[Federal Register Volume 69, Number 51 (Tuesday, March 16, 2004)]
[Notices]
[Pages 12361-12376]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-5596]
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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, January 20, 2004, through March 4, 2004.
The last biweekly notice was published on March 2, 2004 (69 FR 9857).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the
[[Page 12362]]
Commission make a final No Significant Hazards Consideration
Determination, any hearing will take place after issuance. The
Commission expects that the need to take this action will occur very
infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to 301-415-3725 or by email 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
[[Page 12363]]
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 email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: November 11, 2003.
Description of amendment request: The proposed amendment would
amend Appendix A, Technical Specifications (TS), of Facility Operating
License No. NPF-62 for Clinton Power Station (CPS). The proposed
changes would revise several CPS TS instrument channel trip setpoint
Allowable Values.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment implements revised Allowable Values for
the following instrument functions.
Main Steam Isolation Valve--Closure
Anticipated Transient Without Scram Recirculation
Pump Trip Reactor Steam Dome Pressure--High
Reactor Vessel Pressure--Low (Injection
Permissive)
Reactor Vessel Water Level--Low Low Low, Level 1
Reactor Vessel Water Level--Low Low, Level 2
High Pressure Core Spray (HPCS) System Reactor
Vessel Water Level--High, Level 8
Reactor Core Isolation Cooling (RCIC) Storage
Tank Level--Low
HPCS System Suppression Pool Water Level--High
(Pump Suction Transfer)
Automatic Depressurization System (ADS)
Initiation Permissive, Low Pressure Core Spray (LPCS) Pump Discharge
Pressure--High
ADS Initiation Permissive, Low Pressure Coolant
Injection (LPCI) Pumps Discharge Pressure--High
RCIC System Suppression Pool Water Level--High
(Pump Suction Transfer)
Main Steam Line Pressure--Low, and
Safety Relief Valve (SRV) Relief and Low-Low Set
(LLS) functions channel calibration surveillance requirement
The proposed changes do not require modification to the
facility. There is no impact on the accident analysis as a result of
the proposed changes to the Allowable Values. The analytical limit,
which is used as input to the accident analysis, does not change.
The proposed changes will be implemented through revision of the
associated surveillance test procedures, where the revised Allowable
Value will replace the existing value.
Derivation of the Allowable Value in accordance with Regulatory
Guide 1.105, ``Instrument Setpoints,'' uses the analytical limit as
a fixed starting point from which instrument uncertainties are added
or subtracted, as appropriate. Calculation of the Allowable Value to
plant-specific parameters provides additional confidence that
protective instrumentation that passes the surveillance testing
criteria will perform its design function without exceeding the
associated safety analysis limit.
The revised Allowable Values for the affected equipment are not
considered an initiator to any previously analyzed accident and
therefore, cannot increase the probability of any previously
evaluated accident. Implementation of the revised Allowable Values
will ensure that the instrumentation will perform its required
function to meet the accident analysis assumptions. The proposed
Allowable Values will ensure that the fuel is adequately cooled,
containment and drywell are isolated as required, primary
containment temperature and pressure design limits are met, and
overpressurization of the nuclear steam supply system is prevented
following an accident or transient. The proposed changes do not
increase the probability of any accident previously evaluated.
Since the proposed changes ensure the same level of protection
as assumed in the accident analyses, the conclusions of the accident
scenarios remain valid. As a result, no changes to radiological
release parameters are involved. Therefore, the proposed changes do
not increase the consequences of an accident previously evaluated.
In summary, the proposed changes do 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?
The proposed changes do not affect the design, functional
performance or operation of the facility. Similarly, they do not
affect the design or operation of any structures, systems, or
components involved in the mitigation of any accidents, nor do they
affect the design or operation of any component in the facility such
that new equipment failure modes are created. Setpoints remain the
same and therefore, there is no impact on the operation of any of
the associated systems.
As such the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed changes do not involve a change to the plant design
or operation. The proposed changes will be implemented through
revisions to the associated surveillance test procedures where the
revised Allowable Value replaces the existing Allowable Value. No
changes to the instrument setpoints are involved. Since the
availability of the systems will be maintained and since the system
designs are unaffected, the proposed changes ensure the
instrumentation is capable of performing their intended functions.
The proposed changes do not affect the accident analyses that assume
the operability of the instrumentation associated with these
Allowable Values. The margins associated with the analytical limits
are not impacted by the proposed Allowable Values since the
analytical limits remain unchanged.
Therefore, operation of CPS in accordance with the proposed
changes will 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: Edward J. Cullen, Deputy General Counsel
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: August 6, 2003, as supplemented on
February 13, 2004.
Description of amendment request: This amendment would revise the
Technical Specifications (TSs) to incorporate reference to the 10 CFR
50.55a, Codes and Standards, in lieu of the existing criteria of
Regulatory Guide 1.35.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed revision to Technical Specification 4.4.2.1 and
associated Bases Section incorporates reference to the criteria of
10 CFR 50.55a, ``Codes and standards,'' in lieu of the existing
criteria of Regulatory Guide 1.35. This change provides consistency
between the Technical Specification tendon surveillance program
criteria and the regulatory requirements specified in 10 CFR
50.55a(b)(2)(vi). These regulatory requirements and the associated
surveillance program ensure that the reactor building tendon
prestressing system is capable of maintaining the structural
integrity of the containment during operating
[[Page 12364]]
and accident conditions. The reactor building prestressing system is
not an initiator of any accident. Therefore, this change is not
related to the probability of any accident previously evaluated.
This change ensures that the containment tendon surveillance program
addresses the appropriate regulatory criteria. This change does not
result in any reduction in the effectiveness of the existing
surveillance program. The tendon surveillance program will continue
to ensure that the containment structure is capable of performing
its intended safety function in the event of a design basis
accident. Therefore, this change has no affect on the consequences
of an accident previously evaluated.
The proposed changes to Technical Specification Definition 1.22,
Technical Specification 3.1.6.6 and associated Bases, and Technical
Specification 3.24 Bases are only administrative changes or
corrections and have no affect on plant design or operations.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed revision to Technical Specification 4.4.2.1 and
associated Bases Section incorporates reference to the criteria of
10 CFR 50.55a, ``Codes and standards,'' in lieu of the existing
criteria of Regulatory Guide 1.35. This change provides consistency
between the Technical Specification tendon surveillance program
criteria and the regulatory requirement specified in 10 CFR
50.55a(b)(2)(vi). The proposed Technical Specification change does
not result in any reduction in effectiveness of the existing tendon
surveillance program. The tendon surveillance program will continue
to satisfy the applicable Technical Specification and regulatory
required criteria, thus ensuring that the containment structure will
perform its design safety function. This change has no affect on the
design and operation of plant structures, systems, and components.
This change does not introduce any new accident precursors and does
not involve any alterations to plant configurations, which could
initiate a new or different kind of accident.
The proposed changes to Technical Specification Definition 1.22,
Technical Specification 3.1.6.6 and associated Bases, and Technical
Specification 3.24 Bases are only administrative changes or
corrections and have no affect on plant design or operations.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed revision to Technical Specification 4.4.2.1 and
associated Bases Section incorporates reference to the criteria of
10 CFR 50.55a, ``Codes and standards,'' in lieu of the existing
criteria of Regulatory Guide 1.35. The change provides consistency
between the Technical Specification tendon surveillance program
criteria and the regulatory requirement specified in 10 CFR
50.55a(b)(2)(vi). The containment examination and inspection
requirements specified in 10 CFR 50.55a(b)(2)(vi) meet the same
standards as the criteria specified in Regulatory Guide 1.35. The
proposed Technical Specification change does not result in any
reduction in effectiveness of the existing tendon surveillance
program. The tendon surveillance program will continue to satisfy
the applicable Technical Specification and regulatory required
criteria, thus ensuring that the containment structure will perform
its design safety function in accordance with existing margins of
safety for containment integrity.
The proposed changes to Technical Specification Definition 1.22,
Technical Specification 3.1.6.6 and associated Bases, and Technical
Specification 3.24 Bases are only administrative changes or
corrections and have no affect on plant design or operations.
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: 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.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: December 15, 2003.
Description of amendments request: The proposed amendment would
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV)
Vent and Drain Valves,'' to allow a vent or drain line with one
inoperable valve to be isolated instead of requiring the valve to be
restored to Operable status within 7 days.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on February 24, 2003 (68 FR 8637), on possible
amendments to revise the action for one or more SDV vent or drain lines
with an inoperable valve, including a model safety evaluation and model
no significant hazards consideration (NSHC) determination, using the
consolidated line-item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on April 15,
2003 (68 FR 18294). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 15, 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.
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with one valve inoperable instead of requiring the valve to be
restored to operable status within 7 days. With one SDV vent or
drain valve inoperable in one or more lines, the isolation function
would be maintained since the redundant valve in the affected line
would perform its safety function of isolating the SDV. Following
the completion of the required action, the isolation function is
fulfilled since the associated line is isolated. The ability to vent
and drain the SDVs is maintained and controlled through
administrative controls. This requirement assures the reactor
protection system is not adversely affected by the inoperable
valves. With the safety functions of the valves being maintained,
the probability or consequences of an accident previously evaluated
are not significantly increased.
Criterion 2--The proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in the margin of safety.
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDVs is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of an SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin of safety.
[[Page 12365]]
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: William Burton, Acting.
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 amendment request: February 4, 2004.
Description of amendment request: The proposed amendment would
revise the Technical Specifications Index and Technical Specifications
(TS) 4.4.1.3.2, ``Reactor Coolant System Hot Shutdown Surveillance
Requirements,'' and 3.4.1.4.1.b, ``Reactor Coolant System Cold
Shutdown--Loops Filled Limiting Condition For Operation.'' The proposed
change to the Index is an administrative update to restore consistency
with other sections of the TS. The proposed change to TS 4.4.1.3.2 and
TS 3.4.1.4.1.b eliminates a requirement that the wide-range
instrumentation be inoperable before the narrow-range instrumentation
can be used for confirmation of the minimum steam generator secondary
side water level. The primary reason for this proposed change to TS
4.4.1.3.2 and TS 3.4.1.4.1.b is to provide the operational flexibility
needed for a smooth transition through the applicable range of
operating conditions.
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.
There is no impact on previously evaluated accidents because the
proposed amendment does not affect the capability of any structure,
system, or component to perform its design function. The functional
capability of the narrow range instrumentation is not impacted by
the operability status of the wide range instrumentation. The
existing minimum values specified by Technical Specifications for
the wide range and the narrow range instrumentation conservatively
incorporate the applicable uncertainties necessary to make either
instrument suitable for use over the expected range of operating
conditions. As a result, the proposed amendment does not affect the
operating procedures and administrative controls that have the
function of preventing or mitigating any [previously] evaluated
accident.
Therefore, the proposed change 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 proposed amendment does not change the design function or
operation of any structure, system, or component. The proposed
amendment does not involve any physical change to plant equipment.
Use of the narrow range instrumentation while the wide range
instrumentation is operable does not create any new or different
failure mechanisms, malfunctions, or accident initiators than those
already considered in the design and licensing bases.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment does not affect the margin of safety
because the existing minimum values specified by Technical
Specifications for the wide range and the narrow range
instrumentation are not changed. Those minimum values conservatively
incorporate the applicable uncertainties necessary to make either
instrument suitable for use over the expected range of operating
conditions. The calculation of those uncertainties for use of the
narrow range instrumentation is unaffected by the operating status
of the wide range instrumentation.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, [Carolina Power & Light Company] concludes
that the proposed amendment involves no significant hazards
consideration under the standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ``no significant hazards consideration''
is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen Howe.
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 3, 2003.
Description of amendment request: Pursuant to Title 10 of the Code
of Federal Regulations, Section 50.90, Duke Energy Corporation
requested an amendment to the McGuire Nuclear Station Facility
Operating Licenses and Technical Specifications. The proposed change
would add a note to Limiting Condition of Operation 3.7.11, ``Auxiliary
Building Filtered Ventilation Exhaust System (ABFVES)'', that would
allow the Auxiliary Building pressure boundary to be opened
intermittently under administrative control. Changes to the
corresponding Bases would also be made to establish the administrative
controls that are required to minimize the consequences of the open
pressure boundary.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No, the Auxiliary Building Filtered Ventilation Exhaust System
(ABFVES) is not assumed to be an initiator of any analyzed accident.
Therefore, the proposed change contained in this license amendment
request has no significant impact on the probability of occurrence
of any previously analyzed accident.
The ABFVES provides a means of filtering air from the area of
the active emergency core cooling system (ECCS) components, thereby
providing environmental control for temperature and humidity in the
ECCS pump room area and the Auxiliary Building. During emergency
operations, the ABFVES exhausts air from the mechanical penetration
area and the ECCS pump room area and discharges it through the
system filters. For cases where the Auxiliary Building pressure
boundary is opened intermittently under administrative controls,
appropriate compensatory measures would be required by the proposed
Technical Specification to ensure the pressure boundary can be
rapidly restored. Based on the compensatory measures available to
the plant operators and the administrative controls required to
rapidly restore an opened pressure boundary, the accident
consequences do not cause a significant increase in dose above the
applicable General Design Criter[i]a, Standard Review Plan, or 10
CFR [Part] 100 limits.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No, there are no changes being made to actual plant hardware
which will result in
[[Page 12366]]
any new accident causal mechanisms. Also, no changes are being made
to the way in which the plant is being operated. Therefore, no new
accident causal mechanisms will be generated.
3. Does this change involve a significant reduction in a margin
of safety?
No, margin of safety is related to the ability of the fission
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 significantly degraded
by the proposed changes. When the Auxiliary Building pressure
boundary is open on an intermittent basis, as permitted by the
changes proposed in this license amendment request, administrative
controls would be in place to ensure that the integrity of the
pressure boundary could be rapidly restored. Therefore, it is
expected that the plant, and the operating personnel, would maintain
the ability to mitigate design basis events, and that none of the
fission product barriers would be significantly affected by this
change. Therefore, the proposed change is not considered to result
in 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: 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, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: October 15, 2003.
Description of amendment request: The amendments would add a new
Technical Specification (TS) 3.9.7, ``Unborated Water Source isolation
Valves,'' and would revise TS 3.9.2, ``Nuclear Instrumentation,'' to
delete the requirement for Boron Dilution Mitigation System automatic
valve actuations and makeup water pump trip during Mode 6 and to agree
with the wording of NUREG-1431, ``Standard Technical Specifications
Westinghouse Plants,'' Revision 2. The licensee proposed these changes
to provide configuration control of the dilution valves during Mode 6
to preclude the possibility of a boron dilution event and to provide an
opportunity to conduct maintenance on the volume control tank valves,
refueling water storage tank valves, and their respective power
supplies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Operation of the facilities in accordance with this amendment
would not involve a significant increase in the probability or
consequences of an accident previously evaluated. The BDMS [Boron
Dilution Mitigation System] system is designed to mitigate the
consequences of an inadvertent boron dilution event. The probability
of the dilution accident will be reduced by administratively
isolating potential dilution flow paths. Thus, with the proposed
changes, boron dilution is not considered a credible accident during
refueling.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Operation of the facilities in accordance with this amendment
would not create the possibility of a new or different kind of
accident from any accident previously evaluated. No new accident
causal mechanisms are created as a result of this proposed
amendment. No changes are being made to any structure, system, or
component which will introduce any new accident causal mechanisms.
This amendment request does not impact any plant systems that are
accident initiators and does not impact any safety analysis.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Operation of the facilities in accordance with this amendment
would not involve a significant reduction in a margin of safety. The
design criterion and margin of safety for the current BDMS is that
the dilution event is terminated prior to the loss of all shutdown
margin. The same criterion will be met following the isolation of
dilution valves. Therefore, there is no 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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: February 18, 2004.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (TS) to maintain
hydrogen recombiners and hydrogen and oxygen monitors. Licensees were
generally required to implement upgrades as described in NUREG-0737,
``Clarification of TMI [Three Mile Island Nuclear Station] Action Plan
Requirements,'' and Regulatory Guide (RG) 1.97, ``Instrumentation for
Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs
Conditions During and Following an Accident.'' Implementation of these
upgrades was an outcome of the lessons learned from the accident that
occurred at TMI, Unit 2. Requirements related to combustible gas
control were imposed by Order for many facilities and were added to or
included in the TSs for nuclear power reactors currently licensed to
operate. The revised 10 CFR 50.44, ``Standards for Combustible Gas
Control System in Light-Water-Cooled Power Reactors,'' eliminated the
requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design and
qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on September
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the
model NSHC determination in its application dated February 18, 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 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
[[Page 12367]]
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 and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
RG 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 and oxygen 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. Also, as part of the rulemaking to
revise 10 CFR 50.44, the Commission found that Category 2, as
defined in RG 1.97, is an appropriate categorization for the oxygen
monitors, because the monitors are required to verify the status of
the inert containment.
The regulatory requirements for the hydrogen and oxygen 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,
classification of the oxygen monitors as Category 2, and removal of
the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy 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 the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen 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 and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen 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 and oxygen 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.
Category 2 oxygen monitors are adequate to verify the status of
an inerted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: January 15, 2004.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) Section 5.5.12, ``Primary
Containment Leakage Rate Testing Program,'' to reflect a one-time
deferral of the primary containment Type A test to no later than
February 27, 2011, for Unit 2, and no later than July 13, 2009, for
Unit 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change will revise Dresden Nuclear Power Station
(DNPS) Units 2 and 3 Technical Specifications (TS) Section 5.5.12,
``Primary Containment Leakage Rate Testing Program,'' to reflect a
one-time deferral of the primary containment Type A test to no later
than February 27, 2011, for Unit 2, and no later than July 13, 2009,
for Unit 3. The current Type A test interval of 10 years, based on
past performance, would be extended on a one-time basis to 15 years
from the last Type A test.
The function of the primary containment is to isolate and
contain fission products released from the reactor coolant system
(RCS) following a design basis loss-of-coolant accident (LOCA) and
to confine the postulated release of radioactive material to within
limits. The test interval associated with Type A testing is not a
precursor of any accident previously evaluated. Therefore, extending
this test interval on a one-time basis from 10 years to 15 years
does not result in an increase in the probability of occurrence of
an accident. The successful performance history of Type A testing
provides assurance that the DNPS primary containments will not
exceed allowable leakage rate values specified in the TS and will
continue to perform their design function following an accident. The
risk assessment of the proposed change has concluded that there is
an insignificant increase in total population dose rate and an
insignificant increase in the conditional containment failure
probability.
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?
The proposed change for a one-time extension of the Type A tests
for DNPS Units 2 and 3 will not affect the control parameters
[[Page 12368]]
governing unit operation or the response of plant equipment to
transient and accident conditions. The proposed change does not
introduce any new equipment or modes of system operation. No
installed equipment will be operated in a new or different manner.
As such, no new failure mechanisms are introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
DNPS Units 2 and 3 are General Electric BWR/3 plants with Mark I
primary containments. The Mark I primary containment consists of a
drywell, which encloses the reactor vessel, reactor coolant
recirculation system, and branch lines of the RCS; a toroidal-shaped
pressure suppression chamber containing a large volume of water; and
a vent system connecting the drywell to the water space of the
suppression chamber. The primary containment is penetrated by
access, piping, and electrical penetrations.
The integrity of the primary containment penetrations and
isolation valves is verified through Type B and Type C local leak
rate tests (LLRTs) and the overall leak-tight integrity of the
primary containment is verified by a Type A integrated leak rate
test (ILRT) as required by 10 CFR 50, Appendix J, ``Primary Reactor
Containment Leakage Testing for Water-Cooled Power Reactors.'' The
tests are performed to verify the essentially leak-tight
characteristics of the primary containment at the design basis
accident pressure. The proposed change for a one-time extension of
the Type A tests do not affect the method for Type A, B, or C
testing, or the test acceptance criteria. In addition, based on
previous Type A testing results, EGC does not expect additional
degradation, during the extended period between Type A tests, which
would result in a significant reduction in a margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendment request: January 15, 2004.
Description of amendment request: Modify Technical Specification
Surveillance Requirement 3.4.3.2, SR 3.5.1.10, and SR 3.6.1.6.1 to
provide an alternative means for testing the main steam Electromatic
relief valves and the dual function Target Rock safety/relief valves.
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 changes modify Technical Specifications (TS)
Surveillance Requirement (SR) 3.4.3.2, SR 3.5.1.10, and SR 3.6.1.6.1
to provide an alternative means for testing the main steam line
relief valves, automatic depressurization system valves, and low set
relief valves. Accidents are initiated by the malfunction of plant
equipment, or the catastrophic failure of plant structures, systems
or components. The performance of relief valve testing is not a
precursor to any accident previously evaluated and does not change
the manner in which the valves are operated. The proposed testing
requirements will not contribute to the failure of the relief valves
nor any plant structure, system or component. Exelon Generation
Company, LLC has determined that the proposed change in testing
methodology provides an equivalent level of assurance that the
relief valves are capable of performing their intended safety
functions. Thus, the proposed changes do not affect the probability
of an accident previously evaluated.
The performance of relief valve testing provides confidence that
the relief valves are capable of depressurizing the reactor pressure
vessel (RPV). This will protect the reactor vessel from
overpressurization and allow the combination of the Low Pressure
Coolant Injection and Core Spray systems to inject into the RPV as
designed. The low set relief logic causes two low set relief valves
to be opened at a lower pressure than the relief mode pressure
setpoints and causes the low set relief valves to stay open longer,
such that reopening of more than one valve is prevented on
subsequent actuations. Thus, the low set relief function prevents
excessive short duration relief valve cycles with valve actuation at
the relief setpoint, which limits induced thrust loads on the relief
valve discharge line for subsequent actuations of the relief valve.
The proposed changes do not affect any function related to the
safety mode of the duel function safety/relief valves. The proposed
changes involve the manner in which the subject valves are tested,
and have no effect on the types or amounts of radiation released or
the predicted offsite does in the events of an accident. The
proposed testing requirements are sufficient to provide confidence
that the relief valves are capable of performing their intended
safety functions. In addition, a stuck open relief valve accident is
analyzed in the Updated Final Safety Analysis Report. Since the
proposed testing requirements do not alter the assumptions for the
stuck open relief valve accident, the radiological consequences of
any accident previously evaluated 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?
The proposed changes do not affect the assumed accident
performance of the main steam relief valves, nor any plant
structure, system, or component previously evaluated. The proposed
changes do not install any new equipment, and installed equipment is
not being operated in a new or different manner. The proposed change
in test methodology will ensure that the valves remain capable of
preforming their safety functions due to meeting the testing
requirements of the American Society of Mechanical Engineers Boiler
and Pressure Vessel Code, with the exception of opening the valve
following installation or maintenance for which a relief request has
been submitted, proposing an acceptable alternative. No setpoints
are being changed which would alter the dynamic response of plant
equipment. Accordingly, no new failure modes are introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed changes will allow testing of the valve actuation
electrical circuitry, including the solenoid, and mechanical
actuation components, without causing the relief valve to open. The
relief valves will be manually actuated prior to installation in the
plant. Therefore, all modes of relief valve operation will be tested
prior to entering the mode of operation requiring the valve to
perform their safety functions. The proposed changes do not affect
the valve setpoint or the operational criteria that directs the
relief valves to be manually opened during plants transients. There
are no changes proposed which alter the setpoints at which
protective actions are initiated, and there is no change to the
operability requirements for equipment assumed to operate for
accident mitigation.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Vice President,
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
Kennett Square, PA 19348.
[[Page 12369]]
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania
Date of amendment request: January 27, 2004.
Description of amendment request: The proposed change would revise
Technical Specification 3.4.5 to allow repair of steam generator tubes
by installation of leak limiting Alloy 800 sleeves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The leak limiting Alloy 800 sleeves are designed using the
applicable American Society for Mechanical Engineers (ASME) Boiler
and Pressure Vessel Code [ASME Code] and, therefore, meet the design
objectives of the original steam generator (SG) tubing. The applied
stresses and fatigue usage for the sleeves are bounded by the limits
established in the ASME Code. Mechanical testing has shown that the
structural strength of sleeves under normal, upset, emergency, and
faulted conditions provides margin to the acceptance limits. These
acceptance limits bound the most limiting (three times normal
operating pressure differential) burst margin recommended by NRC
Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR Steam
Generator Tubes.'' Burst testing of sleeve-tube assemblies has
confirmed the analytical results and demonstrated that no
unacceptable levels of primary-to-secondary leakage are expected
during any plant condition.
The leak limiting Alloy 800 sleeve depth-based structural limit
is determined using NRC guidance and the pressure stress equation of
ASME Code, Section III with additional margin added to account for
the configuration of long axial cracks. An Alloy 800 sleeved tube
will be plugged on detection of an imperfection in the sleeve or in
the pressure boundary portion of the original tube wall in the leak
limiting sleeve/tube assembly.
Evaluation of the repaired SG tube testing and analysis
indicates no detrimental effects on the leak limiting Alloy 800
sleeve or sleeved tube assembly from reactor system flow, primary or
secondary coolant chemistries, thermal conditions or transients, or
pressure conditions as may be experienced at Beaver Valley Power
Station (BVPS) Unit [No.] 1. Corrosion testing and historical
performance of sleeve-tube assemblies indicates no evidence of
sleeve or tube corrosion considered detrimental under anticipated
service conditions.
The implementation of the proposed change has no significant
effect on either the configuration of the plant or the manner in
which it is operated. The consequences of a hypothetical failure of
the leak limiting Alloy 800 sleeve-tube assembly is bounded by the
current SG tube rupture (SGTR) analysis described in the BVPS Unit
No. 1 Updated Final Safety Analysis Report. Due to the slight
reduction in the inside diameter caused by the sleeve wall
thickness, primary coolant release rates through the parent tube
would be slightly less than assumed for the SGTR analysis and
therefore, would result in lower total primary fluid mass release to
the secondary system. A main steam line break or feedwater line
break will not cause a SGTR since the sleeves are analyzed for a
maximum accident differential pressure greater than that predicted
in the BVPS Unit No. 1 safety analysis. The sleeve-tube assembly
leakage during plant operation would be minimal and is well within
the allowable Technical Specification leakage limits.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The leak limiting Alloy 800 sleeves are designed using the
applicable ASME Code as guidance, and therefore meet the objectives
of the original SG tubing. As a result, the functions of the SG will
not be significantly affected by the installation of the proposed
sleeve. The proposed sleeves do not interact with any other plant
systems. Any accident as a result of potential tube or sleeve
degradation in the repaired portion of the tube is bounded by the
existing SGTR accident analysis. The continued integrity of the
installed sleeve-tube assembly is periodically verified by Technical
Specification requirements and a sleeved tube will be plugged on
detection of an imperfection in the sleeve or in the pressure
boundary portion of the tube wall in the leak limiting sleeve/tube
assembly.
Implementation of the proposed change has no significant effect
on either the configuration of the plant, or the manner in which it
is operated.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The repair of degraded SG tubes with leak limiting Alloy 800
sleeves restores the structural integrity of the degraded tube under
normal operating and postulated accident conditions. The reduction
in core cooling margin due to the addition of Alloy 800 sleeves is
not significant because the cumulative effect of all repaired
(sleeved) and plugged tubes will continue to be less than the
currently allowed core cooling margin threshold established by the
total steam generator tube plugging level. The design safety factors
utilized for the sleeves are consistent with the safety factors in
the ASME Boiler and Pressure Vessel Code used in the original SG
design. The sleeve and portions of the installed sleeve-tube
assembly that represent the reactor coolant pressure boundary will
be monitored and a sleeved tube will be plugged on detection of an
imperfection in the sleeve or in the pressure boundary portion of
the original tube wall in the leak limiting sleeve/tube assembly.
Use of the previously identified design criteria and design
verification testing assures that the margin to safety is not
significantly different from the original SG tubes.
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: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
FirstEnergy Nuclear Operating Company (FENOC), et al., Docket Nos. 50-
334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1
and 2), Beaver County, Pennsylvania
Date of amendment request: January 26, 2004.
Description of amendment request: The proposed change would revise
the BVPS-1 and 2 Updated Final Safety Analysis Report (UFSAR)
description of the design-basis bounding limitations for the ultimate
heat sink design. The proposed change would allow the design
descriptions in the BVPS-1 and 2 UFSARs to credit the current Technical
Specification (TS) 3.7.5.1 requirement at each unit to shut down when
the Ohio River level reaches a low level below 654 feet mean sea level
(msl). This UFSAR revision would preclude design consideration for
design-basis accidents associated with power operation from occurring
when the Ohio River level is below 654 feet msl since the units would
be required to be shut down.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change will revise the BVPS Unit No. 1 and Unit
No. 2 UFSAR
[[Page 12370]]
description of the design basis bounding limitations for the
ultimate heat sink design. FENOC's proposed change will allow the
design description in each BVPS Unit's UFSAR to credit the current
[TS] 3.7.5.1 requirement at each BVPS Unit to shutdown when the Ohio
River level reaches a low level below 654 feet Mean Sea Level (msl).
This UFSAR revision will, therefore, preclude design consideration
for design bases accidents associated with power operation from
occurring when the Ohio River level is below 654' msl since the
plant will already be shutdown. This LAR [license amendment request]
does not propose any Technical Specification changes nor any
physical plant changes.
Since no physical plant changes nor any instrument setpoint
changes are being requested, it [the proposed change] would not
result in an increase in [the] probability of an accident previously
evaluated. Since the proposed change only clarifies the limiting
design basis ultimate heat sink scenario, consistent with both
Units' original licensing bases, it would not result in a
significant increase in the consequences of an accident previously
evaluated.
In conclusion, the request to amend the UFSARs for BVPS Unit
Nos. 1 and 2 to clarify the limiting design basis ultimate heat sink
scenario, consistent with both Units' original licensing bases, 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 previously evaluated?
No. The proposed changes only clarif[y] the limiting design
basis ultimate heat sink scenario, consistent with both Units'
original licensing bases. Since this is not a change to [the]
original licensing bases and the design for the River Water System,
Service Water System, Intake Structure, and [the] ultimate heat sink
will remain valid for all credible plant conditions, this does not
induce a new mechanism that would result in a different kind of
accident from those previously analyzed.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed changes [sic] only clarif[y] the limiting
design basis ultimate heat sink scenario, consistent with both
Units' original licensing bases. The proposed bounding conditions
bound the credible BVPS Unit 1 and Unit 2 operating conditions. The
design for the River Water System, Service Water System, Intake
Structure, and ultimate heat sink continue to meet General Design
Criteria 2 and 44 and the recommendations of Regulatory Guide 1.27,
Revision 2.
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: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: January 28, 2004.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) to maintain
hydrogen recombiners and hydrogen monitors. Licensees were generally
required to implement upgrades as described in NUREG-0737,
``Clarification of TMI [Three Mile Island] Action Plan Requirements,''
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions
During and Following an Accident.'' Implementation of these upgrades
was an outcome of the lessons learned from the accident that occurred
at TMI, Unit 2. Requirements related to combustible gas control were
imposed by Order for many facilities and were added to or included in
the TSs for nuclear power reactors currently licensed to operate. The
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in
Light-Water-Cooled Power Reactors,'' eliminated the requirements for
hydrogen recombiners and relaxed safety classifications and licensee
commitments to certain design and qualification criteria for hydrogen
and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on September
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the
model NSHC determination in its application dated January 28, 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 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, 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,
[[Page 12371]]
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 the TSs
will not result in a significant reduction in their functionality,
reliability, and availability.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: February 3, 2004
Description of amendment request: This amendment request proposes
to revise a footnote to clarify a surveillance requirement and
associated bases for emergency diesel generator testing.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Section 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
FPL Energy Seabrook, LLC (FPLE Seabrook) proposes to revise
footnote (* * *) of Technical Specification (TS) Surveillance
Requirement (SR) 4.8.1.1.2a.5 to remove the link created between
actions b. and c. of TS 3.8.1.1 and the loaded surveillance testing
requirements of SR 4.8.1.1.2a.6. This revision to footnote (* * *)
is a change to the Technical Specifications that does not modify the
physical design or operation of the plant and will not create a
possibility of an accident. Strict compliance with the footnote
requires paralleling the only operable EDG [emergency diesel
generator] unit with the off-site grid upon entry into action
statement[s] b. or c. of TS 3.8.1.1. Operation of the only operable
EDG unit in this manner may increase its vulnerability for failure
if power from the off-site grid is disturbed or lost. EDG unit
availability for subsequent emergency demands may also be adversely
affected.
The proposed change will eliminate the undesirable link that
presently exists between action statement[s] b. and c. of TS 3.8.1.1
and SR 4.8.1.1.2a.6 but will maintain the primary purpose of the SR,
which is to ensure that the EDG unit is capable of starting from
standby conditions and attaining rated voltage and frequency.
Additionally, the proposed change is consistent with the methodology
used in NRC [Nuclear Regulatory Commission] NUREG-1431, Revision 3,
``Standard Technical Specifications Westinghouse Plants.''
Therefore, the proposed change does not involve a significant
increase [in] the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not affect any plant structures,
systems, or components. The operation of plant systems and equipment
will not be affected by this proposed change. The proposed change to
footnote (* * *) does not have the capability to initiate accidents.
The proposed change will eliminate the undesirable link that
presently exists between action statement[s] b. and c. of TS 3.8.1.1
and SR 4.8.1.1.2a.6. However, the proposed change will maintain the
primary purpose of the SR and supporting footnote, which is to
ensure that the EDG unit is capable of starting from standby
conditions and attaining rated voltage and frequency. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes do not involve a change in the operational
limits or physical design of the plant. The proposed changes do not
change the function or operation of plant equipment or affect the
response of that equipment if it is called on to operate. The
performance capability of the EDG units will not be affected. The
proposed change will maintain the primary purpose of the SR and
supporting footnote, which is to ensure that the EDG unit is capable
of starting from standby conditions and attaining rated voltage and
frequency. Therefore, the proposed change 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: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
Acting NRC Section Chief: Darrell J. Roberts.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 29, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.4.9 Pressure Temperature (P/T)
Curve figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 for Heatup/Cooldown-Core
not Critical, Pressure Test and Heatup/Cooldown-Core Critical
conditions.
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?
The proposed revisions to the Cooper Nuclear Station (CNS) P/T
curves are based on the recommendations in Regulatory Guide (RG)
1.99, Revision 2, and are therefore in accordance with the latest
Nuclear Regulatory Commission (NRC) guidance. The evaluation for the
P/T curves for 32 EFPY [Effective Full Power Years] was performed
using the approved methodologies of 10 CFR [Part] 50, Appendix G.
The curves generated from these methods provide guidance to ensure
that the P/T limits will not be exceeded during any phase of reactor
operation. Accordingly, the proposed revision to the CNS P/T curves
is based on
[[Page 12372]]
an NRC accepted means of ensuring protection against brittle reactor
vessel fracture, and compliance with 10 CFR [Part] 50 Appendix G.
Therefore, this proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Based on the above, NPPD [Nebraska Public Power District]
concludes that the proposed TS change to TS 3.4.9 P/T curves,
figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 does not significantly
increase the probability or consequences of an accident previously
evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change updates existing P/T operating limits to
correspond to the current NRC guidance. The proposed TS change
provides more operating flexibility in the P/T curves for in-service
leakage and hydrostatic pressure testing, non-nuclear heatup and
cooldown, and criticality, with the benefits primarily in the area
of pressure test being performed at a lower temperature. The
proposed change does not involve a physical change to the plant, add
any new equipment or any new mode of operation. These changes
demonstrate compliance with the brittle fracture requirements of 10
CFR [Part] 50 Appendix G, and therefore do not create the
possibility for a new or different kind of accident from any
accident previously evaluated.
Based on the above, NPPD concludes that the proposed TS change
to TS 3.4.9 P/T curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The proposed change to the CNS P/T curves does not create a
significant reduction in the margin of safety. The proposed change
revises the existing CNS P/T curves to be consistent with
recommendations of RG 1.99, Revision 2, the current NRC guidance
given to ensure compliance with 10 CFR [Part] 50 Appendix G.
For P/T curve development ASME [American Society of Mechanical
Engineers] Section Xl Code [Boiler and Pressure Vessel Code] Case N-
640 uses the Kic fracture toughness curve as the lower bound for
fracture toughness. P/T curves based on the Kic fracture toughness
limits enhance industrial safety by expanding the P/T window in the
low-temperature operating region. The potential benefits are a
reduction in the duration of the pressure test and, associated
increase in personnel safety, while conducting inspections in
primary containment. Therefore, operational flexibility is gained
while maintaining an adequate margin of safety to Reactor Pressure
Vessel brittle fracture. As stated above, the development of the P/T
curves to 32 EFPY was performed per the guidelines of 10 CFR [Part]
50 Appendix G, and thus, the margin of safety is not significantly
reduced as the result of the proposed TS change.
Based on the above, NPPD concludes that the proposed TS change
to TS 3.4.9 P/T curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 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: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: February 2, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV)
Vent and Drain Valves,'' to allow a vent or drain line with one
inoperable valve to be isolated instead of requiring the valve to be
restored to Operable status within 7 days.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on February 24, 2003 (68 FR 8637), on possible
amendments to revise the action for one or more SDV vent or drain lines
with an inoperable valve, including a model safety evaluation and model
no significant hazards consideration (NSHC) determination, using the
consolidated line-item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on April 15,
2003 (68 FR 18294). The licensee affirmed the applicability of the
model NSHC determination in its application dated February 2, 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.
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with one valve inoperable instead of requiring the valve to be
restored to operable status within 7 days. With one SDV vent or
drain valve inoperable in one or more lines, the isolation function
would be maintained since the redundant valve in the affected line
would perform its safety function of isolating the SDV. Following
the completion of the required action, the isolation function is
fulfilled since the associated line is isolated. The ability to vent
and drain the SDVs is maintained and controlled through
administrative controls. This requirement assures the reactor
protection system is not adversely affected by the inoperable
valves. With the safety functions of the valves being maintained,
the probability or consequences of an accident previously evaluated
are not significantly increased.
Criterion 2--The proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in the margin of safety.
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDVs is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of an SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: John A. Nakoski.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 9, 2004
Description of amendment request: The proposed change allows entry
into a mode or other specified condition in the applicability of a
technical specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of Title 10 of the Code of
[[Page 12373]]
Federal Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting
Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would
be eliminated, several notes or specific exceptions are revised to
reflect the related changes to LCO 3.0.4, and Surveillance Requirement
(SR) 4.0.4 is revised to reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated February 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 proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 9, 2004.
Description of amendment request: The proposed amendment revises TS
5.5.7, ``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 February 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 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.
Criterion 2--The proposed change does not create the possibility
of a new or different
[[Page 12374]]
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: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: March 28, 2003, as supplemented
December 5, 2003.
Brief description of amendments: These amendments revise the
Technical Specifications by eliminating the requirements associated
with hydrogen recombiners and hydrogen monitors.
Date of issuance: March 2, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 262 and 239.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 13, 2003 (68 FR
25651)
The December 5, 2003, supplemental letter provided clarifying
information that did not enlarge the scope of the amendment as noticed
in the original Federal Register notice or change the no significant
hazards consideration.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated March 2, 2004.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: July 21, 2003, as supplemented
February 5, 2004.
Brief Description of amendments: The amendment revised the Updated
Final Safety Analysis Report (UFSAR) to describe temporary operation of
the turbine building ventilation system in a once-through versus
recirculation configuration during outages.
Date of issuance: February 26, 2004.
Effective date: Effective as of the date of issuance shall be
implemented in accordance with 10 CFR 50.71(e).
Amendment Nos.: 230 and 258.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
approved changes to the UFSAR.
Date of initial notice in Federal Register: August 5, 2003 (68 FR
46241). The February 5, 2004, supplemental letter provided clarifying
information only and did not change the initial proposed no significant
hazards consideration or expand the scope of the initial application.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated February 26, 2004.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: March 20, 2003, as supplemented
by letters dated June 10, September 30, and October 22, 2003
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to update the heatup, cooldown,
criticality, and inservice test pressure and temperature limits for the
reactor coolant system of each unit to a maximum of 34 Effective Full
Power Years. Additionally, the amendments revise the Low Temperature
Overpressure (LTOP) System TSs in order to reflect the revised
pressure-temperature limits and the revised LTOP enable temperature.
Date of issuance: March 4, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
[[Page 12375]]
Amendment Nos.: 212 and 206.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 23, 2003 (68
FR 74264).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 4, 2004.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: June 30, 2003, as supplemented by letter
dated December 16, 2003.
Brief description of amendment: The amendment revises the control
room emergency ventilation system surveillance requirements (SRs) by
modifying an existing SR related to the makeup flow rate to show that
it is applicable to the VSF-9 train and by adding a new makeup flow
rate SR that is applicable to the 2VSF-9 train.
Date of issuance: March 2, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 221.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43384).
The December 16, 2003, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 2, 2004.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendments: March 28, 2003, as supplemented
by letters dated October 23 and December 5, 2003.
Brief description of amendments: The amendments revise the
technical specifications to reduce the main steam line low pressure
primary containment isolation allowable value.
Date of issuance: February 18, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 206/198, 219/213.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 23, 2003 (68
FR 74265). The October 23 and December 5, 2003, submittals provided
clarifying information that did not change the initial proposed no
significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 18, 2004.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: March 31, 2003, as supplemented
June 26, 2003.
Brief description of amendments: The amendments revise Appendix A,
Technical Specifications (TS), of Facility Operating License Nos. NPF-
11 and NPF-18. Specifically, the change increases the upper limit
associated with TS Table 3.3.5.1-1, ``Emergency Core Cooling System
Instrumentation,'' Function 3.e, ``HPCS System Flow Rate--Low
(Bypass),'' Allowable Value from less than or equal to (<=) 1704
gallons per minute (gpm) to <= 2194 gpm.
The change increases the Allowable Value band to account for
instrumentation deadband, as-left setting tolerances and setpoint drift
and to resolve historical difficulties during calibration. The current
Allowable Value was initially provided in the LaSalle County Station TS
during conversion to Improved Technical Specifications (ITS) format.
This value was based on vendor supplied data and believed at the time
to adequately account for these parameters. The upper Allowable Value
limit is being increased based on historical performance data for the
High Pressure Core Spray (HPCS) system flow switches. The increase in
the allowed bypass flow rate does not affect the capability of the HPCS
system in performing its intended safety function.
Date of issuance: March 4, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 165 and 151.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 13, 2003 (68 FR
25654). The supplement dated June 26, 2003, provided clarifying
information that did not change the scope of the March 31, 2003,
application nor the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 4, 2004.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: December 16, 2003 as
supplemented January 29 and February 13, 2004.
Brief description of amendment: This amendment revised the
Technical Specifications to allow a one-time extension of the steam
generator tube inservice inspection interval from March 9, 2004, to
March 31, 2005.
Date of issuance: February 26, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 262.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 6, 2004 (69 FR
695).
The supplements dated January 29 and February 13, 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 the amendment is
contained in a Safety Evaluation dated February 26, 2004.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: November 20, 2003, as
supplemented by letter dated February 5, 2004.
Brief description of amendment: The amendment revised Section
2.1.1.2 of the Technical Specifications to reflect the results of
cycle-specific calculations performed for the upcoming Operating Cycle
10, which would employ a mixed core consisting of predominantly GE11
fuel bundles with some new GE14 fuel bundles.
[[Page 12376]]
Date of issuance: February 25, 2004.
Effective date: As of the date of issuance, to be implemented prior
to startup from Refueling Outage 9.
Amendment No.: 112.
Facility Operating License No. NPF-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 23, 2003 (68
FR 74267).
The supplemental letter of February 5, 2004, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination. The staff's related evaluation of
the amendment is contained in a Safety Evaluation dated February 25,
2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: May 22, 2003, as supplemented
July 9, November 5, December 15, 2003, and January 30, February 9, and
February 20, 2004.
Brief description of amendment: The amendment revised the Kewaunee
Nuclear Power Plant operating license and technical specifications to
increase the licensed rated power by 6.0 percent from 1673 megawatts
thermal to 1772 megawatts thermal.
Date of issuance: February 27, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 172.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34670).
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 amendment is contained
in a Safety Evaluation dated February 27, 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: April 2, 2003, as supplemented
by letters dated August 8 and November 13, 2003.
Brief description of amendments: The amendments revise certain
operational requirements of the Diablo Canyon Nuclear Plant Technical
Specifications for the ventilation filter testing program, the control
room ventilation system, the auxiliary building ventilation system, and
the fuel handling building ventilation system. The amendments also
incorporate a selective implementation of the alternative source term.
Date of issuance: February 27, 2004.
Effective date: February 27, 2004, and shall be implemented within
180 days from the date of issuance.
Amendment Nos.: Unit 1--163; Unit 2--165.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37579).
The August 8 and November 13, 2003, supplemental letters 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 27, 2004.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: July 29, 2003, as supplemented
January 12, 2004.
Brief description of amendment: This amendment revises the
Technical Specifications (TSs) references in the Surveillance
Requirement (SR) 4.0.5 and associated Basis, and Bases 3/4.4.2, 3/
4.4.6, and 3/4.4.10. In the current plant TSs, the reference for
inservice testing (IST) and inservice inspection (ISI) activities is
the American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME BPV Code), Section XI. The licensee proposed to reference
the ASME Code for Operation and Maintenance of Nuclear Power Plants
(ASME OM Code) and the ASME BPV Code, Section XI for IST activities and
ISI activities respectively. These changes reflect the fact that the
pump and valve testing requirements previously contained in Subsections
IWP and IWV of the ASME BPV Code, Section XI, have been replaced by the
requirements in the 1998 Edition of the ASME OM Code, 2000 Addenda, for
the licensee's third 120-month IST interval. These TS changes are
required to implement the IST program update in accordance with the
requirements of 10 CFR.55a(f)(5)(ii). The licensee also proposed
certain other language changes.
Date of issuance: February 18, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 166.
Facility Operating License No. NPF-12: Amendment revised the TSs.
Date of initial notice in Federal Register: October 14, 2003 (68 FR
59219). The supplemental letter 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 February 18, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear
Plant, Unit 3, Limestone County, Alabama
Date of application for amendments: October 1, 2003, as
supplemented December 19, 2003.
Description of amendment request: The amendment revised the safety
limit minimum critical power ratio values in Technical Specification
(TS) 2.1.1.2.
Date of issuance: February 24, 2004.
Effective date: February 24, 2004.
Amendment No.: 246.
Facility Operating License No. DPR-68: Amendment revised the TSs.
Date of initial notice in Federal Register: October 28, 2003 (68 FR
61481). The December 19, 2003, letter 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 February 24, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 8th day of March 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-5596 Filed 3-15-04; 8:45 am]
BILLING CODE 7590-01-P