[Federal Register Volume 69, Number 41 (Tuesday, March 2, 2004)]
[Notices]
[Pages 9857-9871]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-4343]
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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, February 5, 2004, through February 19,
2004. The last biweekly notice was published on February 17, 2004 (69
FR 7517).
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
[[Page 9858]]
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also 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 file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
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.
[[Page 9859]]
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 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, Docket No. 50-219, Oyster Creek
Generating Station, Ocean County, New Jersey, Three Mile Island Nuclear
Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: January 30, 2004.
Description of amendment request: The licensee proposes to revise
the operating licenses to reflect the current 100% ownership of AmerGen
by Exelon Generation Company. In particular, the proposed amendments
will remove PECO and British Energy from the licenses, and will remove
certain license conditions in their entirety which were imposed to
acknowledge the indirect foreign ownership in AmerGen by British Energy
plc. Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes are administrative in nature and would
merely conform the facility operating licenses to reflect the
current ownership structure of AmerGen. No actual plant equipment or
accident analyses will be affected by the proposed changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature and would
merely conform the facility operating licenses to reflect the
current ownership structure of AmerGen. No actual plant equipment or
accident analyses will be affected by the proposed changes and no
failure modes not bounded by previously evaluated accidents will be
created.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is administrative in nature and would merely
conform the facility operating licenses to reflect the current
ownership structure of AmerGen. No actual plant equipment or
accident analyses will be affected by the proposed changes.
Additionally, the proposed changes will not relax any criteria used
to establish safety limits, will not relax any safety system
settings, or will not relax the bases for any limiting conditions
for operation. Therefore, the proposed changes do not involve a
significant reduction in any 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.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2 (IP2), Westchester County, New York
Date of amendment request: January 29, 2004.
Description of amendment request: The proposed amendment would
increase the maximum authorized reactor core power level from 3114.4
megawatt thermal (MWt) to 3216 MWt. This represents a nominal increase
of 3.26% rated thermal power. 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 evaluations and analyses associated with this proposed
change to core power level have demonstrated that all applicable
acceptance criteria for plant systems, components, and analyses
(including the Final Safety Analysis Report Chapter 14 safety
analyses) will continue to be met for the proposed increase in
licensed core thermal power for IP2. The subject increase in core
thermal power will not result in conditions that could adversely
affect the integrity (material, design, and construction standards)
or the operational performance of any potentially affected system,
component or analysis. Therefore, the probability of an accident
previously evaluated is not affected by this change. The subject
increase in core thermal power will not adversely affect the ability
of any safety-related system to meet its intended safety function.
Further, the radiological dose evaluations in support of this power
uprate effort show all acceptance criteria are met.
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 evaluations of this proposed amendment show that all
applicable acceptance criteria for plant systems, components, and
analyses (including FSAR [final safety analysis report] Chapter 14
safety analyses) will continue to be met for the proposed power
increase in IP2 licensed core thermal power. The subject increase in
core thermal power will not result in conditions that could
adversely affect the integrity (material, design, and construction
standards) or operational performance of any potentially affected
system, component, or analyses. The subject increase in core thermal
power will not adversely affect the ability of any safety-related
system to meet its safety function. Furthermore, the conditions and
changes associated with the subject increase in core thermal power
will neither cause initiation of any accident, nor create any new
credible limiting single failure. The power uprate does not result
in changing the status of events previously deemed to be non-
credible being made credible. Additionally, no new operating modes
are proposed for the plant as a result of this requested change.
Therefore, the subject increase in core thermal power level will
not create the
[[Page 9860]]
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 evaluations associated with this proposed change show that
all applicable acceptance criteria for plant systems, components,
and analyses (including FSAR Chapter 14 safety analyses) will
continue to be met for this proposed increase in IP2 licensed core
thermal power. The subject increase in core thermal power will not
result in conditions that could adversely affect the integrity
(material, design, and construction standards) or operational
performance of any potentially affected system, component, or
analysis. The subject power uprate will not adversely affect the
ability of any safety-related system to meet its intended safety
function.
Therefore, the subject increase in core thermal power 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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: February 9, 2004.
Description of amendment request: The proposed amendment would
remove the pressurizer heatup and cooldown limits, and the associated
action and surveillance requirements, from the Technical Specifications
and place them in a licensee controlled document.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of an accident is unchanged as a result of the
proposed change to delete the ANO-2 [Arkansas Nuclear One, Unit 2]
pressurizer heatup and cooldown rates and associated action,
surveillance requirement, and bases from the TS [Technical
Specification]. The cooldown and heatup rates are not initiators to
any accidents or pressurizer transients discussed in the ANO-2 SAR
[Safety Analysis Report]. Therefore, the probability of an accident
is not changed.
The purpose of the pressurizer heatup and cooldown limits is to
ensure that given transient events will not negatively affect the
pressurizer structural integrity beyond Code [American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code]
allowables. These limits will be maintained within ASME Code
allowables in a licensee controlled document in accordance with 10
CFR 50.59. Therefore, the consequences of an accident are not
increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The limitations imposed on the pressurizer heatup and cooldown
rates are provided to assure that the pressurizer is operated within
the design criteria assumed for the flaw evaluation and fatigue
analysis performed in accordance with the ASME Code Section XI,
subsection IWB-3600 requirements. The ANO-2 SAR has analyzed the
conditions that would result from a thermal or pressurization
transient on the ANO-2 pressurizer. The proposed deletion of the
pressurizer heatup and cooldown rates and relocation of the limits
to a licensee controlled document does not change the way that the
pressurizer is designed or operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established by the rules contained in
the ASME Section III Code. Any future changes to the cooldown or
heatup rates will be evaluated using 10 CFR 50.59, ``Changes, Tests
and Experiments,'' and are required to meet the ASME Code margins.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
277, Peach Bottom Atomic Power Station, Unit 2, York and Lancaster
Counties, Pennsylvania
Date of application for amendment: February 12, 2004.
Description of amendment request: The proposed amendment would
revise technical specification (TS) Table 3.3.6.1-1, ``Primary
Containment Isolation Instrumentation,'' to increase the TS Allowable
Value (AV) related to the setpoint for the Main Steam Tunnel
Temperature--High system isolation function for those instruments
located within the Reactor Building. A new Function, 1.f, would be
added to represent the Reactor Building Main Steam Tunnel Temperature--
High. Existing Function 1.e would be renamed to clarify that it
represents only the Turbine Building Main Steam Tunnel Temperature--
High.
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 leak detection instrumentation associated with the proposed
amendment is designed to detect Main Steam Line leakage in the range
of one to ten percent of rated steam flow. This design basis remains
unchanged. This ensures that the criteria for acceptance as
established in the original licensing bases remains valid. The
previous analysis for establishing the allowable value for Main
Steam Line Tunnel High temperature in the Reactor Building can be
improved using industry standard, state of the art computer modeling
techniques. The new analysis using the GOTHIC computer code is
appropriate because it accurately accounts for the building heat
structures, HVAC effects, and outside air temperatures. The proposed
change increases the operating margin, which reduces the potential
for unnecessary plant transients. Raising the setpoint causes a
greater time to detect the leak, but remains bounded by existing
analysis for the design basis break of the main steam line
documented in Table 14.9.8 of the Peach Bottom [Updated Final Safety
Analysis Report] UFSAR. There are no impacts on equipment
qualification. Changes to the instrumentation used to detect a steam
line leak do not affect the probability of occurrence of the leak.
Hence, it is concluded that raising the allowable value for Reactor
Building Main Steam Tunnel high temperature does not significantly
increase the probability or consequences of an accident previously
evaluated.
[[Page 9861]]
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 amendment does not impact the physical design or
location of the associated leak detection instrumentation. The leak
detection instrumentation associated with the proposed amendment
will continue to detect main steam line leakage in the range of one
to ten percent of rated steam flow. The instruments will still
initiate the automatic isolation of the appropriate containment
isolation valves to mitigate steam leakage as credited in the
original licensing bases. This proposed amendment is associated only
with the results of a main steam line leak in the Reactor Building
portion of the Main Steam Tunnel and has no impact on the initiation
of this leak. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Steam leaks in the affected area of the Reactor Building will be
detected on a timely basis so that the Group 1 Primary Containment
Isolation Valves are promptly closed. The analysis performed for the
proposed amendment demonstrates that the appropriate instruments
will promptly initiate automatic system isolation upon sensing a
temperature in excess of the new setpoint. Therefore, the proposed
amendment ensures that the criteria for acceptance as established in
the original licensing bases remain valid. Further, the proposed
amendment eliminates a potential cause for unnecessary plant
shutdowns created by conditions other than a main steam line leak.
Equipment qualification and structural integrity of systems,
structures, and components located within the Reactor Building are
not affected by the proposed amendment. Therefore, the proposed
amendment does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: Mr. Edward Cullen, Vice President and
General Counsel, Exelon Generation Company, LLC, 2301 Market Street,
S23-1, Philadelphia, PA 19101.
NRC Acting Section Chief: Darrell J. Roberts.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: February 4, 2004.
Description of amendment request: This amendment request proposes
to update the Technical Specifications (TSs) to correct a non-
conservatism in a TS Table, correct a reference error, update titles,
incorporate formatting changes to increase ease of use, and remove a
permit issuance date to ease administrative burden.
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.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. In addition, the proposed changes do not
affect the manner in which the plant responds in normal operation,
transient or accident conditions nor do they change any of the
procedures related to operation of the plant. The proposed changes
do not alter or prevent the ability of structures, systems and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the acceptance limits
assumed in the Updated Final Safety Analysis Report (UFSAR). The
proposed changes are editorial in nature and only correct, update
and modify the Technical Specifications and Environmental Protection
Plan.
The proposed changes do not affect the source term, containment
isolation or radiological release assumptions used in evaluating the
radiological consequences of an accident previously evaluated in the
Seabrook Station UFSAR. Further, the proposed changes do not
increase the types and amounts of radioactive effluent that may be
released offsite, and do not significantly increase individual or
cumulative occupational/public radiation exposures.
Based on the above, the proposed changes will not significantly
increase 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 changes do not change the operation or the design
basis of any plant system or component during normal or accident
conditions. The proposed changes do not include any physical changes
to the plant. In addition, the proposed changes do not change the
function or operation of plant equipment or introduce any new
failure mechanisms. The plant equipment will continue to respond per
the design and analyses and there will not be a malfunction of a new
or different type introduced by the proposed changes.
The proposed changes are editorial in nature and only update
Seabrook Station Technical Specifications and Environmental
Protection Plan to provide consistency and facilitate ease of use.
The proposed changes do not modify the facility nor do they affect
the plant's response to normal, transient or accident conditions.
The changes do not introduce a new mode of plant operation. The
changes do not affect plant safety. The plant's design and design
basis are not revised and the current safety analyses remain in
effect.
Thus, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The propose changes do not involve a significant reduction in
the margin of safety.
The proposed changes are editorial changes to the Seabrook
Station Technical Specifications and Environmental Protection Plan.
The safety margins established through Limiting Conditions for
Operation, Limiting Safety System Settings and Safety Limits as
specified in the Technical Specifications are not revised nor is the
plant design or its method of operation revised by the proposed
changes.
Thus, it is concluded that 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: 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: December 9, 2003.
Description of amendment request: The proposed amendment request
would: (1) Incorporate into the Updated Safety Analysis Report the
overall Main Steam Isolation Valve (MSIV) Leakage Pathway configuration
(including the post-accident manual actions necessary to establish that
configuration) upon Nuclear Regulatory Commission (NRC) approval, (2)
incorporate into the Cooper Nuclear Station (CNS) licensing basis the
loss-of-coolant accident (LOCA) dose calculation methodology (currently
approved on an interim basis) upon permanent approval by the NRC, and
(3) delete License Condition 2.C.(6), eliminating the commitment to
provide potassium iodide to the control room occupants during LOCA
conditions with core damage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 9862]]
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 ALT [alternate leakage treatment] pathway was determined
using the NRC-endorsed method described in Reference 7.3 [NEDC-
31858P-A Class III, August 1999, ``BWROG [Boiling Water Reactor
Owners Group] Report for Increasing MSIV Leakage Rate Limits and
Elimination of Leakage Control Systems'']. The proposed manual
actions to establish that configuration are designed to assure that
MSIV leakage resulting after a LOCA with core damage will reach the
Main Turbine Condenser via a pathway that has been evaluated as
being seismically robust. The LOCA dose calculation methodology
assumes this leakage reaches the turbine condenser complex. The
manual actions are simple to perform and there are no concerns for
personnel safety in carrying out these actions within the timeframes
established. Accordingly, there is no significant increase in
probability or consequences of a previously evaluated accident.
The LOCA dose calculation methodology is already approved on an
interim basis, as documented in Reference 7.1 [letter to C. Warren
(NPPD) [Nuclear Public Power District] from U.S. Nuclear Regulatory
Commission dated February 21, 2003, ``Cooper Nuclear Station--
Issuance of Amendment Regarding Design Basis Accidents''
Radiological Dose Assessment Methodologies, and Revision to License
Condition 2.C.(6) (TAC No. MB4654)'']. As there are no technical
issues to resolve, the effects of permanent approval on the
probability or consequences of an accident are bounded by the
previous safety conclusions of License Amendment 196.
The deletion of License Condition 2.C.(6), following
implementation of the seismic evaluation and permanent approval of
the LOCA dose calculation methodology, is an administrative change
to the CNS Operating License. Therefore, there are no associated
effects on the probability or consequences of previously evaluated
accidents.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No
The proposed changes only involve the treatment of the Loss-of-
Coolant Accident. No other new or different kinds of accidents can
be created by the proposed changes.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No
The LOCA dose calculation methodology credits MSIV leakage
plateout in the Main Turbine Condenser prior to release to the
Turbine Building. The ALT pathway to the Main Turbine Condenser was
determined using the NRC-endorsed method described in Reference 7.3.
Therefore, the effects on safety margins due to crediting this
configuration are bounded by the NRC Safety Evaluation conclusions
on this methodology. Using the MSIV leakage assumed in the LOCA
analysis and conservative assumptions, there is sufficient time for
the CNS personnel to take the simple actions necessary to configure
the pathway, and thereby assure that the radiological consequences
are bounded by the LOCA dose calculation methodology results.
Accordingly, there is no significant reduction in safety margin.
The LOCA dose calculation methodology is already approved on an
interim basis, as documented in Reference 7.1. As there are no
technical issues to resolve, the effects of permanent approval on
the [ ] [margin of safety] are bounded by the previous safety
conclusions of License Amendment 196.
The deletion of License Condition 2.C.(6), following
implementation of the seismic evaluation and permanent approval of
the LOCA dose calculation methodology, is an administrative change
to the CNS Operating License. Therefore, there are no associated
effects on safety margins.
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.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa;Docket No. 50-305, Kewaunee Nuclear Power
Plant, Kewaunee County, Wisconsin;Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota;Docket No. 50-255, Palisades
Plant, Van Buren County, Michigan;Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin;Docket Nos. 50-282 and 50-306, Prairie Island Nuclear
Generating Plant, Units 1 and 2, Goodhue County, Minnesota
Date of amendment request: January 30, 2004.
Description of amendment request: The proposed amendment deletes
requirements in 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] 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 TS 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 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 30, 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
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 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 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
[[Page 9863]]
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: Jonathan Rogoff, Morgan Lewis, 1111
Pennsylvania Avenue NW.,Washington, DC 20004.
NRC Section Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: December 1, 2003.
Description of amendment request: The proposed changes to the Fort
Calhoun Technical Specifications (TSs) consist primarily of
typographical changes and relocation of material not required to be in
the TSs. The licensee has proposed changes to the following TSs: (1)
Item 14 of Table 3-3 regarding testing of nuclear detector well cooling
annulus exit air temperature detectors, (2) the title of Item of 10a.2
of Table 3-5, (3) TS Section 3.17(5)(ii), (4) TS Section 5.5, ``Review
and Audit,'' (5) TS Section 5.6, ``Reportable Event Action,'' (6) TS
Sections 5.7.1.b, 5.7.1.c, and 5.7.1.d, (7) TS Section 5.9.1.a,
``Startup Report,'' and (8) TS Section 5.9.4.c, ``Fire Protection
Program Deficiency Report.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change relocates requirements for Nuclear Detector
Cooling that do not meet the criteria for inclusion in the TS set
forth in 10 CFR 50.36(c)(2)(ii). The requirements for Nuclear
Detector Cooling are being relocated from TS to the USAR [Updated
Safety Analysis Report], which will be maintained pursuant to 10 CFR
50.59, thereby reducing the level of regulatory control. The level
of regulatory control has no impact on the probability or
consequences of an accident previously evaluated. Therefore, the
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The correction of typographical errors and relocation of
specifications is not an initiator of any previously evaluated
accident. The proposed changes will not prevent safety systems from
performing their accident mitigation function as assumed in the
safety analysis.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change relocates requirements for Nuclear Detector
Cooling that do not meet the criteria for inclusion in TS set forth
in 10 CFR 50.36(c)(2)(ii). The proposed change only affects the
technical specifications and does not involve a physical change to
the plant. Modifications will not be made to existing components nor
will any new or different types of equipment be installed. The
proposed change corrects typographical errors and relocates
information that is unnecessary in the TS. This change will not
alter assumptions made in safety analysis and licensing bases.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change relocates requirements for Nuclear Detector
Cooling that do not meet the criteria for inclusion in TS set forth
in 10 CFR 50.36(c)(2)(ii). The change will not reduce a margin of
safety since the location of a requirement has no impact on any
safety analysis assumptions. In addition, the relocated requirements
for Nuclear Detector Cooling remain the same as the existing TS.
Since any future changes to these requirements or the surveillance
procedures will be evaluated per the requirements of 10 CFR 50.59,
there will be no reduction in a margin of safety.
The additional proposed changes correct typographical errors and
relocate redundant information not required to be in the TS.
[[Page 9864]]
Therefore, this technical specification 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: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
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 amendment requests: December 30, 2003.
Description of amendment requests: The proposed amendment deletes
the requirements from the technical specifications (TS) 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 TS 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 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 December 30, 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.
Category 1 in RG 1.97 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 severe accident
management guidelines (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, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a
neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
[[Page 9865]]
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: December 30, 2003.
Brief description of amendments: The amendment revised the
Administrative Controls Section 5.1.5 to state any Senior Reactor
Operator may be designated to be responsible for the control room
command function.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.92(c), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to Technical Specifications Administrative
Controls Section 5.1.5, involves the use of a more generic
designation of SRO [Senior Reactor Operator] for the unit staff
position responsible for the control room command function. Since
the proposed change is administrative in nature, it does not involve
any physical changes to any structures, systems, or components, nor
will their performance requirements be altered. The proposed change
also does not affect the operation, maintenance, or testing of the
plant. Therefore, the response of the plant to previously analyzed
accidents will not be affected. Consequently, 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 previously evaluated?
As a result of the proposed change to the Technical
Specifications, the qualification requirements for the unit staff
position responsible for the control room command function will
remain unchanged and the plant staff will continue to meet
applicable regulatory requirements. Also, since no change is being
made to design, operation, maintenance, or testing of the plant, no
new methods of operation or failure modes are introduced by the
proposed change. Therefore, the possibility of a new or different
kind of accident from any previously evaluated is not created.
3. Does the proposed change involve a significant decrease in
the margin of safety?
The proposed change to the Technical Specifications will have no
adverse impact on the onsite organizational features necessary to
assure safe operation of the plant since the qualification
requirements for the unit staff position for the control room
command function remain unchanged. The adoption of the more generic
designation of SRO for the individual responsible for control room
command function will also reduce the regulatory burden of having to
devote limited resources to process a license amendment whenever a
title change for this position is implemented, thus improving plant
efficiency. Therefore, the proposed change does not invoice a
significant decrease 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: John A. Nakoski.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: February 3, 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 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 3, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change 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 Limiting
Conditions for Operation (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.
[[Page 9866]]
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: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: January 21, 2004.
Brief description of amendments: The amendment would revise
Technical Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS)
Instrumentation,'' 3.3.2, ``Engineered Safety Feature Actuation System
(ESFAS) Instrumentation,'' and 3.3.6, ``Containment Ventilation
Isolation Instrumentation.'' The purpose of the amendment is to adopt
the completion time, test bypass time, and surveillance frequency time
changes approved by the NRC in Topical Reports WCAP-14333-P-A,
``Probabilistic Risk Analysis of the RPS [reactor protection system]
and ESFAS Test Times and Completion Times,'' and WCAP-15376-P-A,
``Risk-Informed Assessment of the RTS and ESFAS Surveillance Test
Intervals and Reactor Trip Breaker Test and Completion Times.'' The
proposed changes would revise the required actions for certain action
conditions; increase the completion times for several required actions
(including some notes); delete notes in certain required actions; and
increase frequency time intervals (including certain notes) in several
surveillance requirements (SRs).
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?
Response: No.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The same reactor trip system (RTS)
and engineered safety feature actuation system (ESFAS)
instrumentation will continue to be used. The protection systems
will continue to function in a manner consistent with the plant
design basis. These changes to the Technical Specifications [in the
amendment] do not result in a condition where the design, material,
and construction standards that were applicable prior to the change
are altered.
The proposed changes will not modify any system interface. The
proposed changes will not affect the probability of any event
initiators [because the proposed changes are not event initiators].
There will be no degradation in the performance of or an increase in
the number of challenges imposed on safety-related equipment assumed
to function during an accident situation. There will be no change to
normal plant operating parameters or accident mitigation
performance. The proposed changes will not alter any assumptions or
change any mitigation actions in the radiological consequence
evaluations in the FSAR [Comanche Peak Final Safety Analysis
Report].
The determination that the results of the proposed changes are
acceptable [to be considered for plant-specific Technical
Specifications] was established in the NRC Safety Evaluations
prepared for WCAP-14333-P-A (issued by letter dated July 15, 1998)
and for WCAP-15376-P-A (issued by letter dated December 20, 2002).
Implementation of the proposed changes will result in an
insignificant risk impact. Applicability of these conclusions has
been verified through plant-specific reviews and implementation of
the generic analysis results in accordance with the respective NRC
Safety Evaluation conditions [for the two WCAPs].
The proposed changes to the Completion Times, test bypass times,
and Surveillance Frequencies reduce the potential for inadvertent
reactor trips and spurious ESF [engineered safety feature]
actuations, and therefore do not increase the probability of any
accident previously evaluated. The proposed changes do not change
the response of the plant to any accidents and have an insignificant
impact on the reliability of the RTS and ESFAS signals. The RTS and
ESFAS will remain highly reliable and the proposed changes will not
result in a significant increase in the risk of plant operation.
This is demonstrated by showing that the impact on plant safety as
measured by the increase in core damage frequency (CDF) is less than
1.0E-06 per year and the increase in large early release frequency
(LERF) is less than 1.0E-07 per year. In addition, for the
Completion Time changes, the incremental conditional core damage
probabilities (ICCDP) and incremental conditional large early
release probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08,
respectively. These changes meet the acceptance criteria in
Regulatory Guides 1.174 and 1.177. Therefore, since the RTS and
ESFAS will continue to perform their [safety] functions with high
reliability as originally assumed, and the increase in risk as
measured by ``CDF, ``LERF, ICCDP, ICLERP risk metrics is within the
acceptance criteria of existing [NRC] regulatory guidance, there
will not be a significant increase in the consequences of any
accidents.
The proposed changes do 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. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended [safety] function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed changes do not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. The proposed changes are consistent with safety analysis
assumptions and resultant consequences.
Therefore, [the] change[s do] not 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?
Response: No.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. The proposed changes will not affect the normal method of
plant operation. No performance requirements will be affected or
eliminated. The proposed changes will not result in physical
alteration to any plant system nor will there be any change in the
method by which any safety-related plant system performs its safety
function.
There will be no setpoint changes or changes to accident
analysis assumptions.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). There will be no effect on the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions. There will be no impact on the overpower limit, DNBR
[departure from nucleate boiling ratio] limits, FQ [heat
flux hot channel factor], F[Delta]H [nuclear enthalpy
rise hot channel factor], LOCA PCT [loss-of-
[[Page 9867]]
coolant accident peak cladding temperature], peak local power
density, or any other margin of safety. The radiological dose
consequence acceptance criteria listed in the [NRC] Standard Review
Plan will continue to be met.Redundant RTS and ESFAS trains are
maintained, and diversity with regard to the signals that provide
reactor trip and engineered safety features actuation is also
maintained. All signals credited as primary or secondary, and all
operator actions credited in the accident analyses will remain the
same. The proposed changes will not result in plant operation in a
configuration outside the design basis. The calculated impact on
risk is insignificant and meets the acceptance criteria contained in
Regulatory Guides 1.174 and 1.177. Although there was no attempt to
quantify any positive human factors benefit due to increased
Completion Times and bypass test times, it is expected that there
would be a net benefit due to a reduced potential for spurious
reactor trips and actuations associated with testing.
Implementation of the proposed changes is expected to result in
an overall improvement in safety, as follows:
(a) Reduced testing will result in fewer inadvertent reactor
trips, less frequent actuation of ESFAS components, less frequent
distraction of operations personnel without significantly affecting
RTS and ESFAS reliability.
(b) Improvements in the effectiveness of the operating staff in
monitoring and controlling plant operation will be realized. This is
due to less frequent distraction of the operators and shift
supervisor to attend to instrumentation Required Actions with short
Completion Times.
(c) Longer repair times associated with increased Completion
Times will lead to higher quality repairs and improved reliability.
(d) The Completion Time extensions for the reactor trip breakers
will provide the utilities additional time to complete test and
maintenance activities while at power, potentially reducing the
number of forced outages related to compliance with reactor trip
breaker Completion Times, and provide consistency with the
Completion Times for the logic trains.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed NoSignificant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3
Date of amendment request: November 13, 2003.
Brief description of amendment request: The proposed amendment
would allow an increase in the licensed power from 3441 megawatts
thermal (MWt) to 3716 MWt. This represents an increase of approximately
8 percent above the current rated licensed thermal power. The proposed
amendment would also change the operating license and the technical
specifications appended to the operating license to provide for
implementing uprated power operation.
Date of publication of individual notice in Federal Register:
February 5, 2004.
Expiration date of individual notice: March 8, 2004.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 30, 2004.
Brief description of amendment request: The proposed amendment
would revise the Cooper Nuclear Station (CNS) Technical Specifications
(TS), by adding a temporary note to allow a one-time extension of a
limited number of TS Surveillance Requirements (SRs). The temporary
note states that the next required performance of the SR may be delayed
until the current cycle refueling outage, but no later than February 2,
2005, and it expires upon startup from the refueling outage. With the
exception of one SR, the period of additional time requested occurs
during the next planned refueling outage.
Date of publication of individual notice in Federal Register:
February 12, 2004 (69 FR 7023).
Expiration date of individual notice: March 15, 2004.
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].
[[Page 9868]]
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: December 2, 2003.
Brief description of amendment: The amendment revised Surveillance
Requirement (SR) 4.0.2 of the Technical Specifications (TSs) to extend
the delay period, before entering a Limiting Condition for Operation,
following a missed surveillance. The delay period is extended from the
current limit of ``* * * up to 24 hours or up to the limit of the
specified frequency, whichever is less'' to ``* * * up to 24 hours or
up to the limit of the specified frequency, whichever is greater.'' The
revised SR 4.0.2 specifies that a risk evaluation shall be performed
for any surveillance delayed greater than 24 hours and the risk impact
shall be managed. In addition, a new Section 6.21 is added to provide
for a TS Bases Control Program.
Date of Issuance: February 5, 2004.
Effective date: February 5, 2004 and shall be implemented within 60
days of issuance.
Amendment No.: 240.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 6, 2004 (69 FR
692).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 5, 2004.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: June 11, 2003, as supplemented
August 20 and October 13, 2003.
Brief description of amendment: The amendment allows the licensee
to extend its Appendix J, Type A, Containment Integrated Leak Rate
Test, Option B, for H. B. Robinson Steam Electric Plant, Unit No. 2,
from the scheduled May 2004 timeframe to no later than April 9, 2007.
Date of issuance: February 11, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No. 199.
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 23, 2003 (68
FR 74264).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 11, 2004.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket No. 50-270, Oconee Nuclear Station,
Unit 2, Oconee County, South Carolina
Date of application of amendment: October 28, 2003.
Brief description of amendment: The amendment revised the licensing
basis in the Updated Final Safety Analysis Report (UFSAR) to support
installation of a passive low-pressure injection (LPI) cross connect
inside containment. The changes to the UFSAR revise the licensing basis
for selected portions of the core flood and LPI/Decay Heat Removal
piping to allow exclusion of the dynamic effects associated with
postulated rupture of that piping by application of leak-before-break
technology.
Date of Issuance: February 5, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 338.
Renewed Facility Operating License No. DPR-47: Amendment revised
the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68661)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 5, 2004.
No significant hazards consideration comments received: No.
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 application for amendment: May 12, 2003, as revised by
letters dated December 5 and 18, 2003.
Brief description of amendment: By letter dated December 5, 2003,
Entergy submitted a revised application for amendment to Grand Gulf
Nuclear Station, Unit 1 Technical Specification (TS) 3.3.6.1, ``Primary
Containment and Drywell Isolation Instrumentation,'' to add a provision
to the APPLICABILITY function that will eliminate the requirement that
the Residual Heat Removal System Isolation, Reactor Vessel Water Level-
Low, Level 3, be OPERABLE under certain conditions during refueling
outages. Specifically, the proposed change requested in the original
application dated May 12, 2003, would remove the requirement for this
isolation function, specified in Table 3.3.6.1-1, when the upper
containment reactor cavity is at the High Water Level condition
specified in TS 3.5.2, ``Emergency Core Cooling Systems (ECCS)
Shutdown.'' The revised application adds a new surveillance requirement
(SR) (SR 3.3.6.1.9) to verify every four hours that the water level in
the upper containment pool is greater than or equal to 22 feet 8 inches
above the reactor pressure vessel flange, and adds a footnote to Table
3.3.6.1-1, Item 5.b, for MODE 5 that states that the function is not
required when the upper containment reactor cavity and transfer canal
gates are removed and SR 3.3.6.1.9 is met. The proposed SR and footnote
are only applicable in MODE 5. The May 12, 2003, application was
previously noticed in the Federal Register on June 10, 2003 (68 FR
34665).
Date of issuance: January 23, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 163.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 15, 2003 (68
FR 69726). The December 18, 2003, supplemental letter provided
clarifying information that did not change the scope of the December
15, 2003, Federal Register notice or the no significant hazards
consideration determination therein.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 23, 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: August 15, 2003, as supplemented
by letter on September 15, 2003.
Brief description of amendment: The amendment revised the reactor
coolant system pressure-temperature limit curves in Section 3.4.11,
``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,''
of the Technical Specifications. The revised curves are effective up to
22 effective full-power years.
Date of issuance: January 27, 2004.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 110.
[[Page 9869]]
Facility Operating License No. NPF-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 2, 2003 (68
FR 52235).
The staff's related evaluation of the amendment is contained in a
Safety Evaluation dated January 27, 2004.
No significant hazards consideration comments received: No
The September 15, 2003, letter provided clarifying information
within the scope of the original application and did not change the
staff's initial proposed no significant hazards consideration
determination.
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: August 28, 2003.
Brief description of amendment: The amendment revised Section
3.1.7, ``Standby Liquid Control (SLC) System,'' of the Technical
Specifications to support a transition from GE11 to GE14 fuel in the
reactor core. The revised Section 3.1.7 raises the required calculated
average boron concentration in the reactor from a concentration
equivalent to 660 parts per million (ppm) natural boron to 780 ppm
natural boron. The increased concentration is achieved by requiring use
of sodium pentaborate solution enriched with the boron-10 isotope.
Date of issuance: February 13, 2004.
Effective date: As of the date of issuance to be implemented prior
to startup from Refueling Outage 9.
Amendment No.: 111.
Facility Operating License No. NPF-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 30, 2003 (68
FR 56345). The staff's related evaluation of the amendment is contained
in a Safety Evaluation dated February 13, 2004.
No significant hazards consideration comments received: No.
PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: February 14, 2003, as
supplemented on October 2, 2003.
Brief description of amendments: The amendments modify the Salem
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications
(TSs) by: (1) Adding new TS 3/4.7.11, ``Fuel Storage Pool Boron
Concentration,'' to define spent fuel pool boron concentration limits;
(2) relocating fuel assembly storage requirements currently located in
TS 5.6.1.2d to a new TS 3/4.7.12, ``Fuel Assembly Storage in the Spent
Fuel Pool;'' and (3) relocating refueling boron concentration
requirements from TS 3/4.9.1, ``Boron Concentration,'' to the Core
Operating Limits Report.
Date of issuance: February 6, 2004.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 262 and 244.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: April 29, 2003 (68 FR
22753).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 6, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: March 13, 2002, as supplemented
on April 1 and November 21, 2003.
Brief description of amendment: The amendment approves revisions to
the Updated Final Safety Analysis Report (UFSAR) to update the quality
assurance criteria and the basis for the seismic qualification of the
ducting installed as part of the suspended ceiling air delivery system
in the main control room.
Date of issuance: February 12, 2004.
Effective date: As of the date of issuance and shall be implemented
in accordance with 10 CFR 50.71(e).
Amendment No.: 50.
Facility Operating License No. NPF-90: Amendment revised the UFSAR.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18286). The supplemental letters provided clarifying information that
did not expand the scope of the original request 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 12, 2004.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an
[[Page 9870]]
opportunity for public comment. If comments have been requested, it is
so stated. In either event, the State has been consulted by telephone
whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland. Publicly available
records will be accessible from the Agencywide Documents Access and
Management System's (ADAMS) Public Electronic Reading Room on the
Internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html.
If you do not have access to ADAMS or if there are problems in
accessing the documents located in ADAMS, contact the NRC Public
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or
by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1-800-397-4209, 301-415-4737, or by e-mail to [email protected]. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to file such a supplement which satisfies these requirements
with respect to at least one contention will not be permitted to
participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to 301-415-3725 or by email to
[email protected]. A copy of the request for hearing and
[[Page 9871]]
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).
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: February 6, 2004.
Description of amendment request: The amendment changes the
implementation date from 30 days to 120 days for Amendment No. 224
issued on January 16, 2004, that approved a measurement uncertainty
uprate to increase the licensed rated power by 1.6 percent from 1500
megawatts thermal (MWt) to 1524 MWt.
Date of issuance: February 13, 2004.
Effective date: February 13, 2004, and the fully implemented date
for Amendment No. 224 (issued January 16, 2004) is changed to 120 days.
Amendment No.: 225.
Renewed Facility Operating License No. DPR-40: Amendment revises
the implementation date for Amendment No. 224.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Omaha-World Herald. The notice provided an
opportunity to submit comments on the Commission's proposed NSHC
determination. No comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, State consultation, and final NSHC determination
are contained in a safety evaluation dated February 13, 2004.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: February 5, 2004.
Brief description of amendment: The amendment revises Technical
Specification 3.7.5, ``Auxiliary Feedwater (AFW) System'' to
incorporate a one-time provision that extends the allowed outage time
for an inoperable turbine-driven auxiliary feedwater pump.
Date of issuance: February 6, 2004.
Effective date: February 6, 2004.
Amendment No.: 158.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a safety evaluation dated
February 6, 2004.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Dated at Rockville, Maryland, this 20th day of February 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management,Office of Nuclear
Reactor Regulation.
[FR Doc. 04-4343 Filed 3-1-04; 8:45 am]
BILLING CODE 7590-01-P