting Condition for
Operation for the EOC-RPT instrumentation. The revised Technical
Specification will require that either the EOC-RPT instrumentation be
operable or that Minimum Critical Power Ratio and Linear Heat
Generation Rate limits for the inoperable EOC-RPT be placed in effect
as specified in the COLR.
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 End of Cycle Recirculation Pump Trip (EOC-RPT) functions to
insert negative reactivity in response to certain anticipated
transients. The EOC-RPT is a mitigation function and not the
initiator of any evaluated accident or transient. Operation with
inoperable EOC-RPT instrumentation and compliance with new
restrictive Minimum Critical Power Ratio (MCPR) and Linear Heat
Generation Rate (LHGR) operating limits establish sufficient margin
to the core thermal MCPR safety limit (SL) and the thermal
mechanical design limits as would be the case with operable EOC-RPT
instrumentation and existing MCPR and LHGR limits.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not create any new modes of plant or
equipment operation. The proposed change allows the option to apply
an additional penalty factor to the MCPR and LHGR when the EOC-RPT
is inoperable. With the addition of the penalty factor, the margin
to the MCPR SL and the thermal mechanical design limits are
maintained. Therefore, operating the plant with the proposed change
will not create the possibility of a new or different kind of
accident from any previously analyzed.
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.
By establishing a new restrictive MCPR and LHGR operating limit,
there are no changes to the plant design and safety analysis. There
are no changes to the reactor core design instrument setpoints. The
margin of safety assumed in the safety analysis is not affected.
Applicable regulatory requirements will continue to be met and
adequate defense-in[-]depth will be maintained. Sufficient safety
margins will be maintained.
The analytical methods used to determine the revised core
operating limits were reviewed and approved by the NRC, and are
described in Technical Specification 5.6.5. Specific analyses were
prepared by the RBS fuel vendor to develop core operating limits
without crediting the EOC-RPT. Therefore, implementation of the
proposed changes will not involve a significant reduction in the
margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Allen G. Howe.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: February 15, 2005.
Description of amendment request: The proposed amendment would
approve application of an alternative source term methodology with the
exception that Technical Information Document 14844, ``Calculation of
Distance Factors for Power Test Reactor Sites,'' will continue to be
used as the radiation dose basis for equipment qualification.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The implementation of AST assumptions has been evaluated in
revisions to the analyses of the following limiting DBAs at the
Byron Station and Braidwood Station.
Loss-of-Coolant Accident
Fuel Handling Accident
Control Rod Ejection Accident
Locked Rotor Accident
Main Steam Line Break Accident
Steam Generator Tube Rupture Accident
Based upon the results of these analyses, it has been
demonstrated that, with the requested changes, the dose consequences
of these limiting events are within the regulatory guidance provided
by the NRC for use with the AST methodology. This guidance is
presented in RG 1.183, and Standard Review Plan Section 15.0.1. 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 AST methodology follows the guidance provided in RG 1.183
and satisfies the dose limits in 10 CFR 50.67. Even though these
limits are not directly comparable to the previously specified whole
body and thyroid requirements of 10 CFR 50, Appendix A, General
Design Criteria (GDC) 19, ``Control room,'' and 10 CFR 100.11,
``Determination of exclusion area, low population zone, and
population center distance,'' the results of the AST analyses have
demonstrated that the 10 CFR 50.67 limits are satisfied. Therefore,
it is concluded that AST does not involve a significant increase in
the consequences of an accident previously evaluated.
Implementation of AST provides increased operating margins for
the control room ventilation system filtration efficiencies. It also
relaxes containment integrity requirements while handling irradiated
fuel that has decayed for greater than 48 hours and during core
alterations. Automatic initiation of the radiation isolation mode
for the control room is not credited in the accident analysis which
allows relaxation of certain Technical Specification surveillance
requirements.
The equipment affected by the proposed changes is mitigative in
nature and relied upon after an accident has been initiated.
Application of the AST does result in changes to the functions and
operation of various filtration systems as described in the Updated
Final Safety Analysis Report (UFSAR). These effects have been
considered in the evaluations for these proposed changes. While the
operation of various systems does change with the implementation of
AST, the affected systems are not accident initiators; and
application of the AST methodology, itself, is not an initiator of a
design basis accident. The proposed changes to the TS revise certain
equipment performance requirements but do not require any physical
changes to the plant.
As a result, the proposed changes do not affect any of the
parameters or conditions
[[Page 24651]]
that could contribute to the initiation of any accidents. Relaxation
of operability requirements during the specified conditions will not
significantly increase the probability of occurrence of an accident
previously analyzed. Since design basis accident initiators are not
being altered by adoption of the AST, the probability of an accident
previously evaluated is not affected.
Based on the above discussion, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not involve a physical change to the
plant. Implementation of AST provides increased operating margins
for filtration system efficiencies. Application of AST also allows
for the relaxation of containment integrity requirements while
handling irradiated fuel that has decayed for greater than 48 hours
and during core alterations. Automatic initiation of the radiation
isolation mode for the control room is no longer credited in the
accident analysis.
Similarly, the proposed changes do not require any physical
changes to any structures, systems or components involved in the
mitigation of any accidents. Therefore, no new initiators or
precursors of a new or different kind of accident are created. New
equipment or personnel failure modes that might initiate a new type
of accident are not created as a result of the proposed changes.
Based on the above discussion, the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Approval of a change from the original source term methodology
(i.e., TID 14844) to an AST methodology, consistent with the
guidance in RG 1.183, will not result in a significant reduction in
the margin of safety. The safety margins and analytical
conservatisms associated with the AST methodology have been
evaluated and were found acceptable. The results of the revised DBA
analyses, performed in support of the proposed changes, are subject
to specific acceptance criteria as specified in RG 1.183. The dose
consequences of these DBAs remain within the acceptance criteria
presented in 10 CFR 50.67 and RG 1.183.
The proposed changes continue to ensure that the doses at the
exclusion area boundary (EAB) and low population zone boundary
(LPZ), as well as the control room, are within the specified
regulatory limits.
Therefore, based on the above discussion, 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio; Docket
Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2
(BVPS-1 and 2), Beaver County, Pennsylvania; Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: February 22, 2005.
Description of amendment request: The requested change will delete
Technical Specification requirements related to Occupational Radiation
Exposure Reports and Monthly Operating Reports.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated February 22, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve a significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chiefs: Gene Y. Suh, Richard J. Laufer.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: March 22, 2005.
Description of amendment request: The proposed amendments revise
the Technical Specifications (TS) for several Reactor Protection System
functional units. The steam/feedwater flow mismatch coincident with
steam generator water level--low reactor trip is being deleted, the
reactor trip on turbine trip interlock is being changed from P-7 to P-
8, the value of the P-8 interlock setpoint is being changed from 45
percent rated thermal power (RTP) to 40 percent RTP, and the value of
the P-8 interlock allowable value is being changed from 48 percent RTP
to 43 percent RTP.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes revise the operability requirements,
surveillance requirements and the interlock setpoint for two Reactor
Trip System functional units. The affected trip functional units are
not initiators of any accident previously evaluated. The proposed
changes to the affected trip functional units do not adversely
affect the initiators of any accident previously evaluated. A best
estimate
[[Page 24652]]
analysis has shown that a turbine trip without a reactor trip below
40% power does not challenge the pressurizer PORVs [power operated
relief valves] or the steam generator safety valves; thereby, not
adversely affecting the probability of a small break LOCA [loss of
coolant accident] due to a stuck open PORV, or an excessive cooldown
event due to a stuck open steam generator safety valve. As a result,
the probability of any accident previously evaluated is not
significantly increased by the proposed changes.
The steam/feedwater flow mismatch coincident with steam
generator water level--low reactor trip is not credited as a primary
trip in any previously evaluated accidents. The reactor trip on
turbine trip below the P-8 interlock is not credited as a primary
trip in any previously evaluated accidents. Therefore, the
mitigation functions that have been assumed in the accident analyses
will continue to be performed by the systems and components
currently credited in the analyses; and the accident analysis
results are not affected by the changes to the affected trip
functional units. The P-8 setpoint is not an initial condition of
any accident previously evaluated. Therefore, the accident analysis
results are not affected by changes to the P-8 setpoint. No safety
analyses previously performed in the Turkey Point Units 3 and 4
UFSAR [Updated Final Safety Analysis Report] required reanalysis for
these proposed changes. All accident analyses acceptance criteria
continue to be met. The proposed changes do not create any new
credible limiting single failure. As a result, the consequences of
any accident previously evaluated are not significantly increased by
the proposed changes.
In conclusion, operation of the facility in accordance with the
proposed amendments does not involve a significant increase in the
probability or consequences of any accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any previously evaluated.
No changes are being made to the plant that would introduce any
new accident causal mechanisms. The proposed changes do not
adversely affect previously identified accident initiators and do
not create any new accident initiators. No new limiting single
failures or accident scenarios are created by the proposed changes.
No new challenges to any installed safety system are created by
these proposed changes. The proposed changes do not result in any
event previously deemed incredible being made credible.
The steam/feedwater flow mismatch coincident with steam
generator water level--low reactor trip is not credited as an
inhibitor of any potential or actual accident initiators. So,
deletion of this reactor trip functional unit will not create the
possibility of a new or different kind of accident from any
previously evaluated.
Changing the interlock for the reactor trip on turbine trip from
P-7 to P-8 changes the power level associated with enabling and
disabling the reactor trip on turbine trip function. The turbine
pressure input to the reactor protection system permissives is not
an accident initiator and is not credited in the accident analyses.
Changing the P-8 allowable and trip setpoint values changes the
power level associated with enabling and disabling the reactor trip
functions currently associated with P-8. The change does not affect
how the associated trip functional units operate or function. Since
these interlock changes do not affect the way that the associated
trip functional units operate or function, the changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
Therefore, operation of the facility in accordance with the
proposed amendments 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
amendments would not involve a significant reduction in a margin of
safety.
No UFSAR safety analyses were changed or modified as a result of
these proposed changes. Therefore, all margins associated with the
current UFSAR safety analyses acceptance criteria are unaffected.
The current UFSAR safety analyses remain bounding. No UFSAR Chapter
14 events explicitly credit the steam/feedwater flow mismatch
reactor trip function and the reactor trip on turbine trip function
below the P-8 setpoint value. The safety systems credited in the
safety analyses will continue to be available to perform their
mitigation functions. Changing the P-8 setpoint from 45% to 40% is
in the conservative direction for the Reactor Coolant Flow--Low
Reactor Trip and the Reactor Coolant Pump Breaker Position Reactor
Trip. Therefore, the proposed changes do not result in a significant
reduction in a margin of safety; and operation of the facility in
accordance with the proposed amendments would 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Michael L. Marshall, Jr.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 28, 2005.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to allow the option of not
measuring the moderator temperature coefficient within 7 effective
full-power days after reaching an equilibrium boron concentration of
300 parts per million. This option would be available if the benchmark
criteria in WCAP-13749-P-A and the revised prediction specified in the
core operating limits report are satisfied.
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[s] do[es] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The probability or consequences of accidents previously
evaluated in the UFSAR [updated final safety analysis report] are
unaffected by this proposed change. There is no change to any
equipment response or accident mitigation scenario, and this change
results in no additional challenges to fission product barrier
integrity. The proposed change does not alter the design,
configuration, operation, or function of any plant system,
structure, or component. Further, the existing limits on moderator
temperature coefficient (MTC) established by the Technical
Specifications (TS), based on assumptions in the safety analyses,
remain unchanged and continue to be satisfied. 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 change[s] do[es] not create the possibility of a
new or different kind of accident from any 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
or removes any plant equipment, nor alters the design, physical
configuration, or mode of operation of any plant structure, system,
or component. The MTC is a variable that must remain within
prescribed limits, but it is not an accident initiator. No physical
changes are being made to the plant, so no new accident causal
mechanisms are being introduced. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. The proposed change[s] do[es] not involve a significant
reduction in the margin of safety.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
on the availability, operability, or performance of the safety-
related systems and components. 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
[[Page 24653]]
this change. A change to a surveillance requirement is proposed
based on an alternative method of confirming that the surveillance
is met. The Technical Specifications establish limits for the
moderator temperature coefficient (MTC) based on assumptions in the
accident analyses. Applying the conditional exemption from the MTC
measurement changes the method of meeting the surveillance
requirement; however, this change does not modify the TS values and
ensures adherence to the current TS limits. The basis for the
derivation of the MTC limits from the moderator density coefficient
(MDC) assumed in the accident analysis is unchanged. Further, the
safety analysis assumption of a constant MDC and its assumed value
will not change. Therefore, the margin of safety as defined in the
TS is not reduced and 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 Section Chief: Darrell J. Roberts.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 28, 2005.
Description of amendment request: The proposed amendment would
revise Seabrook Station, Unit No. 1 (Seabrook) Technical Specification
(TS) 3/4.9.13, ``Spent Fuel Assembly Storage.'' This revision would
reflect a revised criticality safety analysis supporting a two-zone
spent fuel pool consisting of BORAFLEX[reg] and BORAL[reg] fuel
assembly storage racks. Additionally, the proposed change would create
TS 3/4.9.15, ``Spent Fuel Pool Boron Concentration.''
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 license amendment incorporates the results of a
revised criticality analysis for the spent fuel pool without making
any physical changes to the facility. The revised criticality
analysis for the spent fuel pool (1) credits boron during movement
of fuel in the spent fuel pool, (2) assumes no neutron-absorbing
material in the BORAFLEX [reg] storage racks, and (3) applies a
conservative penalty in the analysis of BORAL [reg] racks. These
changes do not increase the probability of a fuel assembly being
misplaced within the spent fuel pool. The movement of fuel
assemblies will continue to be controlled by approved procedures,
and the placement of spent fuel will be controlled by the revised
Technical Specifications. The proposed changes do not alter or
prevent the ability of structures, systems, or components (SSCs) to
perform their intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in the Updated
Final Safety Analysis Report (UFSAR).
The proposed changes do not affect the source term, containment
isolation or radiological release assumptions used in evaluating the
radiological consequences of an accident previously evaluated in the
Seabrook Station UFSAR. The consequences of a misplaced fuel
assembly are not increased because the analysis demonstrates that
the fuel will remain sub-critical with a minimum of 872 ppm [part
per million] boron in the spent fuel pool. The new technical
specification included in this proposed change will ensure that the
minimum boron concentration is established during the movement of
fuel in the spent fuel pool. Further, the proposed changes neither
increase the types and amounts of radioactivity released offsite nor
increase occupational or public radiation exposures.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes to the TS do not alter the operation of the
spent fuel storage system or its ability to perform its design
function. The proposed changes do not include any physical changes
to the plant and do not introduce a new or different accident from
any type previously evaluated. A misplaced fuel assembly does not
represent a new or different type [of] accident, and the analysis
shows that the fuel remains sub-critical for the limiting case of a
misplaced fuel assembly. Similarly, continuing to take credit for
boron in the spent fuel under accident conditions does not create
the possibility of a new or different kind of accident. The previous
criticality analyses took credit for soluble boron in the spent fuel
pool water to show acceptable results in the analyses of fuel
misloading events.
Therefore the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The changes proposed by this license amendment ensure that the
spent fuel will remain sub-critical under normal and accident
conditions. The controlled placement of fuel assemblies within the
spent fuel pool will maintain Keff less than or equal to
0.95 as required by TS 5.6.1.1 for spent fuel storage. The proposed
amendment maintains the 0.95 limit on Keff by restricting
the placement of spent fuel and by crediting soluble boron in the
fuel pool water.
To assure that the true reactivity will be less than the
calculated reactivity, the analyses contain conservative assumptions
for calculating the safety limits for the spent fuel rack. With this
proposed change, Keff will be less than or equal to 0.95
with a 95% probability at a 95% confidence level.
Therefore, the proposed amendment does 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: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: Darrell J. Roberts.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of amendment request: April 1, 2005.
Description of amendment request: The licensee proposed to revise
Section 3.8.7, ``Inverters--Operating,'' of the Technical
Specifications (TSs), extending the time allowed to fix inoperable
emergency uninterruptible power supply (UPS) inverters from the current
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. The NRC staff's analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The proposed amendment does not affect the design of the
emergency UPS inverters, the operational characteristics or function of
the inverters, the interfaces between the inverters and other plant
systems, or the reliability of the inverters. An inoperable emergency
UPS inverter was not considered an initiator of a previously analyzed
event. In addition, the required actions and the associated completion
times specified by the TSs are not initiators of previously evaluated
accidents. As a result, extending the completion time
[[Page 24654]]
for an inoperable emergency UPS inverter would not have a significant
impact on the frequency of occurrence for a previously analyzed
accident. Furthermore, the proposed amendment will not result in
modifications to plant activities associated with inverter maintenance,
but rather, provides operational flexibility by allowing additional
time to perform inverter corrective maintenance and post-maintenance
testing on-line. The proposed extension of inoperable time will not
significantly affect the capability of inverters to perform their
safety function, which is to ensure an uninterruptible supply of 120-
volt alternating current (ac) electrical power to the associated power
distribution subsystems. The licensee performed a probabilistic risk
assessment which concluded that the increase in plant risk is small.
Therefore, the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment does not alter the design,
configuration, or method of operation of the emergency UPS inverters or
their associated 120-volt ac uninterruptible power distribution
subsystems, nor does the amendment alter any safety analyses inputs and
assumptions. The proposed extended emergency UPS inverter completion
time does not reduce the number of emergency UPS inverters below the
minimum required for safe shutdown or accident mitigation, and does not
affect the parameters within which NMP2 is operated or the setpoints at
which protective or mitigative actions are initiated. The use of the
alternate safety-related maintenance supply to power the 120-volt ac
uninterruptible power distribution subsystem is consistent with the
NMP2 design. If a station blackout event were to occur while an
emergency UPS inverter is out of service, a dedicated portable power
supply would be connected to provide a continuous source of power to
the connected systems. Accordingly, no new failure modes, system
interactions, or accident responses will be created that could result
in a new or different kind of accident.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Margins of safety are established in
the design of components, the configuration of components to meet
certain performance parameters, and in the establishment of setpoints
to initiate alarms or actions. The proposed amendment will not affect
any margin of safety as defined in the NMP2 Updated Safety Analysis
Report. The amendment does not change the design or operational
parameters of the UPS inverters as compared to original plant design.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the proposed amendment involves no
significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: April 1, 2005.
Description of amendment request: The proposed amendment would
provide one-time extension to the completion time for restoration of a
service water train to operable status in Technical Specification (TS)
3.7.8, ``Service Water System (SWS).''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve a significant increase
in the probability of an accident previously evaluated because the
extended Technical Specification action completion time is not an
accident initiator. Therefore the probability is not increased
significantly.
The proposed amendment does not involve a significant increase
in the consequences of an accident previously evaluated. With
service water pump P-7C inoperable, 100% of the required post-
accident SWS cooling capability remains available with the redundant
train maintained operable. A risk analysis was performed to show
that the consequences are not significantly increased. The
compensatory measures provide additional assurance that there is no
significant increase in the consequences of an accident associated
with extending the Technical Specification action completion time
for the service water system for an additional 96 hours.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed amendment only extends the Technical
Specification action completion time and does not involve a physical
alteration of any system, structure or component (SSC), or change in
the way any SSC is operated. The proposed amendment does not involve
operation of any required SSCs in a manner or configuration
different from those previously recognized or evaluated. No new
failure mechanisms will be introduced by the changes being
requested.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment does not involve a significant reduction
in a margin of safety. With service water pump P-7C inoperable, 100%
of the required post-accident service water system cooling
capability remains available with the redundant train maintained
operable. Therefore, there is no significant reduction in the margin
of safety.
Based on the availability of redundant systems, the compensatory
measures that will be taken, and the low probability of an accident
that could not be mitigated by the available systems, the proposed
amendment would 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 Part
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 Section Chief: L. Raghavan.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania
Date of amendment request: January 28, 2005.
Description of amendment request: The proposed amendment would
revise
[[Page 24655]]
the SSES 2 Technical Specification (TS) Table 3.3.5.1-1 ``Emergency
Core Cooling System Instrumentation,'' to change Function 3.e ``HPCI
[High-Pressure Coolant Injection] System,'' conditions referenced from
Required Action A.1 from ``D'' to ``C.'' This is an editorial revision
to correct a typographical error that has been present since PPL
converted to the Improved Technical Specifications in 1998.
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 Unit 2 TS Table 3.3.5.1 provides a
correction to a typographical error that occurred when preparing a
change to Unit 2 Technical Specification Table 3.3.5.1-1 in the
response to an NRC Request for Additional Information (RAI). The
request was initiated during NRC review of documents submitted by
PPL for the conversion to the Improved Technical Specifications.
This proposed change is considered to be administrative in nature
because it was originally submitted correctly and was inadvertently
changed in response to the RAI.
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.
As stated above, the proposed change to the Unit 2 TS Table
3.3.5.1 provides a correction to a typographical error that occurred
when preparing the response to an NRC Request for Additional
Information. The request was initiated by the NRC during its review
of documents submitted by PPL for the conversion to the Improved
Technical Specifications. This proposed change is administrative in
nature.
Therefore, these 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.
Again, the proposed change to the Unit 2 TS Table 3.3.5.1
provides a correction to a typographical error that occurred when
preparing the response to an NRC Request for Additional Information.
The request was initiated by the NRC during its review of documents
submitted by PPL for the conversion to the Improved Technical
Specifications. This proposed change is administrative in nature.
Therefore, these 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 Section Chief: Richard J. Laufer.
PSEG Nuclear LLC, Docket Nos. 50-272, Salem Nuclear Generating Station,
Unit No. 1 Salem County, New Jersey
Date of amendment request: February 23, 2005.
Description of amendment request: The proposed changes will revise
Technical Specification (TS) Steam Generator (SG) requirements for
Salem Nuclear Generating Station, Unit No. 1. The proposed changes
would replace TS 3/4.4.5 ``Steam Generator (SG)'' with ``Steam
Generator Tube Integrity;'' add a new TS 6.8.4.i, ``Steam Generator
Program;'' and add a new reporting requirement TS 6.9.1.10 ``Steam
Generator Tube Inspection Report.'' Additionally, the proposed changes
would revise TS 3/4.4.6.2, ``Reactor Coolant System Operational
Leakage.'' Specifically, the Limiting Condition for Operation and
ACTION and Surveillance Requirements of TS 3/4.4.6.2 would be revised
to clarify the requirements related to primary-to-secondary leakage.
These changes would facilitate implementation of industry initiative
Nuclear Energy Institute (NEI) 97-08, ``Steam Generator Program
Guidelines,'' to allow a comprehensive, performance-based approach to
managing SG performance at Salem Nuclear Generating Station, Unit No.
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?
Response: No.
The proposed change[s] require[s] a Steam Generator Program that
includes performance criteria that will provide reasonable assurance
that the steam generator (SG) tubing will retain integrity over the
full range of operating conditions (including startup, operation in
the power range, hot standby, cool down and all anticipated
transients included in the design specification). The SG performance
criteria are based on tube structural integrity, accident induced
leakage, and operational leakage.
The structural integrity performance criterion is:
All in-service steam generator tubes shall retain structural
integrity over the full range of normal operating conditions
(including startup, operation in the power range, hot standby, and
cool down and all anticipated transients included in the design
specification) and design basis accidents. This includes retaining a
safety factor of 3.0 against burst under normal steady state full
power operation primary-to-secondary pressure differential and a
safety factor of 1.4 against burst applied to the design basis
accident primary-to-secondary pressure differentials. Apart from the
above requirements, additional loading conditions associated with
the design basis accidents, or combination of accidents in
accordance with the design and licensing basis, shall also be
evaluated to determine if the associated loads contribute
significantly to burst or collapse. In the assessment of tube
integrity, those loads that do significantly affect burst or
collapse shall be determined and assessed in combination with the
loads due to pressure with a safety factor of 1.2 on the combined
primary loads and 1.0 on axial secondary loads.
The accident induced leakage performance criterion is:
The primary-to-secondary accident induced leakage rate for any
design basis accidents, other than a SG tube rupture, shall not
exceed the leakage rate assumed in the accident analysis in terms of
total leakage rate for all SGs and leakage rate for an individual
SG. Leakage is not to exceed 1 gpm per SG.
The operational leakage performance criterion is:
The reactor coolant system operational primary-to-secondary
leakage through any one SG shall be limited to 150 gallons per day.
A steam generator tube rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis of a[n] SGTR event, a bounding primary-to-
secondary leakage rate equal to the operational leakage rate limits
in the licensing basis plus the leakage rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as main steam line break
(MSLB), rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses assume that primary-to-
secondary leakage for all SGs is 1 gallon per minute or increases to
1 gallon per minute as a result of accident-induced stresses. The
accident induced leakage criterion retained by the proposed changes
accounts for tubes that may leak during design basis accidents. The
accident induced leakage criterion limits this leakage to no more
than the value assumed in the accident analysis.
[[Page 24656]]
The SG performance criteria proposed as part of these TS changes
identify the standards against which tube integrity is to be
measured. Meeting the performance criteria provides reasonable
assurance that the SG tubing will remain capable of fulfilling its
specific safety function of maintaining reactor coolant pressure
boundary integrity throughout each operating cycle and in the
unlikely event of a design basis accident. The performance criteria
are only a part of the Steam Generator Program required by the
proposed addition of TS 6.8.4.i. The program defined by NEI 97-06
includes a framework that incorporates a balance of prevention,
inspection, evaluation, repair, and leakage monitoring.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary-to-secondary leakage rates resulting from an accident.
Therefore, limits are included in the Salem TS for operational
leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure
the plant is operated within its analyzed condition. The Salem
analysis of the limiting design basis accident assumes that primary-
to-secondary leak rate after the accident is 1 gallon per minute
with no more than 500 gallons per day through any one SG, and that
the reactor coolant activity levels of DOSE EQUIVALENT I-131 are at
the TS values before the accident.
The proposed change[s] do[es] not affect the design of the SGs,
their method of operation, or primary coolant chemistry controls.
The proposed approach updates the current TS and enhances the
requirements for SG inspections.
The proposed change[s] do[es] not adversely impact any other
previously evaluated design basis accident and [are] an improvement
over the current TS.
Therefore, the proposed changes do not affect the consequences
of a[n] SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
probabilities or consequences of an MSLB, rod ejection, or a reactor
coolant pump locked rotor event.
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 performance based requirements are an improvement
over the requirements imposed by the current TS.
Implementation of the proposed Steam Generator Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the Steam Generator Program will be
an enhancement of SG tube performance. Primary-to-secondary leakage
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed changes do not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change[s] do[es] not impact any
other plant system or component. The change[s] enhance[s] SG
inspection requirements.
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 SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change[s] do[es] not affect tube design or operating environment.
The proposed change[s] [are] expected to result in an improvement in
the tube integrity by implementing the Steam Generator Program to
manage SG tube inspection, assessment, and plugging. The
requirements established by the Steam Generator Program are
consistent with those in the applicable design codes and standards
and are an improvement over the requirements in the current TS.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed changes to the
TS.
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 Section Chief: Darrell J. Roberts.
Southern California Edison Company (SCE), et al., Docket Nos. 50-361
and 50-362, San Onofre Nuclear Generating Station, Unit 2 and Unit 3,
San Diego County, California
Date of amendment requests: March 24, 2005.
Description of amendment requests: The proposed change would revise
the following Technical Specifications (TSs):
TS 1.1, Definitions, correct the definition of SHUTDOWN
MARGIN (SDM).
TS 3.1.1, SHUTDOWN MARGIN (SDM)--Tavg > 2000F, and TS
3.1.2, SHUTDOWN MARGIN (SDM)--Tavg < 2000F, relocate the numerical
shutdown margin requirements to the Core Operating Limits Report
(COLR).
TS 3.1.3, Reactivity Balance, increase the required action
time from 72 hours to 7 days when the ``Core reactivity balance not
within limit.''
TS 3.1.5, Control Element Assembly (CEA) Alignment, TS
3.1.6, Shutdown Control Element Assembly (CEA) Insertion Limits, and TS
3.1.7, Regulating CEA Insertion Limits, remove the requirement to
verify SDM.
TS 3.2.4, Departure From Nucleate Boiling Ratio (DNBR),
relocate to the COLR the power margin that must be accommodated when
the Core Operating Limit Supervisory System (COLSS) is in service and
neither CEA calculator is OPERABLE.
TS 5.7.1.5, CORE OPERATING LIMITS REPORT (COLR), identify
that the limits for TSs 3.1.1 and 3.1.2 shall be in the COLR.
The proposed changes are consistent with the Standard Technical
Specifications for Combustion Engineering Plants, NUREG-1432, Revision
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. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Limiting Conditions of Operation (LCOs) and Core Operating
Limits Report (COLR) will continue to restrict operation to within
the regions that provide acceptable results. The safety analysis
will continue to be performed in accordance with the Nuclear
Regulatory Commission (NRC) approved San Onofre Units 2 and 3 reload
analysis methodology.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not add any new equipment, modify any
interfaces with any existing equipment, alter the equipment's
function, or change the method of operating the equipment. The
proposed change does not alter plant conditions in a manner that
could affect other plant components. The proposed change does not
cause any existing equipment to become an accident initiator.
[[Page 24657]]
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Safety Limits ensure that Specified Acceptable Fuel Design
Limits are not exceeded during steady state operation, normal
operational transients, and anticipated operational occurrences. All
fuel limits and design criteria will continue to be met, based on
the NRC approved San Onofre Units 2 and 3 reload analysis
methodology. Therefore, the proposed change will have no impact on
the margins as defined in the Technical Specification bases.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Robert A. Gramm.
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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: December 16, 2004.
Brief description of amendments: The amendments delete TS 5.6.1,
``Occupational Radiation Exposure Report'' and TS 5.6.4, ``Monthly
Operating Reports,'' as described in the Notice of Availability
published in the Federal Register on June 23, 2004 (69 FR 35067).
Date of issuance: April 27, 2005.
Effective date: April 27, 2005, and shall be implemented within 90
days of the date of issuance.
Amendment Nos.: Unit 1-154, Unit 2-154, Unit 3-154.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5236).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 27, 2005.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: November 17, 2004.
Brief Description of amendments: The amendments eliminate the
requirements to submit monthly operating reports and annual
occupational radiation exposure reports.
Date of issuance: April 19, 2005.
Effective date: April 19, 2005.
Amendment Nos.: 235 and 263.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7763).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 19, 2005.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: November 17, 2004.
Brief description of amendment: The amendment eliminates the
requirements to submit monthly operating reports and annual
occupational radiation exposure reports.
Date of issuance: April 19, 2005.
Effective date: April 19, 2005.
Amendment No.: 204.
Renewed Facility Operating License No. DPR-23. Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7763)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 2005.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: November 17, 2004.
Brief description of amendment: This amendment revises Technical
Specifications by eliminating the requirements to submit monthly
operating reports and annual occupational radiation exposure reports.
Date of issuance: April 19, 2005.
Effective date: April 19, 2005.
Amendment No.: 118.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 2004 (70
FR 7763).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 2005.
[[Page 24658]]
No significant hazards consideration comments received: No.
Entergy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: September 23, 2004, as
supplemented by letter dated January 13, 2005.
Brief description of amendment: The change revises Columbia
Generating Station's licensing basis by replacing the current plant-
specific reactor pressure vessel material surveillance program with the
boiling water reactor vessels and internals project (BWRVIP) integrated
surveillance program (ISP). Specifically, the amendment revises
Columbia's final safety analysis report to include participation in the
ISP as described in the program document BWRVIP-86-A, ``BWR [Boiling
Water Reactor] Vessel and Internals Project Updated BWR Integrated
Surveillance Program (ISP) Implementation Plan,'' dated October 2002.
Date of issuance: April 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 192.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62471).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 28, 2005.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: December 20, 2004.
Brief description of amendment: The amendment deletes TS 5.6.1,
``Occupational Radiation Exposure Report,'' and TS 5.6.4, ``Monthly
Operating Reports,'' as described in the Notice of Availability
published in the Federal Register on June 23, 2004 (69 FR 35067).
Date of issuance: April 14, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 223.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2890).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 14, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York
Date of application for amendment: October 22, 2004.
Brief description of amendment: These amendments revise the
Technical Specifications by eliminating the requirements associated
with hydrogen recombiners and hydrogen monitors.
Date of issuance: April 14, 2005.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 243 and 228.
Facility Operating License Nos. DPR-26 and DPR-64: Amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5240).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 14, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York
Date of application for amendment: October 25, 2004.
Brief description of amendment: These amendments revise the
Technical Specifications by eliminating the requirements to submit
monthly operating reports and occupational radiation exposure reports.
Date of issuance: April 14, 2005.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 242 and 227.
Facility Operating License Nos. DPR-26 and DPR-64: Amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5241).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 14, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. STN
50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: April 30, 2004.
Brief description of amendments: The amendments modify the
technical specification (TS) requirements to adopt the provisions of
the industry/TS Task Force (TSTF) change TSTF-359, ``Increased
Flexibility in Mode Restraints.''
Date of issuance: April 5, 2005.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 141, 141, 134, 134.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 5, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: September 15, 2004.
Brief description of amendments: The proposed amendment will delete
the Technical Specification (TS) requirements related to hydrogen/
oxygen monitors. The proposed TS changes support implementation of the
revisions to Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.44, ``Standards for Combustible Gas Control System in Light-
Water-Cooled Power Reactors,'' that became effective on October 16,
2003. The changes are consistent with Revision 1 of the NRC-approved
Industry/Technical Specifications Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-447, ``Elimination of Hydrogen
Recombiners and Change to Hydrogen and Oxygen Monitors.''
Date of issuance: April 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 213/205.
Facility Operating License Nos. DPR-19, DPR-25: The amendments
revised the Technical Specifications.
[[Page 24659]]
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5243).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 28, 2005.
No significant hazards consideration comments received: No.
Exelon Generating Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: September 15, 2004.
Brief description of amendments: The amendments delete the
Technical Specification requirements to maintain hydrogen recombiners
and hydrogen/oxygen monitors and related Surveillance Requirements. The
revised Title 10 of the Code of Federal Regulations (10 CFR) Section
50.44, ``Combustible Gas Control for Nuclear Power Plants,'' eliminated
the requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design
qualification criteria for hydrogen and oxygen monitors.
Date of issuance: April 22, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 172/158.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5243).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 22, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: November 25, 2003.
Brief description of amendments: The amendment revised the
Technical Specifications (TSs) associated with Reactor Coolant System--
CHEMISTRY. Specifically, the amendment relocates Reactor Coolant
System--CHEMISTRY, in its entirety from the TSs to the Technical
Requirements Manual (TRM). In addition, the amendment deletes the
specific activity requirements related to E-Bar, gross beta and gross
gamma.
Date of issuance: April 18, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 174 and 136.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the TSs.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7522).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 18, 2005.
No significant hazards consideration determination comments
received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: November 17, 2004.
Brief description of amendment: The amendment eliminates the
requirements to submit monthly operating reports and annual
occupational radiation exposure reports.
Date of issuance: April 19, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 217.
Facility Operating License No. DPR-72: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7768).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 2005.
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: June 28, 2004.
Description of amendment request: The amendment revised the
Seabrook Station, Unit No. 1 Technical Specifications (TSs) to align
the language of Surveillance Requirement 4.9.4 with that of Limiting
Condition for Operation 3.9.4, ``Containment Building Penetrations.''
The amendment changes the requirement from ``during core alterations
and the movement of irradiated fuel'' to ``during the movement of
recently irradiated fuel.''
Date of issuance: April 21, 2005.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 102.
Facility Operating License No. NPF-86: The amendment revised the
TSs.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53110).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 21, 2005.
No significant hazards consideration comments received: No.
National Aeronautics and Space Administration, Docket No. 50-30, the
Plum Brook Test Reactor, Sandusky, Ohio
Date of application for amendment: January 14, 2005.
Brief description of amendment: The amendment clarifies the license
requirements for confirmation of Final Status Survey results prior to
backfilling or covering of excavated areas.
Date of issuance: April 21, 2005.
Effective date: The license amendment is effective as of its date
of issuance.
Amendment No.: 12.
Facility License No. TR-3: This amendment consists of changes to
the Facility License.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12743).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation enclosed with the amendment dated April 21,
2005.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410,
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New
York
Date of application for amendments: January 24, 2005.
Brief description of amendments: The amendments deleted Sections
6.6.1 and 5.6.1, ``Occupational Radiation Exposure Report,'' and
Sections 6.6.4 and 5.6.4, ``Monthly Operating Reports,'' from the NMP1
and NMP2 Technical Specifications.
Date of issuance: April 19, 2005.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 188 and 115.
Facility Operating License Nos. DPR-63 and NPF-69: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7769).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 19, 2005.
No significant hazards consideration comments received: No
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2, Luzerne County, Pennsylvania
Date of application for amendment: September 22, 2004.
Brief description of amendment: The amendment extended the validity
of the
[[Page 24660]]
reactor pressure vessel pressure-temperature limit curves from May 1,
2005, to May 1, 2006.
Date of issuance: April 25, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 197.
Facility Operating License No. NPF-22: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70721).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 25, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Dates of application for amendments: February 26 and April 28,
2004, as supplemented by letters dated July 8 and October 20, 2004.
Brief description of amendments: The amendments revised Technical
Specification (TS) Section 5.6.6, Reactor Coolant System (RCS) Pressure
Temperature Limits Report (PTLR), to facilitate future licensee-
controlled changes to the PTLR. The changes include a revised PTLR that
provides new heatup and cooldown limits and Cold Overpressure
Protection System (COPS) setpoints, and to recalculate the minimum size
of the pressurizer power operated relief valve orifice of the RCS vent.
In addition, the changes relocate the COPS arming temperature to the
PTLR, and lower the COPS arming temperature from 350 [deg]F to 220
[deg]F. The licensee also included TS bases changes to support the
changes to the TSs.
Date of issuance: March 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 136 (Unit 1) and 115 (Unit 2).
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19575) and April 22, 2004 (69 FR 34707)
The supplements dated July 8 and October 20, 2004, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 28, 2005.
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: December 2, 2004.
Brief description of amendments: The amendments modify technical
specification (TS) requirements for mode change limitations in Limiting
Condition for Operation 3.0.4 and Surveillance Requirement 4.0.4
consistent with Industry/TS Task Force (TSTF) Standard TS Change
Traveler, TSTF-359, Revision 9, ``Increased Flexibility in Mode
Restraints.'' A notice of availability for this TS improvement using
the Consolidated Line Item Improvement Process was published in the
Federal Register (FR) on April 4, 2003 (68 FR 16579).
Date of issuance: April 11, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 301, 290.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the TSs.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2901)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 11, 2005.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: September 23, 2003, as supplemented by
letter dated June 9, 2004.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) to extend the interval between local
leak rate tests for the containment purge and vent valves with
resilient seats (containment purge valves, hydrogen purge valves, and
containment pressure relief valves).
Date of issuance: April 13, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 116 and 116.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the TSs.
Date of initial notice in Federal Register: November 12, 2003 (68
FR 64140).
The supplement dated June 9, 2004, provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 13, 2005.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 13, 2004.
Brief description of amendment: The amendment revised Surveillance
Requirement (SR) 3.8.1.7 (fast-start test), SR 3.8.1.12 (safety
injection actuation signal test), SR 3.8.1.15 (hot restart test), and
SR 3.8.1.20 (redundant unit test) to clarify what voltage and frequency
limits are applicable during the transient and steady state portions of
the diesel generator start testing performed by these SRs.
Date of issuance: April 21, 2005.
Effective date: April 21, 2005, and shall be implemented within 90
days from the date of issuance.
Amendment No.: 161.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2904)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 21, 2005.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I,
[[Page 24661]]
which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment
[[Page 24662]]
under consideration. The contention must be one which, if proven, would
entitle the petitioner to relief. A petitioner/requestor who fails to
satisfy these requirements with respect to at least one contention will
not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of amendment request: April 11, 2005, as supplemented on April
14, 2005.
Description of amendment request: The amendments revise Technical
Specification (TS) 5.5.9, ``Steam Generator (SG) Tube Surveillance
Program,'' to incorporate changes in the SG inspection scope for
Braidwood Station, Unit 2 only, during refueling outage 11.
Date of issuance: April 25, 2005.
Effective date: April 25, 2005.
Amendment Nos.: 135, 135.
Facility Operating License Nos. NPF-72 and NPF-77: Amendment
revises the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Joliet Herald News, April 15 and 18, 2005,
and Morris Daily Herald, April 19, 2005. The announcement provided an
opportunity to submit comments on the Commission's proposed NSHC
determination. No comments have been received. The Commission's related
evaluation of the amendment, finding of exigent circumstances, state
consultation, and final NSHC determination are contained in a safety
evaluation dated April 25, 2005.
Attorney for licensee: Thomas S. O'Neil.
NRC Section Chief: Gene Y Suh.
Dated at Rockville, Maryland, this 2nd day of May 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. E5-2207 Filed 5-9-05; 8:45 am]
BILLING CODE 7590-01-P