[Federal Register Volume 70, Number 49 (Tuesday, March 15, 2005)]
[Notices]
[Pages 12743-12755]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-4792]
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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 18, 2005, through March 3, 2005.
The last biweekly notice was published on March 1, 2005 (70 FR 9986).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
[[Page 12744]]
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: February 2, 2005.
Description of amendment request: The amendment would revise Tables
3.1.1 and 4.1.1 of the Technical Specifications (TSs) to incorporate
the isolation trip setting and the instrumentation surveillance
requirements of the reactor water cleanup (RWCU) system high energy
line break (HELB) detection and isolation equipment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff's analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The equipment modification, which is the subject of the
proposed amendment, had been installed by the licensee in 1998 using
the provisions of 10 CFR 50.59, and the licensee had been performing
the surveillance requirement as is now proposed for this amendment. The
purpose of the modification was to
[[Page 12745]]
ensure that the RWCU system can be isolated on an HELB downstream of
the RWCU system isolation valves. The proposed addition of the RWCU
HELB detection/isolation equipment setpoints and surveillance
requirements to the TSs satisfies the 10 CFR 50.36 requirements for
limiting conditions for operation (LCOs) and surveillance requirements
(SRs) that should be included in the TSs. Thus, the proposed amendment
would not alter the physical design or operational procedures
associated with any plant structure, system, or component (i.e., the
RWCU system will be isolated by existing equipment, in case there is an
HELB, in the same way as before the amendment). Consequently, the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment does not lead to any changes in the
physical design, safety limits, or safety analysis assumptions
associated with the operation of the plant. The proposed amendment
would only add requirements to the TSs for the operability and
surveillance testing of the RWCU system HELB detection/isolation
equipment. Accordingly, the proposed amendment does not introduce any
new accident initiators, nor does it reduce or adversely affect the
capabilities of any plant structure or system in the performance of
their safety function. Therefore, the proposed amendment does not
create the possibility of a new or different kind of accident from any
accident previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. The proposed amendment will not affect
any margin of safety as defined in the Oyster Creek Nuclear Generating
Station Final Safety Analysis Report. The amendment only adds LCOs and
SRs to assure that the RWCU system HELB detection/isolation equipment
is operable under the plant operating conditions when an RWCU system
HELB is possible. The amendment does not change the RWCU system
isolation time as compared to original plant design. Therefore, the
proposed amendment does not involve a significant reduction in a margin
of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the proposed amendment involves no
significant hazards consideration.
Attorney for licensee: Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendment request: January 27, 2005.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) testing frequency for the
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.''
The proposed change would revise the test frequency of SR 3.1.4.2,
control rod scram time testing, from ``120 days cumulative operation in
MODE 1'' to ``200 days cumulative operation in MODE 1.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in licensing amendment applications in the Federal Register on August
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the
model NSHC determination in its application dated January 27, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The frequency
of surveillance testing is not an initiator of any accident
previously evaluated. The frequency of surveillance testing does not
affect the ability to mitigate any accident previously evaluated, as
the tested component is still required to be operable. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change does not result in any new or different modes of plant
operation. 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 proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change continues to test the control rod scram time to ensure the
assumptions in the safety analysis are protected. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Jr.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: February 14, 2005.
Description of amendment request: The proposed amendments would
revise Technical Specification Surveillance Requirement (SR) 3.3.7.1 to
extend the frequency of the channel functional test to once every 31
days to once every 92 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated:
The proposed LAR [license amendment request] extends the current
31 day surveillance frequency to a 92 day surveillance frequency.
The proposed LAR does not alter the method of operating or
configuration for any structure, system, or component. Extension of
the surveillance interval will not affect any accident analysis or
the plant safety system response to the accident. The extension of
the surveillance interval will not affect the ability of ES
[engineered safeguards] to actuate Engineered Safeguards Protective
System (ESPS) equipment. Therefore, the proposed LAR does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
[[Page 12746]]
The proposed change does not necessitate a change in parameters
governing plant operation. Consequently, the proposed LAR does not
alter the nature of events postulated in the UFSAR [Updated Final
Safety Analysis Report] nor does the LAR introduce any unique
precursor mechanisms. Therefore, the proposed amendment will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Involve a significant reduction in the margin of Safety
The proposed change does not adversely affect any plant safety
limits, setpoints, or design parameters. The changes will not
adversely affect the fuel, fuel cladding, RCS [reactor coolant
system], or containment integrity. The proposed change to the
frequency for SR [surveillance requirement] 3.3.7.1 will not impact
the operation of the ESPS Digital Automatic Actuation Logic Channels
nor the actuation of ESPS equipment. Additionally, the channel
functional testing of the ESPS Digital Channels will continue to be
performed within an acceptable timeframe following implementation of
the proposed change. As such, 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: Anne W. Cottingham, Winston and Strawn LPP,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 22, 2004.
Description of amendment request: The requested change would delete
Technical Specification (TS) 6.9.1.5, ``Occupational Radiation Exposure
Report,'' and 6.9.1.6, ``Monthly Operating Reports,'' as described in
the Notice of Availability published in the Federal Register on June
23, 2004 (69 FR 35067).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Allen G. Howe.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: January 27, 2005.
Description of amendment request: The proposed License Amendment
Request (LAR) would allow the licensee to utilize a probabilistic
methodology to determine the contribution to main streamline break
(MSLB) leakage rates for the once-through steam generator (OTSG) from
the tube end crack (TEC) alternate repair criteria (ARC) described in
Crystal River Unit 3 (CR-3) Improved Technical Specification (ITS)
5.6.2.10.2.f. This LAR involves no change to the CR-3 ITS.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
This LAR proposes to change the method to determine the
projected MSLB leakage rates for TEC. Potential leakage from OTSG
tubes, including leakage contribution from TEC, is bounded by the
MSLB evaluation presented in the Final Safety Analysis Report
(FSAR). The inspection required by the ARC will continue being
performed as required by CR-3 ITS 5.6.2.10. This inspection provides
continuous monitoring of tubes with TEC indications remaining in
service, and ensures that degradation of new tubes containing TEC
indications is detected. The proposed change in method to determine
MSLB leakage rates for TEC does not change any accident initiators.
Therefore, granting this LAR does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does not create the possibility of a new or different type of
accident from any accident previously evaluated.
This LAR proposes to change the method to determine the
projected MSLB leakage rates for TEC. The change introduces no new
failure modes or accident scenarios. The proposed change does not
change the assumptions made in Topical Report BAW-2346P, Revision 0,
which demonstrated structural and leakage integrity for all normal
operating and accident conditions for CR-3. The design and
operational characteristics of the OTSGs are not impacted by the use
of a probabilistic methodology to determine MSLB leakage rates.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does not involve a significant reduction in the margin of
safety.
This LAR proposes to change the method to determine the
projected MSLB leakage rates for TEC. The resulting leakage
estimates will be lower than the estimates from the old method.
However, the estimates from the proposed method will be more
realistic and do not impact the acceptance criteria. The methodology
relies on the same accident analyses described in Topical Report
BAW-2346P, Revision 0, and License Amendment Request 249,
Revision 0, and utilizes the same leakage test data and leakage
limit. The FSAR analyzed accident scenarios are not affected by the
change and remain bounding. The limits established in CR-3 ITS
3.4.12, and 5.6.2.10.2.f have not been changed. Therefore, the
proposed change does not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 12747]]
review, it appears that the three standards of 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Jr.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: January 20, 2005.
Description of amendment request: The submittal requests revision
to several Technical Specifications (TSs) using eight TS Task Force
(TSTF) generic changes. The eight TSTFs (nos. 5, 93, 95, 101, 258, 299,
308, and 361) delete redundant safety limit violation notification
requirements; extend the pressurizer heater surveillance frequency from
92 days to 18 months; extend the completion time for reducing the Power
Range High trip setpoint from 8 to 72 hours; change the auxiliary
feedwater pump test frequency to be consistent with the inservice test
program frequency; remove redundant requirements and add other
requirements to Section 5.0, Administrative Controls; clarify the
requirements regarding the frequency of testing for cumulative and
projected dose contributions from radioactive effluents; and add a note
to the residual heat removal requirements during Mode 6 low water level
operations that allows one required residual heat removal (RHR) loop to
be inoperable for up to 2 hours for surveillance testing provided the
other RHR loop is operable and in operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes revise administrative requirements,
actions, action times, surveillance requirements and surveillance
frequencies. The revised requirements are not an initiator of any
accident previously evaluated. As a result, the probability of any
accident previously evaluated is not significantly increased by the
proposed changes. The Technical Specifications continue to require
the systems, structures, and components associated with the revised
requirements to be operable. Therefore, any mitigation functions
assumed in the accident analyses will continue to be performed. As a
result, the consequences of any accident previously evaluated are
not significantly increased. Therefore, the proposed amendments do
not involve a significant increase in the probability or
consequences of any accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any previously evaluated.
The proposed changes do not alter the design or physical
configuration of the plant. No changes are being made to the plant
that would introduce any new accident causal mechanisms. The
proposed changes do not affect any other plant equipment. Therefore,
operation of the facility in accordance with the proposed amendments
does not create the possibility of a new or different kind of
accident from any previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes do not change the design or function of
plant equipment. The proposed changes do not significantly reduce
the level of assurance that any associated plant equipment will be
available to perform its function. The proposed changes provide
operating flexibility without significantly affecting plant
operation. Therefore, operation of the facility in accordance with
the proposed amendments would 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 50.92(copyright))
are satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Michael L. Marshall, Jr.
National Aeronautics and Space Administration, Docket No. 50-30, the
Plum Brook Test Reactor, Sandusky, Ohio
Date of amendment request: January 14, 2005.
Description of amendment request: The proposed amendment will
clarify the license requirements for confirmation of Final Status
Survey results prior to backfilling or covering of excavated areas. The
amendment will allow performance of the Final Status Survey for an area
that has been excavated and allow backfilling of the area without the
performance of confirmatory surveys by the NRC. Backfilling without the
performance of a confirmatory survey would be allowed only when the NRC
Staff has determined that there is appropriate safety or technical
justification for backfilling and that, based on the NRC Staff review
of the completed Final Status Survey for the affected area, there is
reasonable assurance that the Licensee's surveys have demonstrated that
the affected area satisfies the unrestricted release criteria.
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:
The proposed amendment to License TR-3 is necessary to assure
that, in limited situations, areas excavated for the performance of
dismantlement and remediation activities do not result in
unnecessary industrial safety hazards. The proposed changes do not
involve a significant hazard as shown in the following:
A. The proposed amendment to License TR-3 does not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
The accident scenarios applicable to the decommissioning of the
Plum Brook Reactor Facility are described in section 3.3 of the
Decommissioning Plan for the Plum Brook Reactor. The Decommissioning
Plan describes postulated events that could result in a release of
radioactive materials from the site and analyzes the radiation dose
consequences of these events and demonstrates that no adverse public
health and safety impacts are expected from these events.
Performance of Final Status Surveys is a continuation of
decommissioning and is an activity that involves measurements and
analysis of residual radioactivity in areas in which decommissioning
has already been performed. It is a process used to confirm that
radioactivity has been removed to achieve the acceptance criteria
specified in 10 CFR 20, Subpart E. These surveys are subjected to
NRC review, and in some instances NRC confirmatory surveys. These
surveys are performed in areas where other decommissioning
activities are already complete and where there is no credible event
that could initiate the analyzed accidents. Backfilling an excavated
area, or otherwise rendering it inaccessible, will have no impact on
other decommissioning activities, or on postulated accidents from
other decommissioning activities. Therefore, the proposed amendment
will have no affect on the probability or consequences of accidents
previously analyzed.
B. The proposed amendment to License TR-3 will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Accidents previously analyzed in the Decommissioning Plan assess
different
[[Page 12748]]
scenarios that could cause the dispersion of radioactive material to
the environment. These scenarios arise from dismantlement activities
associated with the decommissioning. Excavation of areas to support
dismantlement and remediation is an activity that is described in
the Decommissioning Plan. Excavated areas will be surveyed in
accordance with the Final Status Survey Plan to verify that they
have been radiologically remediated to the unrestricted release
criteria. These areas may be backfilled without the performance of
confirmatory surveys performed by the NRC or their Contractor.
However, this will only be permitted after NRC review of the
Licensee's survey results and a determination that there is
reasonable assurance that no residual radioactivity in excess of the
release criteria remains. Therefore, there will be adequate
verification by the regulatory agency that there is reasonable
assurance that there is no potential for the dispersion of
radioactive material to the environment from the backfilled area.
The methods and processes used for control of work activities and
for control and monitoring of radioactivity will remain the same as
those used prior to this amendment. Therefore, no new or different
types of accidents are created by this proposed amendment.
C. The proposed amendment to License TR-3 will not involve a
significant reduction in a margin of safety.
As discussed previously, the activities that will be performed
at the facility are as previously described and evaluated in the
accident analyses presented in the Decommissioning Plan. The
radiological criteria to be used in applying for termination of the
NRC Licenses will remain the same as originally proposed and are
consistent with the criteria of 10 CFR 20 Subpart E. The results of
Final Status Surveys performed by the Licensee will remain subject
to review by the U.S. NRC for adequate implementation of the Final
Status Survey Plan. Therefore, the margins of safety applicable to
assessing the long term dose to members of the public from exposure
to the facility after termination of the license remain unchanged.
In addition, since this amendment does not impact any previously
reviewed accident analyses as previously discussed, no margins of
safety are affected by this proposed amendment.
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 the Licensee: J. William Sikora, Esquire, 21000
Brookpark Road, Mail Stop 500-118, Cleveland, Ohio 44135.
NRC Section Chief: Patrick M. Madden.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: October 15, 2004.
Description of amendment request: The proposed amendment revises
Technical Specification 5.5.6, ``Reactor Coolant Pump Flywheel
Inspection Program,'' to extend the allowable inspection interval to 20
years.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
availability of a model safety evaluation and model no significant
hazards consideration (NSHC) determination for referencing in license
amendment applications in the Federal Register on October 22, 2003 (68
FR 60422). The licensee affirmed the applicability of the model NSHC
determination in its application dated October 15, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per
year). Moreover, considering the uncertainties involved in this
evaluation, the risk associated with the postulated failure of an
RCP motor flywheel is significantly low. Even if all four RCP motor
flywheels are considered in the bounding plant configuration case,
the risk is still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits; or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2--The proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or alter the methods governing normal plant operation. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements, and does not
alter any assumptions made in the safety analysis. The proposed
change is consistent with the safety analysis assumptions and
current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in a margin of safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: February 1, 2005.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.7.17, ``Spent Fuel Pool
Storage,'' Technical Specification 4.3, ``Fuel
[[Page 12749]]
Storage'' and the corresponding TS bases, using revised spent fuel pool
(SFP) criticality analysis methodology which takes credit for soluble
boron in the spent fuel pool.
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.
This license amendment proposes to revise the plant licensing
basis by: (1) Replacing the spent fuel pool criticality analyses;
and (2) revising the spent fuel storage Technical Specifications
3.7.17, ``Spent Fuel Pool Storage'' and 4.3, ``Fuel Storage''
utilizing the proposed analyses. The proposed Technical
Specification revisions allow spent fuel to be stored in different
configurations.
The proposed changes relate to prevention of criticality
accidents in the spent fuel pool. Since the current spent fuel pool
criticality analyses and Technical Specifications ensure that a
criticality accident does not occur, criticality accidents have not
been previously evaluated. Likewise the proposed spent fuel pool
criticality analyses and Technical Specifications ensure that a
criticality accident does not occur. Thus the proposed changes do
not involve a significant increase in the consequences of an
accident previously evaluated.
Events that could cause a criticality accident were evaluated
and analyses demonstrated that the current Technical Specification
required soluble boron is more than adequate to assure that a
criticality accident does not occur. Thus the proposed changes do
not involve a significant increase in the probability of an accident
previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in 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.
This license amendment proposes to revise the plant licensing
basis by: (1) Replacing the spent fuel pool criticality analyses;
and (2) revising the spent fuel storage Technical Specifications
3.7.17, ``Spent Fuel Pool Storage'' and 4.3, ``Fuel Storage''
utilizing the proposed analyses. The proposed Technical
Specification revisions allow spent fuel to be stored in different
configurations.
The proposed licensing basis changes do not involve a change in
system operation, or procedures involved with the fuel storage
system. It does revise the allowable storage configurations. The
proposed changes provide a conservative basis for evaluating spent
fuel pool criticality and storage of fuel assemblies in a safe
configuration which meets criticality evaluation acceptance
criteria. There are no new failure modes or mechanisms created
through use of the proposed analyses or proposed Technical
Specifications. Use of these licensing basis changes for storage of
fuel assemblies does not involve any modification in the operational
limits of plant systems. There are no new accident precursors
generated with use of these licensing basis 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.
This license amendment proposes to revise the plant licensing
basis by: (1) Replacing the spent fuel pool criticality analyses;
and (2) revising the spent fuel storage Technical Specifications
3.7.17, ``Spent Fuel Pool Storage'' and 4.3, ``Fuel Storage''
utilizing the proposed analyses. The proposed Technical
Specification revisions allow spent fuel to be stored in different
configurations.
The proposed licensing basis change will result in a
conservative calculation of the required spent fuel pool soluble
boron concentration for the proposed fuel storage configurations.
The current Technical Specification required spent fuel pool boron
concentration significantly exceeds the proposed criticality
analyses required boron concentration. The proposed analyses
demonstrate that the criticality analysis acceptance criteria for
the proposed fuel storage configurations are met. The proposed
analyses utilize industry accepted analysis codes which have been
benchmarked for the spent fuel pool criticality analyses proposed
for the Prairie Island Nuclear Generating Plant. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: October 1, 2004.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) to maintain
hydrogen recombiners and hydrogen and oxygen monitors. A notice of
availability for this TS improvement using the consolidated line item
improvement process was published in the Federal Register on September
25, 2003 (68 FR 55416). 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 in 1979. Requirements
related to combustible gas control were imposed by order for many
facilities and were added to, or included in, the TSs for nuclear power
reactors currently licensed to operate. The revised Title 10 of the
Code of Federal Regulations (10 CFR) Section 50.44, ``Standards for
Combustible Gas Control System in Light-Water-Cooled Power Reactors,''
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 Nuclear Regulatory Commission (NRC) staff issued a notice of
availability of a model no significant hazards consideration (NSHC)
determination for referencing 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 October 1, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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 NRC 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.
[[Page 12750]]
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 NRC found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents. Also, as part of the
rulemaking to revise 10 CFR 50.44, the NRC 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 TSs will not prevent an
accident management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and site survey monitoring that support modification of
emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TSs, 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 TSs, 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 TSs, 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 NRC 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 TSs will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: Darrell Roberts.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: February 4, 2005.
Description of amendment request: The proposed change will relocate
Technical Specification (TS) requirements regarding the Traversing In-
core Probe (TIP) System to the Hope Creek Updated Final Safety Analysis
Report (UFSAR). Additionally, the associated TS Bases would be deleted.
Formatting changes would also be made, as required, in order to
incorporate these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will relocate the requirements from the
Technical Specifications (TS) to the Hope Creek Updated Final Safety
Analysis Report (UFSAR) a licensee-controlled document. The
relocated requirements will be retained in licensee-controlled
documents, which will be maintained under the requirements of the
provisions of 10 CFR 50.59. Since any changes to licensee-controlled
documents are required to be evaluated per 10 CFR 50.59, no increase
in the probability or consequences of an accident previously
evaluated is allowed.
In addition, the proposed change will not affect any equipment
important to safety, in structure or operation. This change will not
alter operation of process variables[,] Structures, systems or
components as described in the UFSAR or licensing basis. Therefore,
the proposed change does not involve a significant increase in the
probability or radiological consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not change the design function or
operation of any plant equipment. No new failure mechanisms,
malfunctions, or accident initiators are being introduced by the
proposed changes. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not reduce the margin of safety since
they have no impact on safety analysis assumptions. Any future
changes to the TIP system requirements will be evaluated under
10CFR50.59. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
[[Page 12751]]
NRC Section Chief: Darrell J. Roberts.
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: October 6, 2004.
Brief description of amendments: The proposed amendments will (1)
Revise Technical Specification (TS) 3.8.1, ``AC Sources-Operating,'' to
allow surveillance requirement (SR) testing of the onsite standby
diesel generators (DGs) during power operation, by removing specific
surveillance test MODE restrictions, (2) incorporate changes based on
industry approval TSs Task Force (TSTF) standard TS change traveler,
TSTF-283, Revision 3, (3) add a new note to TS 3.8.1 Limiting Condition
for Operation (LCO) that permits one DG to be connected in parallel
with offsite power in order to conduct the required surveillance
testing, and (4) delete the expired TS LCO 3.8.1, Required Action A.3
one time 21 day completion time allowance for Startup Transformer XST2
preventive maintenance.
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 by focusing on the three standards set forth in 10 CFR
50.92. The licensee's analysis of the issue of no significant hazards
consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design of plant equipment is not being modified by the
proposed changes. In addition, the DGs and their associated
emergency loads are accident mitigating features. As such, testing
of the diesel generators (DGs) themselves is not associated with any
potential accident-initiating mechanism. Therefore, there will be no
significant impact on any accident probabilities by the approval of
the requested changes.
The changes include an increase in the online time that a DG
under test will be paralleled to the grid (for SRs 3.8.1.10 and
3.8.1.14) or unavailable due to testing (per SR 3.8.1.13). However,
the overall time that the DG is paralleled in all modes (outage /
non-outage) should remain unchanged. As such, the ability of the
tested DG to respond to a design basis accident [(DBA)] could be
adversely impacted by the proposed changes. However, the impacts are
not considered significant based, in part, on the ability of the
remaining DG to mitigate a DBA or provide safe shutdown. With regard
to SR 3.8.1.10 and SR 3.8.1.14, experience shows that testing per
these SRs typically does not perturb the electrical distribution
system and share[s] the same electrical configuration alignment as
the current monthly surveillance. In addition, operating experience
and qualitative evaluation of the probability of the DG or bus loads
being adversely affected concurrent with or due to a significant
grid disturbance, while the DG is being tested, support the
conclusion that the proposed changes do not involve any significant
increase in the likelihood of a safety-related bus blackout or
damage to plant loads.
The SR changes that are consistent with TSTF-283 have been
approved generically and for individual licensees. The on-line tests
allowed by the TSTF are only to be performed for the purpose of
establishing OPERABILITY. Performance of these SRs during restricted
MODES will require an assessment to assure plant safety is
maintained or enhanced.
Deletion of expired TS LCO 3.8.1 Required Action A.3 one time 21
day Completion Time allowance for Startup Transformer XST2
preventive maintenance is an administrative change only.
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 accident from any accident previously evaluated?
Response: No.
The proposed changes would not create any new accidents since no
changes are being made to the plant that would introduce any new
accident causal mechanisms. Equipment will be operated in the same
configuration as currently allowed for other DG SRs that allow
testing during at-power operation. Deletion of expired TS LCO 3.8.1
Required Action A.3 one time 21 day Completion Time allowance for
Startup Transformer XST2 preventive maintenance is an administrative
change only. This license amendment request does not impact any
plant systems that are accident initiators; neither does it
adversely impact any accident mitigating systems.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not involve a significant reduction in
the margin of safety. The margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The proposed changes do not
directly affect these barriers, nor do they involve any significant
adverse impact on the DGs which serve to support these barriers in
the event of an accident concurrent with a loss of offsite power.
The proposed changes to the testing requirements for the plant DGs
do not affect the OPERABILITY requirements for the DGs, as
verification of such OPERABILITY will continue to be performed as
required (except during different allowed MODES). The changes have
an insignificant impact on DG availability, as continued
verification of OPERABILITY supports the capability of the DGs to
perform their required function of providing emergency power to
plant equipment that supports or constitutes the fission product
barriers. Only one DG is to be tested at a time, so that the
remaining DG will be available to safely shut down the plant if
required. Consequently, performance of the fission product barriers
will not be impacted by implementation of the proposed amendment.
In addition, the proposed changes involve no changes to
setpoints or limits established or assumed by the accident analysis.
On this and the above basis, no safety margins will be impacted.
Deletion of expired TS LCO 3.8.1 Required Action A.3 one time 21
day Completion Time allowance for Startup Transformer XST2
preventive maintenance is an administrative change only.
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
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Allen G. Howe.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia; Docket Nos. 50-280 and 50-281, Surry Power Station, Units No.
1 and No. 2, Surry County, Virginia
Date of amendment request: August 30, 2004.
Description of amendment request: The proposed amendment revises
the Reactor Coolant Pump (RCP) Flywheel Inspection Programs to extend
the allowable inspection interval to 20 years.
The NRC staff issued a model safety evaluation and model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 24,
2003 (68 FR 37590). The notice of availability of the model application
was issued on October 22, 2003 (68 FR 60422). The licensee affirmed the
applicability of the model NSHC determination in its application dated
August 30, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an
[[Page 12752]]
analysis of the issue of no significant hazards consideration is
presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines [contained] in RG [Regulatory Guide] 1.174 (<1.0E-6 per
year). Moreover, considering the uncertainties involved in this
evaluation, the risk associated with the postulated failure of an
RCP motor flywheel is significantly low. Even if all four RCP motor
flywheels are considered in the bounding plant configuration case,
the risk is still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits; or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2--The proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or alter the methods governing normal plant operation. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements, and does not
alter any assumptions made in the safety analysis. The proposed
change is consistent with the safety analysis assumptions and
current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in a margin of safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: June 1, 2004, as supplemented July 23,
2004, and February 18, 2005.
Description of amendment request: These amendments would lower the
BVPS-2 overpressure protection system enable temperature, allow one
inoperable residual heat removal loop during surveillance testing,
remove the BVPS-1 list of figures and list of tables from the Index of
the BVPS-1 Technical Specifications (TSs), and make minor changes to
achieve consistency between the units and with the Standard TSs for
Westinghouse plants and with some TS Task Force changes.
Date of publication of individual notice in Federal Register :
February 25, 2005 (70 FR 9391).
Expiration date of individual notice: March 11, 2005, for comments;
April 26, 2005, for hearing.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management
[[Page 12753]]
Systems (ADAMS) Public Electronic Reading Room on the Internet at the
NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not
have access to ADAMS or if there are problems in accessing the
documents located in ADAMS, contact the PDR Reference staff at 1 (800)
397-4209, (301) 415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: May 20, 2004, as supplemented by
letter dated October 19, 2004.
Brief description of amendment: The amendment revised the Technical
Specifications, Section 3.2.B.4, regarding control rod operability
requirements for inoperable control rods, clarifying the application of
the action requirements for inoperable control rods.
Date of issuance: February 25, 2005.
Effective date: February 25, 2005, and shall be implemented within
30 days of issuance.
Amendment No.: 253.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 2004 (69 FR
40671).
The October 19, 2004, letter provided clarifying information within
the scope of the original application and did not change the staff's
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of this amendment is contained in a
Safety Evaluation dated February 25, 2005.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: July 15, 2004 as supplemented
on January 31, 2005.
Brief description of amendments: The amendment adds references to
the list of approved core operating limits analytical methods in
Technical Specification 5.6.5.b for Calvert Cliffs, Unit Nos. 1 and 2.
Date of issuance: February 24, 2005.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 271 and 248.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 29, 2004 (69
FR 78056). The supplement dated January 31, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of these amendments is contained in
a Safety Evaluation dated February 24, 2005.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: December 19, 2003, as
supplemented by letter dated January 14, 2004.
Brief description of amendment: The amendment modified Technical
Specifications to adopt the provisions of Industry/TS Task Force (TSTF)
change TSTF-359, ``Increased Flexibility in Mode Restraints.'' The
availability of TSTF-359 for adoption by licensees was announced in the
Federal Register on April 4, 2003 (68 FR 16579).
Date of issuance: February 14, 2005.
Effective date: As of the date if issuance and shall be implemented
within 180 days.
Amendment No.: 203.
Renewed Facility Operating License No. DPR-23: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7519).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 14, 2005.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: September 21, 2004.
Brief description of amendment: This amendment deletes the
technical specification requirements associated with hydrogen
recombiners, and hydrogen and oxygen monitors.
Date of issuance: March 3, 2005.
Effective date: March 3, 2005, and shall be implemented within 120
days from the date of issuance.
Amendment No.: 189.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 21, 2004 (69
FR 76488).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 3, 2005.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: June 28, 2004.
Description of amendment request: The amendment revised the
Seabrook Station, Unit No. 1 Technical Specifications (TSs) by
eliminating the requirements associated with hydrogen recombiners and
hydrogen monitors.
Date of issuance: February 22, 2005.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 99.
Facility Operating License No. NPF-86: The amendment revised the
TSs.
Date of initial notice in Federal Register: September 14, 2004 (69
FR 55471).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 22, 2005.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: October 6, 2003, as supplemented by
letters dated May 5, May 24, July 8, September 13, 2004, and January
13, 2005.
Description of amendment request: The amendment revised the
Seabrook Station, Unit No. 1 licensing basis to incorporate a full-
scope application of an alternative source term methodology in
accordance with 10 CFR 50.67.
Date of issuance: February 24, 2005.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 100.
Facility Operating License No. NPF-86: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68670).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 2005.
No significant hazards consideration comments received: No.
[[Page 12754]]
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 17, 2004, as supplemented by
letters dated March 17, 2004, April 1, 2004, May 26, 2004, September
13, 2004 (2 letters), October 12, 2004, October 28, 2004, December 3,
2004, December 28, 2004, and January 28, 2005.
Description of amendment request: The amendment revised the
Seabrook Station, Unit No. 1 operating license to increase the licensed
rated power by 5.2 percent from 3411 megawatts thermal to 3587
megawatts thermal.
Date of issuance: February 28, 2005.
Effective date: As of its date of issuance, and shall be
implemented within 12 months.
Amendment No.: 101.
Facility Operating License No. NPF-86: The amendment revised the
Technical Specifications and the operating license.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34701).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 28, 2005.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 25, 2004, as supplemented by
letters dated December 29, 2004, and January 26, 2005.
Brief description of amendment: The amendment revises Technical
Specification 2.1.1.2 for the dual recirculation loop and single
recirculation loop Safety Limit Minimum Critical Power Ratio values to
reflect results of a cycle-specific calculation.
Date of issuance: February 1, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 210.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 23, 2004 (69
FR 68183).
The supplements dated December 29, 2004, and January 26, 2005,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 1, 2005.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: October 15, 2004.
Brief description of amendment: The amendment revised Section 1.7,
which defines the term ``Instrument Channel Calibration,'' by adding
two new sentences pertaining to calibration of resistance temperature
detector or thermocouple sensors.
Date of issuance: February 17, 2005.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 187.
Facility Operating License No. DPR-63: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70719).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 17, 2005.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: September 7, 2004.
Brief description of amendment: The amendment deleted the technical
specification requirements to submit a monthly operating report and an
annual occupational radiation exposure report.
Date of issuance: February 18, 2005.
Effective date: February 18, 2005, and shall be implemented within
60 days from the date of issuance.
Amendment No.: 231.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62477).
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated February 18, 2005.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 21, 2004, as supplemented on October
29, and December 16, 2004.
Brief description of amendment: The amendment revised Technical
Specifications (TSs) Section 2.3(4), ``Emergency Core Cooling System--
Trisodium Phosphate (TSP),'' regarding the volume and the form of the
TSP; and TS Section 3.6(2)d.(i), ``Safety Injection and Containment
Cooling Systems Tests,'' the surveillance requirement for the TSP
volume.
Date of issuance: March 1, 2005.
Effective date: March 1, 2005, and shall be implemented within 60
days from the date of issuance.
Amendment No.: 232.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: The October 29, and
December 16, 2004, supplemental letters provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated March 1, 2005.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2, (SSES-2) Luzerne County, Pennsylvania
Date of application for amendment: September 8, 2004, as
supplemented by letter dated February 1 and 14, 2005.
Brief description of amendments: The amendment changed the SSES-2
Technical Specifications by revising the Unit 2 Cycle 13 Minimum
Critical Power Ratio Safety Limits in Section 2.1.1.2 and the
references listed in Section 5.6.5.b.
Date of issuance: February 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 194.
Facility Operating License No. NPF-22: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 2005 (69 FR
698). The supplements dated February 1 and 14, 2005, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed,
[[Page 12755]]
and did not change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 28, 2005.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: May 21, 2003, as supplemented
December 1, 2003 (two letters), February 16, March 1 and 8, April 22,
May 21, July 8 and 14, August 6 and 18, September 10, October 14 and
18, December 3 and 6, 2004, and January 27, 2005.
Brief description of amendment: The amendment revises the Technical
Specifications to reflect design modifications to the Control Room
Emergency Air Treatment System and elimination of the requirements for
the Containment Post Accident Charcoal Filters. The amendment also
changes the source term for the Dose Calculation Methodology to the
Alternate Source Term, and revises both the reactor coolant dose
equivalent I-131 specific activity limit and the containment spray NaOH
concentration limit.
Date of issuance: February 25, 2005.
Effective date: As of the date of issuance to be implemented upon
completion of the installation and testing of the new Control Room
Emergency Air Treatment System.
Amendment No.: 87.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40718). The supplemental letters provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 25, 2005.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: March 1, 2004.
Brief description of amendment: The amendment modifies Technical
Specification requirements to adopt the provisions of Industry/TS Task
Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: March 1, 2005.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 88.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications and/or License.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34706).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 1, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: May 21, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications to delete the requirements to maintain
hydrogen recombiners and hydrogen monitors and oxygen monitors.
Date of issuance: February 23, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 244 and 188.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57992).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 23, 2005.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: September 10, 2004.
Brief description of amendments: The amendments delete the
technical specification requirements to submit monthly operating
reports and annual occupational radiation exposure reports.
Date of issuance: March 2, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 115/115.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62479).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 2, 2005.
No significant hazards consideration comments received: No.
Dated in Rockville, Maryland, this 4th day of March 2005.
For the Nuclear Regulatory Commission.
James E. Lyons,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 05-4792 Filed 3-14-05; 8:45 am]
BILLING CODE 7590-01-P