[Federal Register Volume 69, Number 71 (Tuesday, April 13, 2004)]
[Notices]
[Pages 19561-19582]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-8047]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, March 19 through April 1, 2004. The last
biweekly notice was published on March 30, 2004 (69 FR 16615).
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
[[Page 19562]]
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 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
[[Page 19563]]
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 Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: February 27, 2004.
Description of amendment request: The licensee proposed to relocate
the average power range monitor (APRM)-based stability protection
settings for Option II stability solution to the Core Operating Limits
Report (COLR). The Option II solution demonstrates that existing
quadrant-based APRM trip systems will initiate a reactor scram for
postulated reactor instability and avoid violating the minimum critical
power ratio safety limit. Use of Option II was previously approved by
the Nuclear Regulatory Commission staff thru Amendment No. 235, dated
October 18, 2002.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes will relocate the Average Power Range
Monitor (APRM) based stability protection settings for the Option II
stability solution from the Technical Specifications (TS) to the
Core Operating Limits Report (COLR). The APRM based stability
protection settings are not an initiator or a precursor to an
accident. Furthermore, changes to the stability protection settings
do not physically modify or change the function, or system
interfaces, of the APRM Neutron Flux Scram and Neutron Flux Control
Rod Block systems or components. The APRM based stability protection
settings provide automatic protection to assure that anticipated
coupled neutronic/thermal-hydraulic instabilities will not
compromise established fuel safety limits. The proposed TS changes
cannot increase the consequences of a previously evaluated accident
because the changes do not alter any Limiting Safety System Setting,
but only relocate the applicable stability protection settings to
the COLR. The applicable stability protection settings will continue
to be determined by an NRC approved methodology.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes will relocate the APRM based stability
protection settings for the Option II stability solution from the TS
to the COLR. The APRM based stability protection settings for the
Option II stability solution assure anticipated coupled neutronic/
thermal-hydraulic instabilities will not compromise established fuel
safety limits. These changes do not introduce any new accident
precursors and do not involve any alterations to plant
configurations which could initiate a new or different kind of
accident. The proposed changes do not affect the intended function
of the APRM system nor do they affect the operation of the system in
a way which would create a new or different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will relocate the APRM based stability
protection settings for the Option II stability solution from the TS
to the COLR. The APRM based stability protection settings for
protection against reactor instability assure anticipated coupled
neutronic/thermal-hydraulic instabilities will not compromise
established fuel safety limits. No fuel thermal limits or other
design and licensing basis acceptance criteria are adversely
affected. No other events are adversely affected. The margin of
safety, as defined in the TS, for all events is maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Richard J. Laufer.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: March 8, 2004.
Description of amendment request: The proposed amendment would
delete Operating License Condition 2.C.(6) ``Long Range Planning
Program.'' The original objective of this requirement was to enable the
licensee to better control and manage resources regarding major
activities. The license condition does not have any direct effect on
plant design or operation. Since imposition of this requirement on May
27, 1988, the licensee has developed internal processes to control and
manage work activities, thus leading the licensee to determine that
this license condition is no longer needed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 19564]]
issue of no significant hazards consideration. The NRC staff has
reviewed the licensee's analysis against the three standards of 10 CFR
50.92(c). The NRC staff's analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The subject license condition was not a factor in the
scenario of any previously analyzed postulated design-basis accident or
anticipated operational transient. No hardware design change is
involved with the proposed amendment. Thus, the proposed deletion of
the license condition would create no adverse effect on the functional
performance of any plant structure, system, or component (SSC). All
SSCs will continue to perform their design functions with no decrease
in their capabilities to mitigate the previously analyzed consequences
of postulated accidents and anticipated operational transients.
Accordingly, the deletion of the license condition will lead to no
increase in the consequences of an accident previously evaluated, and
no increase in the probability of an accident previously evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment is not the result of a hardware
design change, nor does it lead to the need for a hardware design
change. There is no change in the methods the unit is operated. As a
result, all SSCs will continue to perform as previously analyzed by the
licensee, and previously evaluated and accepted by the NRC staff.
Therefore, the proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the proposed deletion of the
license condition will not lead the licensee to exceed or alter a
design basis or safety limit, and will not result in operating any
component in a less conservative manner, the proposed amendment will
not affect in any way the performance characteristics and intended
functions of any SSC. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Richard J. Laufer.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: March 8, 2004.
Description of amendment request: The proposed amendment would
delete Operating License Condition 2.C.(9) ``Long Range Planning
Program.'' The original objective of this requirement was to enable the
licensee to better control and manage resources regarding major
activities. The license condition does not have any direct effect on
plant design or operation. Since imposition of this requirement on May
27, 1988, the licensee has developed internal processes to control and
manage work activities, thus leading the licensee to determine that
this license condition is no longer needed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the three standards of 10 CFR 50.92(c). The NRC staff's
analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The subject license condition was not a factor in the
scenario of any previously analyzed postulated design-basis accident or
anticipated operational transient. No hardware design change is
involved with the proposed amendment. Thus, the proposed deletion of
the license condition would create no adverse effect on the functional
performance of any plant structure, system, or component (SSC). All
SSCs will continue to perform their design functions with no decrease
in their capabilities to mitigate the previously analyzed consequences
of postulated accidents and anticipated operational transients.
Accordingly, the deletion of the license condition will lead to no
increase in the consequences of an accident previously evaluated, and
no increase in the probability of an accident previously evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment is not the result of a hardware
design change, nor does it lead to the need for a hardware design
change. There is no change in the methods the unit is operated. As a
result, all SSCs will continue to perform as previously analyzed by the
licensee, and previously evaluated and accepted by the NRC staff.
Therefore, the proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the proposed deletion of the
license condition will not lead the licensee to exceed or alter a
design basis or safety limit, and will not result in operating any
component in a less conservative manner, the proposed amendment will
not affect in any way the performance characteristics and intended
functions of any SSC. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esquire, Vice
President, General Counsel and Secretary, Exelon Generation Company,
LLC, 300 Exelon Way, Kennett Square, PA 19348.
NRC Section Chief: Richard J. Laufer.
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: December 12, 2003.
Description of amendments request: The proposed amendment would
delete Technical Specification (TS) Section 5.5.3, ``Post-Accident
Sampling,'' requirements to maintain a Post-Accident Sampling System
(PASS). Licensees were generally required to implement PASS upgrades as
a result of 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
[[Page 19565]]
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 August 11, 2000 (65 FR 49271) 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 October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated December 12, 2003.
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 radionuclides
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 [a] 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 [a] 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: James M. Petro, Jr., 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. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: June 25, 2003.
Description of amendment request: The proposed amendments would
correct two inadvertent editorial changes made by Duke during the
submittal of Technical Specification (TS) Amendment 194/175 which
revised TS 3.3.1 (Reactor Trip System Instrumentation) and TS Amendment
197/178 which revised TS 4.2.1 (Design Features, Fuel Assemblies).
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 this LAR [License Amendment Request] involve a
significant increase in the probability or consequences of an
accident previously evaluated?
No. Approval and implementation of this LAR will have no affect
on accident probabilities or consequences since the proposed changes
are editorial in nature and were previously reviewed and approved by
the NRC [Nuclear Regulatory Commission].
2. Does this LAR create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. This LAR does not involve any physical changes to the plant.
Therefore, no new accident causal mechanisms will be generated. The
proposed changes are editorial in nature and were previously
reviewed and approved by the NRC. Consequently, plant accident
analyses will not be affected by these changes.
3. Does this LAR involve a significant reduction in a margin of
safety?
No. Margin of safety is related to the confidence in the ability
of the fission
[[Page 19566]]
product barriers to perform their design functions during and
following accident conditions. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
The performance of these barriers will not be affected by the
proposed changes since they are editorial in nature and have been
previously reviewed and approved by the NRC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
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: January 15, 2004, as supplemented by
letter dated March 15, 2004.
Description of amendment request: The proposed amendments would
revise the Technical Specifications associated with the control rod
drive (CRD) trip devices. These amendments are needed to support
implementation of the reactor trip breaker (RTB) replacement.
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) Involve a significant increase in the probability or
consequences of an accident previously evaluated[.]
The proposed LAR [license amendment request] modifies the
Technical Specifications [TS] to incorporate new TS requirements
associated with the new Control Rod Drive (CRD)/Reactor Trip Breaker
(RTB) configuration. The proposed LAR will continue to ensure that
the CRD trip devices will be operable to ensure that the reactor
remains capable of being tripped at any time it is critical.
Reliable CRD reactor trip circuit breakers and associated support
circuitry provides assurance that a reactor trip will occur when
initiated. The new RTBs will have the same seismic and quality group
qualifications as the existing components in the CRDCS [CRD control
system] system [sic]. The new RTBs will enhance the reliability of
the system by resolving age-related degradation issues and replacing
obsolete equipment. Therefore, the proposed LAR does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated[.]
The proposed LAR modifies the Technical Specifications to
incorporate new TS requirements associated with the new CRD/RTB
configuration. The systems affected by implementing the proposed
changes to the TS are not assumed to initiate design basis
accidents. Rather, the systems affected by the changes are used to
mitigate the consequences of an accident that has already occurred.
The proposed TS changes do not affect the mitigating function of
these systems. The reliability of the mitigating systems will be
improved by implementation of the RTB Upgrade. Consequently, these
changes do not alter the nature of events postulated in the Safety
Analysis Report nor do they introduce any unique precursor
mechanisms. Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) Involve a significant reduction in a margin of safety.
The proposed TS changes do not unfavorably affect any plant
safety limits, set points, or design parameters. The changes also do
not unfavorably affect the fuel, fuel cladding, RCS [reactor coolant
system], or containment integrity. Therefore, the proposed TS
change, which adds TS requirements associated with the CRD/RTB
upgrade, do not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
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 amendment request: March 3, 2004.
Description of amendment request: The proposed amendments would
revise the administrative Technical Specifications (TSs) for the
Reactor Coolant Pump Flywheel Inspection Program to extend the
allowable inspection interval to 20 years.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 24, 2003 (68 FR 37590), on possible amendments
to extend the inspection interval for reactor coolant pump (RCP)
flywheels, 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 October 22, 2003 (68
FR 60422). The licensee affirmed the applicability of the model NSHC
determination in its application dated March 3, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per
year). Moreover, considering the uncertainties involved in this
evaluation, the risk associated with the postulated failure of an
RCP motor flywheel is significantly low. Even if all four RCP motor
flywheels are considered in the bounding plant configuration case,
the risk is still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits; or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
[[Page 19567]]
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or alter the methods governing normal plant operation. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements, and does not
alter any assumptions made in the safety analysis. The proposed
change is consistent with the safety analysis assumptions and
current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel. Therefore,
the proposed 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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: March 9, 2004.
Description of amendment request: The proposed amendment would
extend the completion time (CT) from 1 hour to 24 hours for Condition B
of Technical Specification (TS) 3.5.1, ``Accumulators.'' The
accumulators are part of the emergency core cooling system and consist
of tanks partially filled with borated water and pressurized with
nitrogen gas. The contents of the tank are discharged to the reactor
coolant system (RCS) if, as during a loss-of-coolant accident, the
coolant pressure decreases to below the accumulator pressure. Condition
B of TS 3.5.1 specifies a CT to restore an accumulator to operable
status when it has been declared inoperable for a reason other than the
boron concentration of the water in the accumulator not being within
the required range. This change was proposed by the Westinghouse Owners
Group participants in the TS Task Force (TSTF) and is designated TSTF-
370. TSTF-370 is supported by NRC-approved Topical Report WCAP-15049-A,
``Risk-Informed Evaluation of an Extension to Accumulator Completion
Times,'' submitted on May 18, 1999. The NRC staff issued a notice of
opportunity for comment in the Federal Register on July 15, 2002 (67 FR
46542), on possible amendments concerning TSTF-370, 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 March 12, 2003 (68 FR 11880). The licensee affirmed the
applicability of the following NSHC determination in its application
dated March 9, 2004.
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 basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of WCAP-15049-A, the proposed change will
allow plant operation with an inoperable accumulator for up to 24
hours, instead of 1 hour, before the plant would be required to
begin shutting down. The impact of the increase in the accumulator
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total
plant core damage frequency (CDF) less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049-A for the accumulator CT increase meet the criterion of
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications,'' for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049-A, design
basis accumulator success criteria are not considered necessary to
mitigate large break loss-of-coolant accident (LOCA) events, and
were only included in the WCAP-15049-A evaluation as a worst case
data point. In addition, WCAP-15049-A states that the NRC has
indicated that an incremental conditional core damage frequency
(ICCDP) greater than 5E-07 does not necessarily mean the change is
unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
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
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049-A evaluation demonstrates that the small
increase in risk due to increasing the CT for an inoperable
accumulator is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1.1, is to ensure that a sufficient volume of borated
[[Page 19568]]
water will be immediately forced into the core through each of the
cold legs in the event the RCS pressure falls below the pressure of
the accumulators, thereby providing the initial cooling mechanism
during large RCS pipe ruptures. As described in Section 9.2 of WCAP-
15049-A, the proposed change will allow plant operation with an
inoperable accumulator for up to 24 hours, instead of 1 hour, before
the plant would be required to begin shutting down. The impact of
this on plant risk was evaluated and found to be very small. That
is, increasing the time the accumulators will be unavailable to
respond to a large LOCA event, assuming accumulators are needed to
mitigate the design basis event, has a very small impact on plant
risk.
Since the frequency of a design basis large LOCA (a large LOCA
with loss of offsite power) would be significantly lower than the
large LOCA frequency of the WCAP-15049-A evaluation, the impact of
increasing the accumulator CT from 1 hour to 24 hours on plant risk
due to a design basis large LOCA would be significantly less than
the plant risk increase presented in the WCAP-15049-A evaluation.
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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: December 24, 2003.
Description of amendment request: The proposed amendment would
delete requirements in the Pilgrim Nuclear Power Station Technical
Specifications (TSs) 3.7.A.7.c and 4.7.A.7.c, associated with hydrogen
analyzers. The NRC staff issued a notice of opportunity for comment in
the Federal Register on August 2, 2002 (67 FR 50374), on possible
amendments to eliminate the hydrogen analyzers from TSs, including a
model safety evaluation and model no significant hazards consideration
(NSHC) determination, using the Consolidated Line Item Improvement
Process (CLIIP). The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on September 25, 2003 (68 FR
55416). The licensee affirmed the applicability of the relevant
portions of the model NSHC determination in its application dated
December 24, 2003.
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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
RG [Regulatory Guide] 1.97 Category 1, is intended for key variables
that most directly indicate the accomplishment of a safety function
for design-basis accident events. The hydrogen monitors no longer
meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44, the Commission found that
Category 3, as defined in RG 1.97, is an appropriate categorization
for the hydrogen monitors because the monitors are required to
diagnose the course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed 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.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the SAMGs [Severe Accident
Management Guidelines], 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 the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or 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 the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides 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 hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, 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.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate 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 intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
Based upon the reasoning presented above, the requested change does
not involve a significant hazards consideration. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: J.M. Fulton, Esquire, Assistant General
Counsel,
[[Page 19569]]
Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts 02360-5599.
NRC Section Chief: Darrell J. Roberts, Acting.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 15, 2004.
Description of amendment request: The amendment proposes to move
the Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical
Specification (TS) 3.4.8.2, pressurizer heatup and cooldown limits to
the Technical Requirements Manual (TRM), which is reviewed in
accordance with Section 50.59 of Title 10 of the Code of Federal
Regulations (10 CFR), ``Changes, tests, and experiments.'' The
associated action statement, surveillance requirement, and bases are
also proposed for relocation.
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 probability of an accident is unchanged as a result of the
proposed change to delete the Waterford 3 pressurizer heatup and
cooldown rates and associated action, surveillance requirement, and
bases from the TS. The cooldown and heatup rates are not initiators
to any accidents or pressurizer transients discussed in the
Waterford 3 Final Safety Analysis Report (FSAR). Therefore, the
probability of an accident is not changed.
The purpose of the pressurizer heatup and cooldown limits is to
ensure that given transient events will not negatively affect the
pressurizer structural integrity beyond Code allowables. These
limits will be maintained within ASME [American Society of
Mechanical Engineers] Code allowables in the TRM in accordance with
10 CFR 50.59. Therefore, the consequences of an accident are not
increased.
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 limitations imposed on the pressurizer heatup and cooldown
rates are provided to assure that the pressurizer is operated within
the design criteria assumed for the flaw evaluation and fatigue
analysis performed in accordance with the ASME Code Section XI,
subsection IWB-3600 requirements. The Waterford 3 FSAR has analyzed
the conditions that would result from a thermal or pressurization
transient on the Waterford 3 pressurizer. The proposed deletion of
the pressurizer heatup and cooldown rates and relocation of the
limits to the TRM does not change the way that the pressurizer is
designed or operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established by the rules contained in
the ASME Section III Code. Any future changes to the cooldown or
heatup rates will be evaluated using 10 CFR 50.59 and are required
to meet the ASME Code margins.
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: N.S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: March 12, 2004.
Description of amendment request: The proposed amendments would
modify the Technical Specifications (TS) to eliminate selected response
time testing (RTT) requirements associated with Reactor Protection
System instrumentation and Primary Containment Isolation
instrumentation for Main Steam Line Isolation functions. The proposed
changes are consistent with the Boiling Water Reactor Owners Group
(BWROG) Licensing Topical Report ``System Analyses for the Elimination
of Selected Response Time Testing Requirements,'' NEDO-32291A,
Supplement 1, dated October 1999, as approved by the NRC on June 11,
1999.
The original Licensing Topical Report (LTR) NEDO-32291-A, dated
October 1995, established a generic basis for elimination of many RTTs
for instrument loops that had good performance histories and longer
response time requirements. The justification was based on the adequacy
of surveillance tests other than RTTs to assure that response time
requirements were met for sensors in those loops. Supplement 1 to NEDO-
32291-A was prepared to document an analysis to extend the conclusions
of the original study to cover the logic components in selected
instrumentation loops that have intermediate length response time
requirements. The intent was to demonstrate that elimination of the RTT
requirements for the logic portions of those loops is of no safety
significance. Supplement 1 concludes, for instrument loops meeting the
application criteria of the Licensing Topical Report, that performance
of ongoing TS required surveillance tests other than RTTs (i.e.,
calibration tests, functional tests, and logic system functional tests)
provides adequate assurance that those instrument loops will meet their
respective response time requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in probability or consequences of an accident previously evaluated.
The proposed amendment to the TS eliminates selected RTT
requirements in accordance with the NRC approved BWROG LTR.
Elimination of RTT for selected instrumentation in the Reactor
Protection System and Primary Containment Isolation Instrumentation
does not result in the alteration of the design, material, or
construction standards that were applicable prior to the proposed
change. The response time assumptions used in the accident analyses
remain unchanged. Only the methodology used for response time
verification is changed. All component models used in the affected
trip channels were analyzed for a bounding response time. As
documented in the BWROG LTR and supplement, a degraded response time
will be detected by other TS required tests. The bounding response
time of the relays discussed in the supplement to the LTR can be
used in place of actual measured response times to ensure that the
instrumentation systems will meet the response time requirements of
the accident analysis.
The proposed change will not result in the modification of any
system interface that would increase the likelihood of an accident
since these events are independent of the proposed change. In
addition, the proposed amendment will not change, degrade, or
prevent actions, or alter any assumptions previously made in
evaluating the radiological consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 19570]]
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed action does not involve physical alteration of the
station. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
is no change being made to the parameters within which LaSalle is
operated. There are no setpoints at which protective or mitigative
actions are initiated that are affected by this proposed action. All
Reactor Protection System and Primary Containment Isolation
Instrumentation channels affected by the proposed change will
continue to have an initial response time verified by test before
initially placing the channel in service and after any maintenance
that could affect response time.
The proposed change does not alter assumptions made in the
safety analysis. A review of the failure modes of the affected
sensors and relays indicates that a sluggish response of the
instruments can be detected by other TS surveillances. Changing the
method of periodically verifying instrument response for the
selected instrument channels will not create any new accident
initiators or scenarios. Periodic surveillance of these instruments
will detect significant degradation in the channel characteristic.
This proposed action will not alter the manner in which equipment
operation is initiated, nor will the function demands on credited
equipment be changed. No change is being made to procedures relied
upon to respond to an off-normal event. As such, no new failure
modes are being introduced.
The sensors and relays in the affected channels will be able to
meet the bounding response times as defined and presented in the LTR
Supplement. It has been found acceptable to use component bounding
response times in place of actual measured response times to ensure
that instrumentation systems will meet response time requirements of
the accident analyses. In addition, [Exelon Generation Company, LLC]
EGC's adherence to the conditions listed in the NRC Safety
Evaluations for the LTR and Supplement provides additional assurance
that the instrumentation systems will meet the response time
requirements of the accident analyses.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Implementation of the BWROG LTR methodologies for eliminating
selected response time testing requirements does not involve a
significant reduction in the margin of safety. The current response
time limits are based on the maximum values assumed in the plant
safety analyses. The analyses conservatively establish the margin of
safety. The elimination of the selected response time testing does
not affect the capability of the associated systems to perform their
intended function within the allowed response time used as the basis
for plant safety analyses. Plant and system response to an
initiating event will remain in compliance within the assumptions of
the safety analyses, and therefore, the margin of safety is not
affected. This is based on the ability to detect a degraded response
time of an instrument or relay by the other required TS tests,
component reliability, and redundancy and diversity of the affected
functions, as justified in the reviewed and approved LTR and
Supplement.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: February 27, 2004.
Description of amendment request: The proposed change to the
Technical Specifications (TSs) supports the activation of the trip
outputs of the previously-installed Oscillation Power Range Monitor
(OPRM) portion of the Power Range Neutron Monitoring (PRNM) system.
Specifically, this proposed change will revise TS Sections 3.3.1.1,
``Reactor Protection System Instrumentation,'' and 3.4.1,
``Recirculation Loops Operating Reporting Requirements,'' and their
associated TS Bases, and TS Section 5.6.5, ``Core Operating Limits
Report (COLR).'' In addition, the proposed change deletes the Interim
Corrective Action requirements from the Recirculation Loops Operating
TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. This modification has no impact on any of the
previously installed PRNM functions. Plant operation in portions of
the former restricted region may potentially cause a marginal
increase in the probability of occurrence of an instability event.
This potential increase in probability is acceptable because the
OPRM function will automatically detect the condition and initiate a
reactor scram before the Minimum Critical Power Ratio (MCPR) Safety
Limit is reached. Consequences of the potential instability event
are reduced because of the more reliable automatic detection and
suppression of an instability event, and the elimination of
dependence on the manual operator actions.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The modification replaces procedural actions that
were established to avoid operating conditions where reactor
instabilities might occur with an NRC approved automatic detect and
suppress function.
Potential failures in the OPRM Upscale function could result in
either failure to take the required mitigating action or an
unintended reactor scram. These are the same potential effects of
failure of the operator to take the correct appropriate action under
the current procedural actions. The net effect of the modification
changes the method by which an instability event is detected and by
which mitigating action is initiated, but does not change the type
of stability event that could occur. The effects of failure of the
OPRM equipment are limited to reduced or failed mitigation, but such
failure cannot cause an instability event or other type of accident.
Therefore, since no radiological barrier will be challenged as a
result of activating the OPRM trip function, it is concluded that
this proposed activity does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The current safety analyses assume that the
existing procedural actions are adequate to prevent an instability
event. As a result, there is currently no quantitative or
qualitative assessment of an instability event with respect to its
impact on MCPR.
The OPRM trip function is being implemented to automate the
detection (via direct measurement of neutron flux) and subsequent
suppression (via scram) of an instability event prior to exceeding
the MCPR Safety Limit. The OPRM trip provides a trip output of the
same type as currently used for the Average Power Range Monitor
(APRM). Its failure modes and types are identical to those for the
present APRM output. Currently, the MCPR Safety Limit is not
impacted by an instability event since the event is ``mitigated'' by
manual means via the procedural actions, which prevent plant
operating conditions where an instability event is possible. In both
methods of mitigation (manual and automated), the margin of safety
associated with the MCPR Safety Limit is maintained.
Therefore, since the MCPR Safety Limit will not be exceeded as a
result of an
[[Page 19571]]
instability event following implementation of the OPRM trip
function, it is concluded that the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: Mr. Edward Cullen, Vice President and
General Counsel, Exelon Generation Company, LLC, 2301 Market Street,
S23-1, Philadelphia, PA 19101.
NRC Section Chief: Darrell Roberts, Acting.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: February 27, 2004.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) Section 5.6.2.6, ``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 February 27, 2004.
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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602-1551.
NRC Section Chief: William F. Burton, Acting.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: January 28, 2004.
Description of amendment request: Duane Arnold Energy Center
[[Page 19572]]
implemented improved technical specifications in 1998 via Amendment 223
using NUREG 1433, ``Standard Technical Specifications--General Electric
Plants BWR/4,'' Revision 1, as a model. The proposed amendment would
revise Technical Specification Sections 5.5.11, 1.4, 3.3.1.1, and 5.5.2
to adopt the following selected NRC approved generic changes to the
improved technical specification NUREG.
Technical Specification Task Force (TSTF)-273,
Revision 2, Safety Function Determination Program Clarifications.
TSTF-284, Revision 3, Add ``Met'' versus
``Perform'' to Specification 1.4, Frequency.
TSTF-264, Deletion of Flux Monitors Specific
Overlap Surveillance Requirements.
TSTF-299, Administrative Controls Program
5.5.2.b Test Interval Defined and Allowance for 25 Percent Extension of
Frequency.
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:
Adoption of TSTF-273, Revision 2, and TSTF-284, Revision 3
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves reformatting, renumbering, and
rewording the existing Technical Specifications. The reformatting,
renumbering, and rewording process involves no technical changes to
the existing Technical Specifications. As such, this change is
administrative in nature and does not affect initiators of analyzed
events or assumed mitigation of accident or transient events.
Therefore, 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?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in methods governing normal plant operation. The proposed
change will not impose any new or eliminate any old requirements.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Response: No.
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses' assumptions. This change is
administrative in nature. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
Adoption of TSTF-264, Revision 0
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change deletes Surveillance Requirements.
Surveillances are not initiators to any accident previously
evaluated. Consequently, the probability of an accident previously
evaluated is not significantly increased. The equipment being tested
is still required to be Operable and capable of performing the
accident mitigation functions assumed in the accident analysis. As a
result, the consequences of any accident previously evaluated are
not significantly affected. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
remaining Surveillance Requirements are Consistent with industry
practice and are considered to be sufficient to prevent the removal
of the subject Surveillances from creating a new or different type
of accident. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Response: No.
The deleted Surveillance Requirements do not result in a
significant reduction in the margin of safety. As provided in the
justification, the change has been evaluated to ensure that the
deleted Surveillance Requirements are not necessary for verification
that the equipment used to meet the LCO [limiting condition for
operation] can perform its required functions. Thus, appropriate
equipment continues to be tested in a manner and at a frequency
necessary to give confidence that the equipment can perform its
assumed safety function. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
Adoption of TSTF-299, Revision 0
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change provides more stringent requirements for
operation of the facility. These more stringent requirements do not
result in operation that will increase the probability of initiating
an analyzed event and do not alter assumptions relative to
mitigation of an accident or transient event. The more restrictive
requirements continue to ensure process variables, structures,
systems, and components are maintained consistent with the safety
analyses and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in methods governing normal plant operation. The proposed
change does impose different requirements. However, these changes
are consistent with the assumptions in the safety analyses and
licensing basis. Thus, this change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Response: No.
The proposed change provides additional restrictions which
enhance plant safety. This change maintains requirements within the
safety analyses and licensing basis. 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 request involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Morgan Lewis, 1111
Pennsylvania Avenue NW., Washington, DC 20004.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: February 27, 2004.
Description of amendment request: The proposed amendment would
remove license condition 2.C.(2)(b) to perform large transient testing
as part of the extended power uprate (EPU) power ascension testing
program at the Duane Arnold Energy Center (DAEC).
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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The requested licensing action would remove the current
requirement to perform specific large transient tests as part of the
DAEC EPU power ascension testing program. No other changes are
proposed. Therefore,
[[Page 19573]]
the probability of an accident previously evaluated is not
significantly increased.
The proposed action will not affect any System, Structure, or
Component designed for the mitigation of previously analyzed events.
The proposed change does not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Thus, the proposed change will not increase the consequences of any
previously evaluated accident.
(2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The requested licensing action would remove the current
requirement to perform specific large transient tests as part of the
DAEC EPU power ascension testing program. No other changes are
proposed. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) The proposed amendment will not involve a significant
reduction in a margin of safety.
Performance of these specific large transient tests is not
necessary to ensure acceptable plant operation at the higher thermal
power level. Simple, integrated systems tests are performed in lieu
of the complex, challenging large transient tests. Other required
testing of the specific SSCs that have been modified for EPU ensures
that the plant will respond as expected during any abnormal
operating event, including these specific transients. Thus, the
proposed elimination of the large transient tests will not
significantly reduce any margin of safety from that previously
approved for EPU operation at the DAEC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Morgan Lewis, 1111
Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: January 30, 2004.
Description of amendment request: The proposed amendment would
revise Monticello Nuclear Generating Plant (MNGP) Technical
Specifications (TS) to (1) clarify the permissive set point for the
source range monitor (SRM) detector not-fully-inserted rod block
bypass, (2) correct a typographical error in the surveillance
requirement for suppression pool temperature monitoring, (3) clarify
the set point for the pressure suppression chamber-reactor building
vacuum breakers instrumentation, (4) clarify the operating force
requirements for the pressure suppression chamber--drywell vacuum
breakers surveillance test, and (5) make corrections resulting from
License Amendments (LAs) 130 and 132.
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 SRM Detector-not-fully-inserted rod block bypass set point,
the Pressure Suppression Chamber--Reactor Building Vacuum Breakers
actuation instrumentation set point requirement and the Pressure
Suppression Chamber--Drywell Vacuum Breakers surveillance test
requirements are being clarified in the MNGP TS to ensure these
functions will adequately support safe operation of the facility.
Typographical errors are being corrected along with corrections
resulting from omissions and an oversight from previous LAs. The
proposed TS changes do not introduce new equipment or new equipment
operating modes, nor do the proposed changes alter existing system
relationships. The changes do not affect plant operation, design
function or any analysis that verifies the capability of a SSC
[structure, system or component] to perform a design function.
Further, the proposed changes do not increase the likelihood of the
malfunction of any structure, system or component (SSC) or impact
any analyzed accident. Consequently, the probability of an accident
previously evaluated is not affected.
Therefore, the proposed amendment does 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 SRM Detector-not-fully-inserted rod block bypass set point,
the Pressure Suppression Chamber--Reactor Building Vacuum Breakers
actuation instrumentation set point requirement and the Pressure
Suppression Chamber--Drywell Vacuum Breakers surveillance test
requirements are being clarified in the MNGP TS to ensure these
functions will adequately support safe operation of the facility.
Typographical errors are being corrected along with corrections
resulting from omissions and an oversight from previous LAs. The
changes do not create the possibility of new credible failure
mechanisms, or malfunctions. These changes do not modify the design
function or operation of any SSC. Further the changes do not involve
physical alterations of the plant; no new or different type of
equipment will be installed. The proposed changes do not introduce
new accident initiators. Consequently, the changes cannot create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously analyzed.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
The SRM Detector-not-fully-inserted rod block bypass set point,
the Pressure Suppression Chamber--Reactor Building Vacuum Breakers
actuation instrumentation set point requirement and the Pressure
Suppression Chamber--Drywell Vacuum Breakers surveillance test
requirements are being clarified in the MNGP TS to ensure these
functions will adequately support safe operation of the facility.
Typographical errors are being corrected along with corrections
resulting from omissions and an oversight from previous LAs. These
changes do not exceed or alter a design basis or a safety limit for
a parameter established in the MNGP Updated Safety Analysis Report
(USAR) or the MNGP facility license. Consequently, the changes do
not result in a significant reduction in the margin of safety.
Therefore, the proposed amendment does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: February 10, 2004.
Description of amendment request: The proposed change involves the
extension from 1 hour to 24 hours of the completion time (CT) for
Action (a) of Technical Specification (TS) 3.5.1.1, which defines
requirements for accumulators. Accumulators are part of the emergency
core cooling system and consist of tanks partially filled with borated
water and pressurized with nitrogen gas. The contents of the tank are
discharged to the reactor coolant system (RCS) if, as during a loss-of-
coolant accident, the coolant pressure
[[Page 19574]]
decreases to below the accumulator pressure. Action (a) of TS 3.5.1.1
specifies a CT to restore an accumulator to operable status when it has
been declared inoperable for a reason other than the boron
concentration of the water in the accumulator not being within the
required range. This change was proposed by the Westinghouse Owners
Group participants in the TS Task Force (TSTF) and is designated TSTF-
370. TSTF-370 is supported by NRC-approved topical report WCAP-15049-A,
``Risk-Informed Evaluation of an Extension to Accumulator Completion
Times,'' submitted on May 18, 1999. The NRC staff issued a notice of
opportunity for comment in the Federal Register on July 15, 2002 (67 FR
46542), on possible amendments concerning TSTF-370, 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 March 12, 2003 (68 FR 11880). The licensee included in its
application several minor changes to make the plant specific TS more
consistent with the STS and TSTF-370. The licensee affirmed the
applicability of the following NSHC determination in its application
dated February 10, 2004.
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 basis for the accumulator limiting condition for operation
(LCO), as discussed in Basis Section 3.5.1.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of WCAP-15049-A, the proposed change will
allow plant operation with an inoperable accumulator for up to 24
hours, instead of 1 hour, before the plant would be required to
begin shutting down. The impact of the increase in the accumulator
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total
plant core damage frequency (CDF) less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049-A for the accumulator CT increase meet the criterion of
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications,'' for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049-A, design
basis accumulator success criteria are not considered necessary to
mitigate large break loss-of-coolant accident (LOCA) events, and
were only included in the WCAP-15049-A evaluation as a worst case
data point. In addition, WCAP-15049-A states that the NRC has
indicated that an incremental conditional core damage frequency
(ICCDP) greater than 5E-07 does not necessarily mean the change is
unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
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
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049-A evaluation demonstrates that the small
increase in risk due to increasing the CT for an inoperable
accumulator is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Basis Section
3.5.1.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable
accumulator for up to 24 hours, instead of 1 hour, before the plant
would be required to begin shutting down. The impact of this on
plant risk was evaluated and found to be very small. That is,
increasing the time the accumulators will be unavailable to respond
to a large LOCA event, assuming accumulators are needed to mitigate
the design basis event, has a very small impact on plant risk.
Since the frequency of a design basis large LOCA (a large LOCA
with loss of offsite power) would be significantly lower than the
large LOCA frequency of the WCAP-15049-A evaluation, the impact of
increasing the accumulator CT from 1 hour to 24 hours on plant risk
due to a design basis large LOCA would be significantly less than
the plant risk increase presented in the WCAP-15049-A evaluation.
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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
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
Date of amendment request: February 20, 2004.
Description of amendment request: The proposed amendments would
revise Vogtle Electric Generating Plant, Units 1 and 2 Administrative
Controls Section 5.2.2.g of Technical Specification to limit the
requirement of the Shift Technical Advisor function to Modes 1-4 in
accordance with NUREG 0737.
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?
The proposed change to TS [Technical Specification] 5.2.2.g does
not significantly
[[Page 19575]]
increase the probability or consequences of an accident previously
evaluated in the FSAR [Final Safety Analysis Report]. This revision
does not have any effect on the probability of any accident
initiators. The consequences of accidents previously evaluated in
the FSAR are not adversely affected by this proposed change because
the STA [Shift Technical Advisor] is not credited for mitigation of
any accidents. The proposed change which requires the STA function
to be available while in Modes 1-4 is in accordance with the
requirements of NUREG 0737, Item I.A.1.1. Consequently, the
probability or consequences of an accident previously evaluated are
not significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
The proposed change to TS 5.2.2.g does not create the
possibility of a new or different kind of accident from any
previously evaluated. No new accident scenarios, failure mechanism,
or limiting single failures are introduced as a result of the
proposed change. The proposed Technical Specifications change does
not challenge the performance or integrity of any safety-related
systems. The proposed change to TS 5.2.2.g is in accordance with
NUREG 0737.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed change to TS 5.2.2.g will not reduce a margin of
safety because it has no direct effect on any safety analyses
assumptions. The STA function is to evaluate plant conditions and
provide advice to the shift supervisor during plant transients and
accidents. The proposed change limits the requirements for the STA
function to Modes 1-4 in accordance with NUREG 0737. The STA
function is not credited for the mitigation of any accidents
previously evaluated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: John A. Nakoski.
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
Date of amendment request: February 26, 2004.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 5.6.6, ``Reactor Coolant System
(RCS) Pressure and Temperature Limits Report (PTLR)'', to reference the
NRC-approved methodology for developing Pressure-Temperature limits and
Cold Overpressure Protection System setpoints and the methodology used
to justify eliminating the reactor vessel closure head/vessel flange
requirements. The proposed amendment would also revise TS 3.4.12,
``Cold Overpressure Protection System (COPS)'', to change the Reactor
Coolant System vent size.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes to the Technical Specifications [TS]
and PTLRs [Pressure and Temperature Limits Reports] do not affect
any plant equipment, test methods, or plant operation, and are not
initiators of any analyzed accident sequence. Operation in
accordance with the proposed TS will ensure that all analyzed
accidents will continue to be mitigated by the SSCs [systems,
structures and components] as previously analyzed.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed changes do not introduce any new equipment,
create new failure modes for existing equipment, or create any new
limiting single failures. The changes to the P-T [pressure-
temperature] limits and COPS [Cold Overpressure Protection Systems]
setpoints will ensure that appropriate fracture toughness margins
are maintained to protect against reactor vessel failure during both
normal and low temperature operation. The changes to the P-T limits
and COPS setpoints are consistent with the methodology approved by
the NRC [Nuclear Regulatory Commission] in WCAP-14040, Rev. 4. Plant
operation will not be altered, and all safety functions will
continue to perform as previously assumed in accident analyses.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The proposed changes will not adversely affect the operation
of plant equipment or the function of any equipment assumed in the
accident analysis. The utilization of ASME [American Society of
Mechanical Engineers] Code Case N-640 maintains the relative margin
of safety commensurate with that which existed at the time that ASME
B&PV [Boiler and Pressure Vessel] Code, Section XI, Appendix G was
approved in 1974 and will ensure an acceptable margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: John A. Nakoski.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant (BFN), Unit 1, Limestone County, Alabama
Date of amendments request: March 9, 2004 (TS 434).
Description of amendments request: The proposed amendment would
lower the current Reactor Vessel Water Level--Low, Level 3 Allowable
Value in the Unit 1 Technical Specifications for several instrument
functions to reduce the likelihood of unnecessary reactor scrams and
the resultant engineered safety feature actuations by increasing the
operating range between the normal reactor vessel water level and Level
3 trip functions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The Reactor Vessel Water Level--Low, Level 3 functions are
in response to water level transients and are not involved in the
initiation of accidents or transients. Therefore, reducing the BFN,
Unit 1, Level 3 Allowable Value does not increase the probability of
an accident previously evaluated.
Additionally, the results of the safety evaluation associated
with the lowering of the Level 3 Allowable Value concludes that the
previously evaluated transient and accident consequences are not
significantly affected by the change. Therefore, the proposed
amendment does not involve a significant increase in the probability
of consequences or 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?
No. The proposed amendment to lower the BFN, Unit 1, Reactor
Vessel Water Level--Low, Level 3 Allowable Value does not involve a
hardware change and the purpose of the Level 3 function is not
affected. The Level 3 functions will continue to fulfill their
design objective. The proposed changes do not create the possibility
of any new failure mechanisms. No new external threats or release
pathways are created. Therefore, reduction of the Allowable Value
does not result in the possibility of a new or different kind of
accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
[[Page 19576]]
No. The results of the safety evaluation associated with the
reducing the BFN, Unit 1, Reactor Vessel Water Level--Low, Level 3
Allowable Value concluded that transient and accident consequences
remain within the required acceptance criteria. Therefore, the
margin of safety is not reduced for any event evaluated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: William F. Burton, Acting.
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 5, 2004.
Description of amendment request: The proposed amendments would
delete Technical Specifications (TSs) 3.6.4.1, ``Hydrogen Monitors,''
and 3.6.4.2, ``Electric Hydrogen Recombiners-W.'' The proposed changes
support Title 10, Code of Federal Regulations, Part 50, Section 44 (10
CFR 50.44), ``Standards for Combustible Gas Control system in Light-
Water-Cooled Power Reactors'' and are consistent with the Industry/
Technical Specification Task Force (TSTF) Standard TS Change Traveler,
TSTF-447, ``Elimination of Hydrogen Recombiners and change to Hydrogen
and Oxygen Monitors.''
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:
TVA has reviewed the proposed no significant hazards
consideration determination published on September 25, 2003, (68 FR
55416) as part of the consolidated line item improvement process
(CLIIP). TVA has concluded that the proposed determination presented
in the notice is applicable to SQN, and the determination is hereby
incorporated by reference to satisfy the requirements of 10 CFR
50.91(a).
The United States Nuclear Regulatory Commission (NRC) staff has
reviewed the licensee's analysis and, based on this review, it appears
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: William F. Burton, Acting.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 5, 2004.
Description of amendment request: The proposed amendments would
delete surveillance requirement (SR) 4.9.2.c and SRs 4.10.3.2 and
4.10.4.2 from the Technical Specifications (TSs). SR 4.9.2.c requires
channel functional tests for each Source Range neutron flux monitor
within 8 hours prior to initial core alterations. SRs 4.10.3.2 and
4.10.4.2 require channel functional tests for each Power Range and
Intermediate Range neutron flux monitor within 12 hours prior to the
initiation of a physics test. In addition, the proposed changes include
revisions to the associated TS bases (3/4.9.2, 3/4.10.3, and 3/4.10.4).
Basis for proposed no significant hazards consideration
determination: As required by Title 10, Code of Federal Regulations,
Part 50, Section 91(a) (10 CFR 50.91(a)), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed amendment removes the requirement to perform an
additional CHANNEL FUNCTIONAL TEST (CFT) on the Intermediate and
Power Range functions within 12 hours of performing a PHYSICS TEST.
The Intermediate and Power Range instrumentation is determined to be
OPERABLE by periodic SRs which must be confirmed to be within
frequency prior to making the reactor critical. The proposed
amendment also removes the requirement to perform an additional CFT
on the Source Range monitors. The Source Range instrumentation is
determined to be OPERABLE by periodic SRs, which must be confirmed
to be within frequency prior to Mode 6, prior to CORE ALTERATIONS,
and must remain OPERABLE. A CFT for the Source Range, Intermediate
Range, or Power Range instrumentation is not a precursor to, or
assumed to be an initiator of any analyzed accident. Therefore, this
change does not involve a significant increase in the probability of
an accident previously evaluated.
Regarding a significant increase in the consequences of an
accident, several factors must be considered. First the PHYSICS
TESTS are performed in accordance with the TSs in Mode 2. Therefore,
the power level of the reactor is limited to 5 percent or less.
Along with this, the reactor trip function of the Intermediate Range
detectors will be unaffected by the proposed amendment and
therefore, will be available to mitigate a reactivity transient at
low power. Further, the trip setpoint for the Power Range monitors
are decreased during startup. This setpoint reduction provides an
additional measure to limit a reactivity excursion. 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?
No. The proposed changes permit the conduct of normal operating
evolutions during limited periods when additional controls over
reactivity margin are imposed by the TSs. The proposed change does
not introduce any new equipment into the plant or significantly
alter the manner in which existing equipment will be operated. The
proposed changes are not based on a change in the design or
configuration of the plant. The changes to operating allowances are
minor and are only applicable during certain conditions. The
operating allowances are consistent with those acceptable at other
times. The proposed changes delete the requirements for the
performance of a CFT for the Source Range, Intermediate Range, and
Power Range instrumentation within 8 hours of initiating CORE
ALTERATIONS for the Source Range monitors and within 12 hours of
starting a PHYSICS TEST for the Intermediate Range and Power Range
instrumentation. Since the proposed changes only allow activities
that are presently approved and routinely conducted, no possibility
exists for a new or different kind of accident from those previously
evaluated. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. As stated previously, the proposed change deletes the
requirement to perform an additional CFT for the Source Range,
Intermediate Range, and Power Range instrumentation within 8 hours
of initiating CORE ALTERATIONS for the Source Range monitors and
within 12 hours of starting a PHYSICS TEST for the Intermediate
Range and Power Range instrumentation. The Source Range,
Intermediate Range, and Power Range instrumentation channels are
determined to be OPERABLE by meeting the requirements of the
periodic surveillance. These SRs are not affected by the proposed
amendment. The proposed changes do not involve a significant
reduction in a margin of safety because the ability to monitor the
reactor during the applicable operating conditions and modes of
operation will be maintained. The proposed changes do not affect
these operating restrictions and the margin of safety which assures
the ability to
[[Page 19577]]
monitor the reactor is not affected. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
The United States Nuclear Regulatory Commission (NRC) staff has
reviewed the licensee's analysis and, based on this review, it appears
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: William F. Burton, Acting.
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 5, 2004.
Description of amendment request: The proposed amendments would
change Technical Specification (TS) 4.0.5.c. Specifically, the proposed
change would extend the examination frequency for the reactor coolant
pump (RCP) motor flywheel from a 10-year interval to an interval not to
exceed 20 years. This proposed change is consistent with the Industry/
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-421, ``Revision to RCP Flywheel
Inspection Program (WCAP-15666).''
Basis for proposed no significant hazards consideration
determination: As required by Title 10, Code of Federal Regulations,
Part 50, Section 91(a) (10 CFR 50.91(a)), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
TVA has reviewed the proposed no significant hazards
consideration determination published on June 24, 2003 (68 FR
37590), as part of the consolidated line item improvement process
(CLIIP). TVA has concluded that the proposed determination presented
in the notice is applicable to SQN, and the determination is hereby
incorporated by reference to satisfy the requirements of 10 CFR
50.91(a).
The United States Nuclear Regulatory Commission (NRC) staff has
reviewed the licensee's analysis and, based on this review, it appears
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: William F. Burton, Acting.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia
Date of amendment request: March 4, 2004.
Description of amendment request: The proposed amendments would
delete the note in Improved Technical Specification Surveillance
Requirement 3.4.12.7 that permitted the performance of the Channel
Operational Test within 12 hours of entering a mode in which the power-
operated relief valves (PORVs) are required to be operable for low
temperature overpressure protection (LTOP).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do changes involve a significant increase in the probability
or consequences of an accident previously evaluated?
The proposed changes to perform a Channel Operational Test on
each required PORV at least 31 days prior to entering the LTOP Mode
will continue to ensure verification and adjustment, if required, of
its lift setpoint. Changes will not affect the probability of
occurrence of any accident previously analyzed: nor alter the design
assumptions, conditions, and configuration of the facility or the
manner in which the plant is operated and maintained. Therefore, the
proposed changes do not involve a significant increase in the
consequences of any previously analyzed accident.
2. Do changes create the possibility of a new or different kind
of accident from any accident previously evaluated?
The proposed changes to perform a Channel Operational Test on
each required PORV at least 31 days prior to entering the LTOP Mode
will not create any new accident or event initiators. No systems,
structures, or components are being physically modified such that
the design function is being altered. The proposed changes do not
impose any new or different requirements for the performance of the
Channel Operational Test. Therefore, the proposed changes do not
create the possibility of a new or different kind of accident from
those previously analyzed.
3. Do changes involve a significant reduction in a margin of
safety?
The proposed changes do not involve any change to the safety
analysis limits. The level of safety of facility operation is
unaffected by the proposed changes since there is no change in the
intent for the performance of the Channel Operational Test.
Therefore, it is concluded that the margin of safety will not be
reduced by the implementation of the changes.
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: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendment: February 25, 2004.
Brief description of amendments: The amendment would extend the
implementation date for Amendment Nos. 261 and 238 for Calvert Cliffs
Units 1 and 2, respectively, to July 1, 2004. The changes to the
reactor pressure vessel pressure-temperature limits cooldown rates that
were approved by Amendment Nos. 261 and 238 are more conservative than
the plants existing rates and result in a longer cooldown period. The
existing cooldown rates are acceptable through the end of 2004.
Date of publication of individual notice in Federal Register: March
5, 2004 (69 FR 10487).
Expiration date of individual notice: May 5, 2004.
[[Page 19578]]
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: December 23, 2003, as
supplemented by letter dated January 30, 2004.
Brief description of amendment: The amendment modified Technical
Specification (TS) requirements for mode change limitations to adopt
the TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility in
Mode Restraints.''
Date of issuance: March 29, 2004.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 241.
Facility Operating License No. NPF-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 20, 2004 (69
FR 2738).
The January 30, 2004, letter provided clarifying information within
the scope of the original application and did not change the staff's
initial proposed no significant hazards consideration determination.
The staff's related evaluation of the amendment is contained in a
Safety Evaluation dated March 29, 2004.
No significant hazards consideration comments received: No.
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: October 7, 2003, and its
supplement dated December 18, 2003.
Brief description of amendments: The amendments revise Technical
Specification (TS) Section 5.5.6, ``Containment Tendon Surveillance
Program,'' for consistency with the requirements of 10 CFR 50.55a(g)(4)
for components classified as Code Class CC. The amendments also delete
the provisions of Surveillance Requirement (SR) 3.0.2 from this TS. In
addition, the amendments revise TS 5.5.16, ``Containment Leakage Rate
Testing Program,'' to add exceptions to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Testing Program.'' Also, the
paragraphs in Section 5.5.16 have been sequenced to more clearly
separate the requirements of the program. This is considered an
administrative change and is consistent with the guidance in NUREG-
1432, ``Standard Technical Specifications Combustion Engineering
Plants,'' Revision 2.
Date of issuance: March 19, 2004.
Effective date: March 19, 2004, and shall be implemented within 90
days of the date of issuance.
Amendment Nos.: Unit 1-151, Unit 2--151, Unit 3--151.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68659) The December 18, 2003, supplemental letter provided revised
technical specification pages to reflect changes that were approved in
Amendment No. 149, did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 19, 2004.
No significant hazards consideration comments received: No.
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: September 17, 2003 as
supplemented by letter dated February 20, 2004.
Brief description of amendments: The amendments revise the
technical specifications to support the replacement of part-length
control element assemblies (CEAs) with a new design, referred to as
part-strength CEAs. The two designs are geometrically very similar and
contain essentially the same amount and type of neutron absorber in the
lower half of the assemblies, which is the region of the CEAs inserted
into the reactor core during normal operations.
Date of issuance: March 23, 2004.
Effective date: March 23, 2004, and shall be implemented within 60
days of the date of issuance.
Amendment Nos.: Unit 1--152, Unit 2--152, Unit 3--152.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68657). The February 20, 2004, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination.
[[Page 19579]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 23, 2004.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-325, Brunswick Steam
Electric Plant, Unit 1, Brunswick County, North Carolina
Date of application for amendment: October 31, 2003, as
supplemented March 4, March 12, and March 19, 2004.
Brief description of amendment: The amendment revised the Minimum
Critical Power Ratio Safety Limit contained in Technical Specification
2.1.1.2.
Date of issuance: March 26, 2004.
Effective date: Effective as of the date of issuance and shall be
implemented prior to startup for Unit 1, Cycle 15, operation.
Amendment No.: 231.
Facility Operating License Nos. DPR-71: Amendment changes the
Technical Specifications.
Date of initial notice in Federal Register: January 6, 2004 (69 FR
693). The March 4, March 12, and March 19, 2004, supplemental letters
provided clarifying information that did not change the scope of the
proposed amendment as described in the original notice of proposed
action published in the Federal Register and did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 2004.
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: February 14, 2003, as
supplemented by letters dated November 10 and December 10, 2003, and
January 30, 2004.
Brief description of amendment: This amendment revises Technical
Specification (TS) 5.6.3.d to allow an increase in the decay heat load
from 1.0 MBTU/hr to 7.0 MBTU/hr for fuel stored in Spent Fuel Pools C
and D at Shearon Harris Nuclear Power Plant, Unit 1.
Date of issuance: March 26, 2004.
Effective date: March 26, 2004.
Amendment No.: 115.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12948). The November 10 and December 10, 2003, and January 30, 2004,
supplements provided clarifying information that did not change the
scope of the proposed amendment as described in the original notice of
proposed action published in the Federal Register and did not change
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 2004.
No significant hazards consideration comments received: No.
Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear
Plant, Charlevoix County, Michigan
Date of application for amendment: November 20, 2002, and August 6,
2003, as supplemented by letters dated December 1, 2003, and February
20, 2004.
Brief description of amendment: The amendment revises the Big Rock
Point License and Defueled Technical Specifications to remove reactor
operational and administrative requirements that are no longer
applicable due to the transfer of all spent fuel from the spent fuel
pool into dry cask storage at the Big Rock Point Independent Spent Fuel
Storage Installation.
Date of issuance: March 19, 2004.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 125.
Facility Operating License No. DPR-6: Amendment revises the
Defueled Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2800), and November 25, 2003 (68 FR 66133). The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
March 19, 2004.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: March 4, 2003, as supplemented
May 13 and September 18, 2003, and February 12 and March 10, 2004.
Brief description of amendment: The amendment revised selected
sections of the Technical Specifications (TSs) based upon a re-analysis
of fuel handling accidents (FHAs). The revised analysis is based upon
selective implementation of the alternative source term methodology of
Regulatory Guide 1.183, and in accordance with Title 10 of the Code of
Federal Regulations, Section 50.67. Specifically, the amendment
revised: TS 3.7.8, ``Plant Systems, Control Room Envelope
Pressurization System;'' TS 3.9.4, ``Refueling Operations, Containment
Building Penetrations;'' TS 3.9.9, ``Refueling Operations, Containment
Purge and Exhaust Isolation System,'' and TS 3.9.12, ``Refueling
Operations, Fuel Building Exhaust Filter System.''
Date of issuance: March 17, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 219.
Facility Operating License No. NPF-49: The amendment revised the
TSs.
Date of initial notice in Federal Register: March 4, 2003 (68 FR
40711). The May 13 and September 18, 2003, and February 12 and March
10, 2004, supplements contained clarifying information and did not
change the staff's initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: December 5, 2003, as
supplemented on February 9, 2004.
Brief description of amendment: The amendment revised the Safety
Limit Minimum Critical Power Ratio values in Technical Specification
1.1.A.1 to incorporate the results of the cycle-specific core reload
analysis for Vermont Yankee Nuclear Power Station Cycle 24 operation.
Date of Issuance: March 22, 2004.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 217.
Facility Operating License No. DPR-28: The amendment revised the
TSs.
Date of initial notice in Federal Register: January 20, 2004 (69 FR
2741). The supplement dated February 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.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated March 22, 2004.
[[Page 19580]]
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: March 26, 2003, as supplemented
on July 24, 2003.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) regarding reactor pressure vessel (RPV) fracture
toughness and material surveillance requirements (SRs). Specifically,
the amendment revised the pressure-temperature limits for the RPV as
specified in TS Figures 3.6.1, 3.6.2, and 3.6.3. In addition, the
amendment deleted TS 4.6.A.5, which specifies plant-specific RPV
material SRs. These plant-specific SRs are being replaced by
implementing the Boiling Water Reactor Vessel and Internals Project
(BWRVIP) RPV integrated surveillance program (ISP). The details of the
BWRVIP ISP will be added to the Vermont Yankee Nuclear Power Station
Updated Final Safety Analysis Report.
Date of Issuance: March 29, 2004.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 218.
Facility Operating License No. DPR-28: Amendment revised the TSs.
Date of initial notice in Federal Register: April 29, 2003 (68 FR
22747). The supplement dated July 24, 2003, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated March 29, 2004.
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: June 11, 2003, as supplemented
on December 5, December 30, 2003, and February 18, 2004.
Brief description of amendments: The amendments revise technical
specification 3.7.8 to permit a one-time extension from 72 hours to 144
hours for the completion time required to restore a unit specific
essential service water train to operable status.
Date of issuance: March 18, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 136/136, 130/130.
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: September 30, 2003.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 18, 2004.
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: September 8, 2003.
Brief description of amendments: The amendments modified Technical
Specifications requirements to adopt the provisions of Industry/
Technical Specification Task Force (TSTF) change 359, ``Increase
Flexibility in Mode Restraints.''
Date of issuance: March 12, 2004.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 169 and 132.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68668).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 12, 2004.
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: April 15, 2002, as supplemented by
letter dated January 14, 2004.
Description of amendment request: The amendment revises the
Technical Specifications (TSs) to relocate the boron concentration
limits and ``Safety Limits'' figures to the Core Operating Limits
Report. Some limiting conditions and actions are revised to be
consistent with the Improved Standard Technical Specifications.
Date of issuance: March 23, 2004.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 96.
Facility Operating License No. NPF-86: The amendment revises the
TS.
Date of initial notice in Federal Register: May 28, 2002 (67 FR
36931). The January 14, 2004, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the amendment beyond the scope of
the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 23, 2004.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 25, 2003.
Brief description of amendment: The amendment revises the Technical
Specification (TS) for Limiting Condition for Operation requirement
3.5.1 to incorporate TS Task Force Traveler 318 to allow one low
pressure coolant injection pump inoperable in each of the two emergency
core cooling system divisions.
Date of issuance: March 31, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 203.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 14, 2003 (68 FR
59218).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: March 27, 2003, as supplemented
on November 3, 2003, and January 28, 2004.
Brief description of amendments: The amendment revises Technical
Specification Surveillance Requirement 3.2.4.2, ``Rod Group Alignment
Limits.'' The revision expands the alignment limits on allowable rod
cluster control assembly, or rod, deviation from demanded position. The
change applies in Mode 1, when operating at greater than 85 percent of
rated thermal power.
Date of issuance: March 29, 2004.
[[Page 19581]]
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 212 and 217.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 29, 2003 (68 FR
22749).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 29, 2004.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: June 11, 2003.
Brief description of amendments: The amendments revise the
technical specifications to allow use of the power distribution
monitoring system (PDMS) for power distribution measurements as
described in Topical Report WCAP-12462-P-A, ``BEACON: Core Monitoring
and Support System.''
Date of issuance: March 31, 2004.
Effective date: March 31, 2004, and shall be implemented within 180
days from the date of issuance.
Amendment Nos.: Unit 1--164; Unit 2--166.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40717).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: September 18, 2003, as
supplemented December 8, 2003, and February 24, 2004.
Description of amendment request: The amendments revised the
pressure-temperature limit curves in Technical Specification (TS)
3.4.9.
Date of issuance: March 10, 2004.
Effective date: March 10, 2004.
Amendment Nos.: 288 & 247.
Facility Operating License No. DPR-52 and DPR-68: Amendments
revised the TSs.
Date of initial notice in Federal Register: October 28, 2003 (68 FR
61480). The December 8, 2003, and February 24, 2004, letters provided
clarifying information that did not change the scope of the original
request or the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 10, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: March 24, 2003, as supplemented
December 4, 2003, and February 12, 2004.
Brief description of amendment: The amendment revises the design
and licensing basis failure modes and effects analysis for specific
valves in the essential raw cooling water system, component cooling
water system, and control air system to address a condition in which
containment integrity, accident flood levels, and sump boron
concentrations subsequent to a high-energy line break could not be
automatically ensured, and, therefore, manual actions are required.
Date of issuance: March 29, 2004.
Effective date: As of the date of issuance and shall be implemented
in conjunction with the next update to the Updated Final Safety
Analysis Report required by 10 CFR 50.71(e).
Amendment No.: 51.
Facility Operating License No. NPF-90: Amendment revises the
Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18287). The supplemental letters provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 29, 2004.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket No. 50-445, Comanche Peak Steam
Electric Station, Unit No. 1, Somervell County, Texas
Date of amendment request: July 21, 2003, as supplemented by
letters dated January 8, January 21, and March 8, 2004.
Brief description of amendments: The Amendment revises the
Technical Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance
Program,'' to allow the use of Westinghouse (Westinghouse Electric
Station LLC) leak limiting Alloy 800 sleeves for repair of degraded SG
tubes.
Date of issuance: March 24, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 112.
Facility Operating License No. NPF-87: The amendments revised the
Technical Specifications.
Date of initial notice in Federal Register: July 21, 2003.
Supplemental letters dated January 8, January 21, and March 8, 2004
provided clarifying information that did not change the scope of the
original Federal Register notice or the original no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 24, 2004.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: December 8, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 5.5.6, ``Containment Tendon Surveillance
Program,'' for consistency with the requirements of 10 CFR 50.55a(g)(4)
for components classified as Code Class CC. The amendment also deletes
the provisions of Surveillance Requirement (SR) 3.0.2 from this TS. In
addition, the amendment revises TS 5.5.16, ``Containment Leakage Rate
Testing Program,'' to add exceptions to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Testing Program.''
Date of issuance: March 17, 2004.
Effective date: March 17, 2004, and shall be implemented within 90
days from the date of issuance.
Amendment No.: 160.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 6, 2004 (69 FR
700).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 2004.
No significant hazards consideration comments received: No.
[[Page 19582]]
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 17, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 5.5.6, ``Containment Tendon Surveillance
Program,'' for consistency with the requirements of 10 CFR 50.55a(g)(4)
for components classified as Code Class CC. The amendment also deletes
the provisions of Surveillance Requirement (SR) 3.0.2 from this TS. In
addition, the amendment revises TS 5.5.16, ``Containment Leakage Rate
Testing Program,'' to add exceptions to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Testing Program.''
Date of issuance: March 17, 2004.
Effective date: March 17, 2004, and shall be implemented within 90
days from the date of issuance.
Amendment No.: 152.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2003 (68
FR 64140).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 5th day of April 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-8047 Filed 4-12-04; 8:45 am]
BILLING CODE 7590-01-P