[Federal Register Volume 68, Number 23 (Tuesday, February 4, 2003)]
[Notices]
[Pages 5668-5687]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-2415]
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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 Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, January 10, 2003, through January 23, 2003.
The last biweekly notice was published on January 21, 2003 (68 FR
2796).
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.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing 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
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By March 6, 2003, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714,\1\ which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the
[[Page 5669]]
designated Atomic Safety and Licensing Board will issue a notice of a
hearing or an appropriate order.
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\1\ The most recent version of title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29,
2002.
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As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 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
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, 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 filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois; Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket
Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle
County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear
Power Station, Units 1 and 2, Rock Island County, Illinois; Docket Nos.
50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3,
York County, Pennsylvania; Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: December 20, 2002.
Description of amendment request: Nuclear Regulatory Commission
(NRC) Regulatory Issue Summary 2002-05: ``NRC Approval of Boiling Water
Reactor Pressure Vessel Integrated Surveillance Program,'' provides
guidance on implementing the boiling water reactor (BWR) reactor
pressure vessel integrated surveillance program (ISP). The amendment
will modify the Updated Safety Analysis Reports (USARs) by removing the
current facility reactor material surveillance capsule removal
schedules from the facility USARs and specifying that these facilities
will participate in an ISP developed by the BWR Vessel and Internals
Project (BWRVIP). In addition, the Limerick Station will remove the
current facility reactor material
[[Page 5670]]
specimen surveillance schedule from the Technical Specifications.
With the exception of Oyster Creek, the USARs of each of the listed
facilities contain a withdrawal schedule for the reactor pressure
vessel material specimens. For those facilities which are not scheduled
to remove a material specimen as part of the ISP (i.e., Clinton, Quad
Cities, and Limerick), the proposed amendment would remove these plant-
specific schedules from the facility USARs and substitute a description
of the facility's participation in the ISP. For those facilities which
are scheduled to remove a capsule as part of the ISP (i.e., Dresden,
LaSalle, and Peach Bottom), the proposed amendment would revise the
material specimen withdrawal schedule in accordance with the ISP.
Finally, for Oyster Creek, which is not scheduled to remove any further
material specimens, the proposed amendment would revise the USAR to
state that Oyster Creek will participate in the ISP.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change adopts an integrated surveillance program
(ISP) for reactor material specimen surveillances. The ISP ensures
that the reactor pressure vessel (RPV) will continue to meet all
applicable fracture toughness requirements. No physical changes to
the facilities will result from the proposed change. The initial
conditions and methodologies used in accident analyses remain
unchanged. The proposed change does not revise or alter the design
assumptions for systems or components used to mitigate the
consequences of accidents. Thus, accident analyses results are not
affected by this proposed change.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change adopts an ISP for reactor material specimen
surveillances. The ISP ensures that the RPV will continue to meet
all applicable fracture toughness requirements. No physical changes
to the facilities will result from the proposed change.
The proposed change does not affect the design or operation of
any system, structure, or component (SSC) in the plant. The safety
functions of the related SSCs are not changed in any manner, nor is
the reliability of any SSC reduced. The change does not affect the
manner by which the facility is operated and does not change any
facility, structure, system, or component.
No new or different type of equipment will be installed by this
proposed change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change has no impact on the margin of safety of any
Technical Specification. There is no impact on safety limits or
limiting safety system settings. The change does not affect any
plant safety parameters or setpoints. No physical or operational
changes to the facility will result from the proposed changes.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Deputy General Counsel
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 19, 2002, as supplemented by
letter dated December 19, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) by: (1) Modifying the wording
of the current Surveillance Requirements (SRs) 4.0.1 and 4.0.3 to be
consistent with NUREG-1431, Revision 2, Improved Standard Technical
Specifications (ISTS) wording for SR 3.0.1 and SR 3.0.3; and (2)
modifying the ISTS wording, adopted in item 1 above, to allow a delay
period of 24 hours or up to the surveillance frequency interval,
whichever is greater, and to require a risk analysis to be performed
for any surveillance greater than 24 hours consistent with Technical
Specification Task Force (TSTF)-358 for missed surveillances.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the Consolidated Line Item Improvement Process (CLIIP). The NRC
staff subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). Entergy Operations Inc. reviewed
the following proposed NSHC determination published in the Federal
Register as part of the CLIIP for TSTF-358, and concluded in its
application of August 19, 2002, that the proposed NSHC determination
applied to Waterford Steam Electric Station, Unit 3.
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:
Adoption of TSTF-358, Revision 6--Missed Surveillances
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
[[Page 5671]]
3. The proposed change does not involve a significant reduction
in the margin of safety.
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function. Therefore, this change does
not involve a significant reduction in a margin of safety.
Proposed Changes to SR 4.0.1 and 4.0.3
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration for the
adoption of NUREG-1431, Revision 2, for the revised SR 4.0.1 and 4.0.3
wording. The NRC staff has reviewed the licensee's analysis against the
standards of 10 CFR 50.92(c). The NRC staff's review is presented
below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change involves rewording of the existing SRs 4.0.1
and 4.0.3 to be consistent with NUREG-1431, Revision 2. These
modifications involve no technical changes to the existing TS. This
change is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accident or transient events.
Therefore, this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed change involves the rewording of the existing SR 4.0.1
and 4.0.3 to be consistent with NUREG-1431, Revision 2. The change does
not involve a physical alteration of the plant (no new or different
type of equipment installed) or changes in the methods governing normal
plant operation. The change will not impose any new or different
requirements or eliminate any existing requirements. Therefore, the
proposed change does not create the probability of a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
the margin of safety.
The proposed change involves rewording of the existing SRs 4.0.1
and 4.0.3 to be consistent with NUREG-1431, Revision 2. The change is
administrative in nature and will not involve any technical changes.
The change will not reduce a margin of safety because it has no impact
on any safety analysis assumptions. Since this change is administrative
in nature, no question of safety is involved. Therefore, the proposed
change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 16, 2002.
Description of amendment request: The proposed amendment will
revise the current main steam isolation valve (MSIV) Technical
Specification (TS) 3/4 7.1.5 to more closely reflect TS 3.7.2 contained
in NUREG-1432, Revision 2. In addition, this change will remove the
MSIVs from the scope of containment isolation valve (CIV) TS 3/4 6.3
such that only TS 3/4.7.1.5 will apply to the MSIVs. These changes will
provide increased flexibility and clarity regarding the implementation
of the TSs regarding MSIVs.
Basis for proposed no significant hazard 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 to the applicability for the main steam line
isolation valves will not require operability when all MSIVs are
closed in Modes 2, 3, and 4. Analyzed events are assumed to be
initiated by the failure of plant structures, systems or components.
In the closed position the MSIVs are already in their safety
function position. In this position, there can be no increase in the
probability or consequences of an accident.
The consequences of previously analyzed events are dependent on
the initial conditions assumed for the analysis, and the
availability and successful functioning of the equipment assumed to
operate in response to the analyzed event. When the MSIVs are closed
in Modes 2, 3, and 4 they are performing their design function for
containment isolation and for main steam line isolation on the
secondary side of the plant. The proposed change does not alter the
initial conditions assumed in the safety analyses. The plant
parameters assumed for the analyses are maintained within assumed
limits through compliance with the Technical Specifications and
plant procedures. Additionally, the proposed change does not impose
any new safety analyses limits. Therefore, the proposed change does
not involve a significant increase in the consequences of an
accident previously evaluated.
The proposed change increases the allowed outage time for an
inoperable MSIV from 4 hours to 8 hours in Mode 1 and for Modes 2,
3, and 4; will allow both MSIVs to be inoperable, will allow
separate action entry for the inoperable valves, and will allow 8
hours to close each inoperable valve. Analyzed events are assumed to
be initiated by the failure of plant structures, systems or
components. Extending the time available to complete repairs of an
inoperable component does not have a detrimental impact on the
integrity of plant components nor does it increase the probability
that these components will fail. The proposed changes are not
related in any way to the probability of failure of a plant
structure, system or component which would result in the occurrence
of an analyzed event. Because the probability of failure of plant
equipment is not affected, there is no impact on the probability of
occurrence of a previously analyzed accident.
The consequences of previously analyzed events are dependent on
the initial conditions assumed for the analysis, and the
availability and successful functioning of the equipment assumed to
operate in response to the analyzed event. The steam line break
analysis in FSAR [Final Safety Analysis Report] Section 15.1.3
assumes a failure of one MSIV to close. For the containment
isolation function, in the event of an inoperable MSIV coincident
with a LOCA [loss-of-coolant accident], the closed system (i.e., the
steam generator tubes and main steam line piping) remains intact.
The closed system is subjected to a Type A containment leakage test,
is missile protected, and [has] seismic category I piping, and
typically has flow through it during normal operation such
[[Page 5672]]
that any loss of integrity could be continually observed through
leakage detection systems within containment and system walkdowns
outside containment. Therefore, with an inoperable MSIV the safety
analysis (both LOCA and steam line break) remains valid assuming no
additional failures. The increase in core damage frequency and large
early release fraction, resulting from the increased restoration
time, is negligible. The proposed 8 hour Allowed Outage Time is
sufficiently short to ensure that the MSIVs are operable when
required to perform their design function. Even though both MSIVs
will be allowed under separate condition entry, to be inoperable in
Modes 2, 3, and 4 the inoperable valves are still required to be
closed. The 8 hour Allowed Outage Time to close an inoperable valve
is based on the small likelihood of an accident occurring that will
need the MSIV isolation function during this time period and the
fact that the valves are located on a closed system with respect to
containment integrity. The proposed change does not alter the
initial conditions assumed in the safety analyses. The plant
parameters assumed for the analyses are maintained within assumed
limits through compliance with the Technical Specifications and
plant procedures. Additionally, the proposed change does not impose
any new safety analyses limits. Therefore, the proposed change does
not involve a significant increase in the consequences of an
accident previously evaluated.
The proposed change will add a Note to the MSIV surveillance to
allow entry into Mode 3 for testing at hot conditions. Analyzed
events are assumed to be initiated by the failure of plant
structures, systems or components. The addition of this allowance
for testing is not related in any way to the probability of failure
of a plant structure, system or component which would result in the
occurrence of an analyzed event. Because the probability of failure
of plant equipment is not affected, there is no impact on the
probability of occurrence of a previously analyzed accident.
The consequences of previously analyzed events are dependent on
the initial conditions assumed for the analysis, and the
availability and successful functioning of the equipment assumed to
operate in response to the analyzed event. The proposed change will
allow entry into Mode 3 in order to perform MSIV testing at hot
conditions. However, prior to this testing, the MSIVs are not known
to be inoperable from any other cause other than not having
performed the Surveillance Requirement to demonstrate closure times
at hot plant conditions, which they are expected to pass. The
proposed change will allow entry into Mode 3 for the condition where
both MSIVs may require closure time testing. This testing allowance
is limited to Mode 3, and must be completed prior to entry into
Modes 1 or 2. The proposed change does not alter the initial
conditions assumed in the safety analyses. The plant parameters
assumed for the analyses are maintained within assumed limits
through compliance with the Technical Specifications and plant
procedures. Additionally, the proposed change does not impose any
new safety analyses limits. Therefore, the proposed change does not
involve a significant increase in the consequences of an accident
previously evaluated.
The proposed change will require MSIVs, that are closed in
accordance with the Mode 2, 3, and 4 Action, be verified closed once
per seven days. Analyzed events are assumed to be initiated by the
failure of plant structures, systems or components. The addition of
this requirement is not related in any way to the probability of
failure of a plant structure, system or component which would result
in the occurrence of an analyzed event. Because the probability of
failure of plant equipment is not affected, there is no impact on
the probability of occurrence of a previously analyzed accident.
The consequences of previously analyzed events are dependent on
the initial conditions assumed for the analysis, and the
availability and successful functioning of the equipment assumed to
operate in response to the analyzed event. The proposed change adds
a Surveillance Requirement to Technical Specification 3/4.7.1.5 to
verify proper MSIV isolation on an actuation signal. This is not a
new Surveillance Requirement for the Technical Specifications.
Technical Specification 3.3.2, Engineering Safety Features Actuation
System Instrumentation, Surveillance Requirement 4.3.2.1 (Table 4.3-
2 Item 4.d) requires a functional test of the actuation relay (K305)
once per 18 months which verifies automatic closure of the MSIVs on
a simulated main steam isolation signal. The proposed change does
not alter the initial conditions assumed in the safety analyses. The
plant parameters assumed for the analyses are maintained within
assumed limits through compliance with the Technical Specifications
and plant procedures. Additionally, the proposed change does not
impose any new safety analyses limits. Therefore, the proposed
change does not involve a significant increase in the consequences
of an accident previously evaluated.
Therefore, none of the proposed change[s] described above
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. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
is no change being made to the parameters within which the plant is
operated, or to the setpoints at which protective or mitigative
actions are initiated. No alteration in the procedures which ensure
the plant remains within analyzed limits is being proposed, and no
change is being made to the procedures relied upon to respond to an
off-normal event. As such, no new failure modes are being
introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
limitations on operating parameters, and the setpoints at which
automatic actions are initiated. No equipment design features are
impacted by this change, no operating parameters are revised, and no
changes to the actuation setpoints are involved.
The design safety function of the MSIVs is to close upon receipt
of a main steam isolation signal. With the MSIVs already closed in
Modes 2, 3 or 4, the design function is satisfied.
The proposed change will increase the allowed outage time from 4
hours to 8 hours in Mode 1, for an inoperable MSIV. The proposed
change will also relax current allowances for MSIVs in Modes 2, 3,
and 4; however, the relaxations are in lower modes of operation
where the potential for an accident that would require the MSIV
isolation function is reduced. The proposed changes will still
ensure that the inoperable MSIV(s) are restored or closed in a
reasonable time of 8 hours. Once closed, the MSIVs meet their design
safety function.
The proposed change will add a note indicating the Surveillance
Requirements must be performed prior to entry into Modes 1 or 2. The
MSIVs are expected to pass the Surveillance Requirement and are not
known to be inoperable for any other reason than not having
performed the valve closure test at hot conditions. The testing is
limited to Mode 3, when the reactor is subcritical, thus verifying
the MSIV closure times prior to power operation.
The proposed change will require MSIVs, which are closed in
accordance with the Mode 2, 3, and 4 Action, be verified closed once
per seven days. This requirement provides additional assurance that
the MSIVs perform their design safety function to close.
The proposed change adds a Surveillance Requirement to Technical
Specification 3/4.7.1.5 to verify proper MSIV isolation on an
actuation signal. This, however, is not a new Surveillance
Requirement for the Technical Specifications. Technical
Specification 3.3.2, Engineering Safety Features Actuation System
Instrumentation, Surveillance Requirement 4.3.2.1 (Table 4.3-2 Item
4.d) requires a functional test of the actuation relay (K305) once
per 18 months which verifies automatic closure of the MSIVs on a
simulated main steam isolation signal.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street, NW., Washington, DC 20005-3502.
[[Page 5673]]
NRC Section Chief: Robert A. Gramm.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 16, 2002.
Description of amendment request: The proposed amendment will add
the topical report entitled ``Fuel Rod Maximum Allowable Gas
Pressure,'' CEN-372-P-A, to the list of analytical methods in Technical
Specification (TS) 6.9.1.11.1 used to determine the Waterford Steam
Electric Station, Unit 3 (Waterford 3) core operating limits.
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 analyzed?
Response: No.
The proposed change does not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident. The proposed
change adds an NRC [Nuclear Regulatory Commission]-approved topical
report to the list of analytical methods used to determine the core
operating limits. The effect of the addition of this new reference
is to revise the fuel design criterion for internal rod pressure to
accept rod pressures that may exceed nominal Reactor Coolant System
operating pressure. The use of this revised criterion continues to
ensure that the consequences of an accident remain within acceptable
limits. The change also proposes the administrative deletion of
report date and revision levels in the list of references. These
changes do not alter any of the assumptions or bounding conditions
currently in the Final Safety Analysis Report.
Waterford 3 performed a large break loss-of-coolant accident
(LOCA) analysis using bounding fuel performance data as described in
CEN-372-P-A. This analysis concluded that the peak cladding
temperature remained within 10 CFR 50.46 limits.
In addition to the LOCA analysis, an evaluation of the potential
for departure from nucleate boiling (DNB) propagation was performed
as described in CEN-372-P-A. The results confirmed that Waterford 3
is bounded by the results evaluated in the topical report and that
DNB propagation will not occur.
Based on these analyses, there is no 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 any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident. Accordingly, no
new failure modes have been defined for any plant system or
component important to safety nor has any new limiting failure been
identified as a result of the proposed change. The intent of the
proposed change is to reference an NRC-approved topical report in
the Technical Specifications. The topical report justifies an
acceptance criterion that allows fuel rod internal pressure to
exceed RCS [reactor coolant system] pressure. There are no new
accidents created by this change. An administrative aspect of this
change, the deletion of date and revision levels, was also
considered and does not create a new or different accident.
The impact of fuel rod internal pressure exceeding reactor
coolant system (RCS) pressure was considered in both an emergency
core cooling system (ECCS) performance analysis and in a DNB
propagation evaluation performed for Waterford 3. These two aspects
were required considerations based on the NRC Safety Evaluation
review of the topical report. The results demonstrated that
Waterford 3 continues to meet 10 CFR 50.46 and that there is no
potential for DNB propagation.
Based on these analyses, there is no possibility of the creation
of a new or different kind of accident from those previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change adds an NRC-approved topical report to the
list of analytical methods used to determine core operating limits.
It also deletes the revision number and dates associated with each
of the topical reports listed. The effect of the addition of the new
reference is to revise the fuel design criterion for fuel rod
internal pressure to accept rod pressures that may exceed nominal
RCS operating pressure. The use of this revised criterion continues
to ensure that the consequences of an accident remain within
acceptable limits. Since the core operating limits will continue to
be established by an NRC-approved methodology and the results will
be verified to meet the established acceptance criteria of 10 CFR
50.46, the change will provide adequate core protection. Thus, the
proposed amendment does not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 20, 2002.
Description of amendment request: The proposed amendment makes
several administrative changes to the Waterford Steam Electric Station,
Unit 3, Technical Specifications (TSs) to revise, delete, correct, or
clarify certain titles, page numbers, and heading information. The
proposed amendment also revises personnel and committee titles that
have been changed, revises administrative reporting requirements to
conform to 10 CFR 50.4, and deletes redundant or unnecessary
requirements from TSs 5.4, 6.6, and 6.7.
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 changes are primarily to correct titles, page
numbering errors, and otherwise make the TS index pages consistent
with other NRC [U. S. Nuclear Regulatory Commission] approved pages.
These changes are all of an administrative nature and have no effect
on any plant equipment or structures. Therefore, these changes do
not increase the probability or consequences of an accident
previously evaluated.
The proposed amendment also deletes TS 5.4.1 and 5.4.2. Values
for RCS [Reactor Coolant System] design pressure, temperature, and
volume are contained in the Final Safety Analysis Report. Any
changes to these are controlled by 10 CFR 50.59. Therefore, removing
the section from the TS will not increase the probability or
consequences of previously evaluated accidents.
The proposed amendment also deletes TS 6.6 and 6.7, and revises
TS 6.9.1 and TS 6.9.2 to administratively conform reporting
requirements to those in 10 CFR [part] 50. Therefore, removing these
sections from the TS will not increase the probability or
consequences of previously evaluated accidents.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are administrative in nature and do not
involve a physical alteration of the plant. No new or different
equipment or modes of operation are being introduced by this
proposed change. Thus, the changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
[[Page 5674]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
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 are primarily administrative in nature
and can not affect any safety barriers. The proposed change to TS
5.4 only deletes unnecessary information. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: December 6, 2002.
Description of amendment request: The proposed amendment would
increase the surveillance interval of the Local Power Range Monitor
(LPRM) calibrations from 1000 megawatt-days/ton to 2000 megawatt-days/
ton.
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:
Operation of the JAF [James A. FitzPatrick] plant in accordance
with the proposed amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92 since it would not:
1. Involve an increase in the probability or consequences of an
accident previously evaluated. The revised surveillance interval
continues to ensure that the LPRM signal is adequately calibrated.
The proposed change results in no change in radiological
consequences of the design basis LOCA [loss-of-coolant accident] as
currently analyzed for JAF. This change will not alter the basic
operation of process variables, structures, systems, or components
as described in the JAF UFSAR [Updated Final Safety Analysis
Report], and no new equipment is introduced by the change in LPRM
surveillance interval. The performance of the APRM [Average Power
Range Monitor] and RBM [Rod Block Monitor] systems are not
significantly affected by the proposed LPRM surveillance interval
increase. Therefore, the probability of accidents previously
evaluated is unchanged.
The consequences of an accident can be affected by the thermal
limits existing at the time of the postulated accident, but LPRM
chamber exposure has no significant effect on the calculated thermal
limits because LPRM accuracy does not significantly deviate with
exposure. For the extended calibration interval, the total nodal
power uncertainty remains less than the uncertainty assumed in the
thermal analysis basis safety limit, maintaining the accuracy of the
thermal limit calculation. Therefore, the thermal limit calculation
is not significantly affected by LPRM calibration frequency, and the
consequences of an accident previously evaluated are unchanged.
The change does not affect the initiation of any event, nor does
it negatively impact the mitigation of any event. Therefore, the
proposed change 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 accident previously evaluated. The proposed change will not
physically alter the plant or its mode of operation. The performance
of the APRM and RBM systems are not significantly affected by the
proposed LPRM surveillance interval increase. As such, no new or
different types of equipment will be installed, and the basic
operation of installed equipment is unchanged. The methods governing
plant operation and testing are consistent with current safety
analysis assumptions. Therefore, the proposed change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Involve a significant reduction in a margin of safety. The
proposed change has no impact on equipment design or fundamental
operation, and there are no changes being made to safety limits or
safety system allowable values that would adversely affect plant
safety as a result of the proposed change. The performance of the
APRM and RBM systems are not significantly affected by the proposed
LPRM surveillance interval increase. The margin of safety can be
affected by the thermal limits existing prior to an accident;
however, uncertainties associated with LPRM chamber exposure have no
significant effect on the calculated thermal limits. The thermal
limit calculation is not significantly affected because LPRM
sensitivity with exposure is well defined. LPRM accuracy remains
within the total nodal power uncertainty assumed in the thermal
analysis basis, thus maintaining thermal limits and the safety
margin.
Since the proposed change does not affect safety analysis
assumptions or initial conditions, the margin of safety in the
safety analyses are maintained. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: January 9, 2003.
Description of amendment request: The proposed Technical
Specification (TS) amendment request changes the definition of a Logic
System Functional Test, deletes the definition of a Simulated Automatic
Actuation, clarifies Surveillance Requirement 4.5.G.1.a regarding
simulated automatic actuation testing, and revises associated TS Bases.
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 change involves surveillance requirements and
definitions of surveillance tests. As such, the proposed change does
not involve any plant physical changes, change any Technical
Specification instrumentation setpoints, or introduce any new mode
of plant operation. The proposed change to surveillance requirements
and definitions does not result in any significant change in the
availability of logic systems or safety-related systems themselves.
Protective functions will be maintained. The proposed change does
not degrade plant design, operation, or the performance of any
safety system assumed to function in the accident analysis.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Create the possibility for a new or different kind of
accident from any previously evaluated.
The proposed change does not: introduce any new accident
initiators or failure mechanisms because the changes do not
introduce any new modes of plant operation, make any physical
changes (no new or different type of equipment will be installed);
or change any Technical Specification instrumentation setpoints or
methods of plant operation. The proposed changes will not
substantially impose new requirements or eliminate any existing
requirements.
Therefore, the changes to the surveillance requirements and
testing definitions that encompass this proposed change do not
[[Page 5675]]
create the possibility of a new or different kind of accident than
those previously evaluated.
3. 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. There is no change or impact on any safety
analysis assumptions. The proposed change does not involve any
increase in calculated off-site dose consequences. Operability of
protective instrumentation and the associated systems is unaffected,
and performance of equipment will not be significantly affected.
Since the proposed change is consistent with the BWR/4 Standard
Technical Specifications, NUREG-1433, Revision 2, approved by the
NRC [Nuclear Regulatory Commission] staff, revising the Technical
Specifications in a manner which clarifies and reflects the approved
level of detail ensures that safety margins are acceptable.
Therefore, there is no significant reduction in the margin of safety
as a result of this Technical Specification change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: March 14, 2002.
Description of amendment request: The proposed license amendment
request (LAR) will allow exercising and testing the Inclined Fuel
Transfer System (IFTS) prior to the beginning of the refueling outage,
thus increasing system reliability and refuel outage efficiency. The
proposed LAR does not provide for the movement of fuel. The proposed
LAR supplements Amendment No. 100 by including a time limit on the
removal of the IFTS blind flange, providing a requirement to install
the upper pool IFTS gate prior to IFTS blind flange removal, and
limiting the unbolted configuration on the IFTS blind flange when it is
rotated.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change permits removal of the Inclined Fuel
Transfer System (IFTS) blind flange for a maximum duration of 60
days per cycle when primary Containment operability is required in
MODES 1 (Power Operation), 2 (Startup), or 3 (Hot Shutdown). The
proposed change also limits the duration the IFTS blind flange may
be unbolted when in MODES 1, 2, or 3. The proposed change does not
involve modifications to plant systems or design parameters that
could contribute to the initiation of any accidents previously
evaluated.
Regarding the probability and consequences of design basis and
beyond design basis accidents, a comprehensive technical evaluation
was completed in accordance with Regulatory Guide (RG) 1.174, ``An
Approach for Using Probabilistic Risk Assessment In Risk-Informed
Decisions On Plant-Specific Changes to the Licensing Basis'' and RG
1.177, ``An Approach for Plant-Specific, Risk-Informed Decision
Making: Technical Specifications.'' This evaluation determined that
the proposed change is technically justified and the associated risk
is insignificant.
The proposed change permits alteration of the containment
boundary for the IFTS penetration. Regarding the consequences of
accidents, the proposed change has been determined via a
probabilistic risk assessment to be acceptable regarding its overall
impact to the plant's risk, consistent with the Nuclear Regulatory
Commission's Safety Goal Policy Statement. The resulting pressures
and temperatures from a design basis Loss Of Coolant Accident (LOCA)
are considered the primary challenge to the integrity of the
containment. Pursuant to Amendment 100, the existing Technical
Specifications require maintaining an adequate water seal to prevent
leakage from the bottom of the IFTS transfer tube and isolating the
drain piping. This water seal is adequate to mitigate the effects of
the design basis peak post-accident pressures and temperatures. The
proposed change requires the installation of the upper IFTS pool
gate to provide protection of the Suppression Pool Make Up system
water inventory. A time limit for IFTS blind flange removal of 60
days per cycle and a 20 hour limit for the unbolted configuration of
the IFTS flange have been established as conservative measures to
limit the associated risk to the containment boundary for all
accident conditions. The proposed change has been found to be
acceptable regarding flooding and seismic design issues.
Therefore, the function of the containment to provide an
adequate boundary in the event of a design basis LOCA is not
compromised with the proposed change and the proposed change does
not result in a significant increase in the probability of the
consequences of previously evaluated accidents.
2. The proposed changes would not create the possibility of a
new or different kind of accident from any previously analyzed.
The proposed change consists of the removal of the IFTS blind
flange when in MODES 1, 2, or 3. The IFTS blind flange is a passive
component that is not part of the primary reactor coolant pressure
boundary and is not involved in the operation or shutdown of the
reactor. Being passive, its presence or absence does not affect any
of the parameters or conditions that could contribute to the
initiation of any incidents or accidents that are created from a
loss of coolant or positive reactivity incident. Re-aligning the
boundary of the primary containment to include portions of the IFTS
is passive in nature and therefore has no influence on the
possibility of creating a new or different kind of accident.
Furthermore, operation of the IFTS is unrelated to the operation of
the reactor and there is no mishap in the process that can lead or
contribute to the possibility of losing any coolant in the reactor
or introducing the chance for positive or negative reactivity or
other accidents different from and not bounded by those previously
evaluated.
Therefore, the proposed change does not result in creating the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed change involves the re-alignment of the primary
containment boundary by removing the IFTS blind flange, which is a
passive component. The margin of safety that has the potential of
being impacted by the proposed change involves the dose consequences
of postulated accidents, which are directly related to potential
leakage through the primary containment boundary. The potential
leakage pathways due to the proposed change have been reviewed, and
leakage can only occur from the administratively controlled IFTS
transfer tube drain piping. Pursuant to Amendment 100, an individual
is currently designated to provide timely isolation of this drain
piping when this proposed change is in effect. The conservatively
calculated dose, which might be received by the designated
individual while isolating the drain piping, is well within the
guidelines of General Design Criterion 19. Furthermore, the drain
piping isolation valve is included in the Primary Containment
Leakage Rate Testing Program to ensure that leakage from the piping
and components located outboard of the blind flange will be
maintained consistent with the leakage rate assumptions of the
accident analysis. It has been determined that the proposed change
would not have a substantial impact on the ultimate pressure
capacity of the containment as it relates to the Large Early Release
Frequency (LERF) nor would it have a substantial impact on LERF from
seismic events. Therefore, the dose consequences of an event would
be unchanged, and the associated margin of safety would also be
unchanged.
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
[[Page 5676]]
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: March 14, 2002.
Description of amendment request: The amendment request proposes a
one-time exception to the requirement in Nuclear Energy Institute (NEI)
94-01 to perform an integrated leak rate test (ILRT) at a frequency of
10 years. The exception is to allow ILRT testing within 15 years from
the last ILRT, completed July 1, 1994. The proposed amendment is
considered risk-informed, therefore Regulatory Guide 1.174, ``An
approach for Using Probabilistic Risk Assessment in Risk-Informed
Decisions on Plant-Specific Changes to the Licensing Basis,'' has been
followed, while using the methodology of Electric Power Research
Institute (EPRI) report, ``Risk Impact Assessment of Revised
Containment Leak Rate Testing Intervals,'' (EPRI TR-104285).
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. This proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed extension to Type A testing cannot increase the
probability of an accident previously evaluated since extension of
the containment Type A testing is not a physical plant modification
that could alter the probability of accident occurence, nor is it an
activity or modification that could lead to equipment failure or
accident initiation.
The proposed extension to Type A testing does not result in a
significant increase in the consequences of an accident as
documented in NUREG-1493. The NUREG notes that very few potential
containment leakage paths are not identified by Type B and C tests.
It concludes that reducing Type A (ILRT) testing frequency to once
per twenty years leads to an imperceptible increase in risk.
Other testing and inspections provide a high degree of assurance
that the containment will not degrade in a manner detectable only by
Type A testing. The last three Type A tests performed at PPNP
identified containment leakage within the acceptable criteria,
indicating a very leak-tight containment. Inspections required by
the ASME Code are performed in order to identify indications of
containment degradation that could affect leak-tightness.
Containment pressure is monitored each shift during plant operation
and would identify containment vessel shell leakage into the annulus
by a decrease in containment pressure. Type B and C testing,
required by Technical Specifications, identifies any containment
leakage from designed penetrations, such as from valves, that would
otherwise be detected by a Type A test. These factors establish that
an extension to the PPNP Type A test interval will not represent a
significant increase in the consequences of an accident.
Thus, the proposed amendment does not involve a significant
increase in the probability or consequences of a previously
evaluated accident.
2. This proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed revision to the Technical Specifications adds a
one-time extension to the current interval for Type A testing for
PPNP. The current test interval of ten years, based on past
performance, would be extended on a one-time basis to fifteen years
from the last Type A test. The proposed extension to Type A testing
does not create the possibility of a new or different type of
accident since there are no physical changes to the plant or changes
to the operation of the plant that could introduce a new failure.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. This proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed revision to the PPNP Technical Specifications adds
a one-time extension to the current interval for Type A testing. The
current test interval of ten years, based on past performance, would
be extended on a one-time basis to fifteen years from the last Type
A test. The proposed extension to Type A testing will not
significantly reduce the margin of safety. The NUREG-1493 generic
study of the effects of extending containment leakage testing found
that a 20-year interval in Type A testing resulted in an
imperceptible increase in risk to the public. NUREG-1493 found that,
generically, the design containment leakage rate contributes only
about 0.1 percent of the overall risk and that decreasing the Type A
testing frequency would have a minimal effect on this risk since 95%
of the Type A detectable leakage paths would already be detected by
Type B and C testing. Furthermore, for PPNP, monitoring containment
vessel pressure each shift during operation further reduces the risk
of any containment leakage path going undetected. The PPNP test and
inspection performance has satisfactorily demonstrated that the
containment remains very leak tight. The proposed change has no
effect on Core Damage Frequency (CDF). The change in Large Early
Release Frequency (LERF) was computed and found to be a ``very
small'' change in accordance with the guidelines of Regulatory Guide
1.174. The computed change in Conditional Containment Failure
Probability (CCFP) and offsite dose have also been evaluated and are
considered to be insignificant.
Therefore, the change does not involve a significant reduction
in a margin of safety.
Based on the above considerations, it is concluded that a
significant hazard would not be introduced as a result of this
proposed change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: October 11, 2002.
Description of amendment request: The proposed amendment would
revise Crystal River Unit 3 Improved Technical Specifications (ITS)
3.3.15 ``Reactor Building Purge Isolation-High Radiation;'' ITS Bases
3.7.15 ``Spent Fuel Assembly Storage;'' ITS 3.9.3 ``Containment
Penetrations;'' and ITS 3.9.6 ``Refueling Canal Water Level'' to
account for handling irradiated fuel within containment that has not
occupied part of a critical reactor core within the previous 72 hours.
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:
Crystal River Unit 3 (CR-3) proposes to revise Improved
Technical Specifications (ITS) 3.3.15, 3.9.3, 3.9.6, and Bases
3.7.15.
Florida Power Corporation (FPC) has determined that this license
amendment request does not involve a significant hazards
consideration as defined in 10 CFR 50.92 based on the following:
(1) Does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change does not increase the probability of a fuel
handling accident in that the proposed change deals with the results
of such an accident, not the cause of such an accident. The proposed
change does not increase the consequences of an accident previously
evaluated in that the CR-3 Alternate Source Term (AST) has been
[[Page 5677]]
approved by the NRC, and this proposed change implements that NRC
approval. The AST for the Fuel Handling Accident (FHA) takes no
credit for containment isolation nor for a filtered release.
(2) Does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed changes to the ITS do not affect nor create a
different type of fuel handling accident. The fuel handling accident
analyses assume that all of the iodine and noble gases that become
airborne, escape, and reach the exclusion area boundary and low
population zone with no credit taken for filtration, containment of
the source term, or for decay or deposition in the containment. The
proposed changes do not involve the addition or modification of
equipment nor do they alter the design of plant systems. The revised
operations are consistent with the fuel handling accident analyses.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
(3) Does not involve a significant reduction in margin of
safety.
The calculated doses to both the public and control room
operators are well within the limits given in 10 CFR 50.67. The
proposed changes do not alter the bases for assurance that safety-
related activities are performed correctly or the basis for any ITS
that is related to the establishment of or maintenance of a safety
margin.
The systems that have been included in the proposed change will
have administrative controls in place to assure that the systems are
available and can be promptly returned to operation to further
reduce dose consequences. These administrative controls will include
a single normal or contingency method to promptly close the
equipment hatch opening. This prompt method need not completely
block the hatch opening nor be capable of resisting pressure, but is
to enable the ventilation systems to draw the release from the
postulated FHA in the proper direction such that it can be
monitored. Therefore, operations of the facility in accordance with
the proposed amendment would not involve a significant reduction in
margin of safety.
Based on the above, FPC concludes that the proposed license
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: R. Alexander Glenn, Associate General
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St.
Petersburg, Florida 33733-4042.
NRC Section Chief: Allen G. Howe.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: December 19, 2002.
Description of amendment request: The proposed amendment would
revise Crystal River Unit 3 Improved Technical Specification 2.1.1,
``Reactor Core Safety Limits.'' The proposed change will permit the use
of the BHTP correlation, which is needed to utilize the Framatome ANP
high thermal performance (HTP) spacer grid design.
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:
FPC [Florida Power Corporation] has evaluated the proposed
License Amendment Request (LAR), which consists of the identified
Improved Technical Specification (ITS) change, against the criteria
of 10 CFR 50.92(c). The ITS change allows the use of the BHTP
Correlation for departure from nucleate boiling (DNB) calculations
of reload cores containing the Mark-B/HTP fuel design.
FPC has concluded that this proposed LAR does not involve a
significant hazards consideration. The following is a discussion of
how each of the criteria is satisfied.
(1) [Does not] [i]nvolve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed safety limit value ensures that fuel integrity will
be maintained during normal operations and anticipated operational
occurrences (AOOs), and that the design requirements will continue
to be met. The proposed methodology for the BHTP departure from
nucleate boiling (DNB) correlation will be generically reviewed and
approved by the NRC prior to its use by Crystal River Unit 3 (CR-3)
in mixed core reload analyses. The core operating limits will be
developed in accordance with the new methodology and any limitations
established by the NRC in its safety evaluation of the new
methodology. The proposed safety limit value does not affect the
performance of any equipment used to mitigate the consequences of an
analyzed accident. There is no impact on the source term or pathways
assumed in accidents previously evaluated. No analysis assumptions
are violated and there are no adverse effects on the factors that
contribute to offsite or onsite dose as the result of an accident.
Therefore, the safety limit value for the BHTP correlation will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) [Does not] [c]reate the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed safety limit value does not change the methods
governing normal plant operation, nor are the methods utilized to
respond to plant transients altered. The BHTP correlation is not an
accident/event initiator. No new initiating events or transients
result from the use of the BHTP correlation and the related safety
limit changes. Therefore, the safety limit value for the BHTP
correlation will not involve the possibility of a new or different
kind of accident from any previously evaluated.
(3) [Does not] [i]nvolve a significant reduction in a margin of
safety.
The proposed safety limit value has been established in
accordance with the methodology for the BHTP correlation, to ensure
that the applicable margin of safety is maintained (i.e., there is
at least 95% probability at a 95% confidence level that the hot fuel
rod in the core does not experience departure from nucleate boiling
(DNB)). The proposed methodology for the BHTP DNB correlation will
be generically reviewed and approved by the NRC prior to its use by
CR-3. The other reactor core safety limits will continue to be met
by analyzing the reload for the mixed core using NRC approved
methods, and incorporation of resultant operating limits into the
Core Operating Limits Report (COLR). Therefore, the safety limit
value for the BHTP correlation will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: R. Alexander Glenn, Associate General
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St.
Petersburg, Florida 33733-4042.
NRC Section Chief: Allen G. Howe.
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: December 20, 2002.
Description of amendment request: This proposed amendment provides
editorial and administrative changes to the Technical Specifications.
The changes correct typographical, spelling, numbering syntax, page
break, and font consistency errors as well as removing blank pages and
associated references. There are no substantive changes made in the
proposed amendment.
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. Will operation of the facility in accordance with this
proposed change
[[Page 5678]]
involve a significant increase in the probability or consequences of
an accident previously evaluated?
No. The proposed amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated because the proposed amendments are purely
administrative or editorial in nature. These amendments make no
substantive Technical Specification changes and do not affect any
assumptions contained in plant safety analyses, the physical design
and/or operation of the plant; and they do not affect Technical
Specifications that preserve safety analysis assumptions. Therefore,
the proposed changes do not affect the probability or consequences
of accidents previously analyzed.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
No. The use of the administratively changed Technical
Specifications does not create the possibility of a new or different
kind of accident from any previously evaluated, since the proposed
amendments will not change the physical plant or the modes of plant
operation defined in the facility operating license. No new failure
mode is introduced due to the administrative changes and
clarifications, since the proposed changes do not involve the
addition or modification of equipment, nor do they alter the design
or operation of affected plant systems, structures, or components.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
No. The operating limits and functional capabilities of the
affected systems, structures, and components are unchanged by the
proposed amendments. The changed Technical Specifications, which
correct administrative and editorial errors, and clarify existing
Technical Specification requirements, do not reduce any of the
margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Allen G. Howe.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 31, 2002.
Description of amendment request: The proposed amendment would
change the reactor vessel material surveillance program to incorporate
the Boiling Water Reactor Vessel and Internals Project (BWRVIP)
Integrated Surveillance Program (ISP) into the licensing basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
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.
Pressure-temperature (P/T) limits (CNS [Cooper Nuclear Station]
Technical Specifications Figures 3.4.9-1, 2, and 3) are imposed on
the reactor coolant system to ensure that adequate safety margins
against non-ductile or brittle fracture exist during normal
operation, anticipated operational occurrences, and system
hydrostatic tests. The P/T limits are based on the nil-ductility
reference temperature, RTNDT, as described in ASME
Section XI, Appendix G. Changes in the fracture toughness properties
of RPV [reactor pressure vessel] beltline materials, resulting from
the neutron irradiation and the thermal environment, are monitored
by a surveillance program in compliance with the requirements of 10
CFR 50 [title 10 of the Code of Federal Regulations part 50]
Appendix H. The effect of neutron fluence on the shift in the
RTNDT of RPV materials is predicted by methods given in
RG [Regulatory Guide] 1.99, Revision 2.
This change is not related to any accidents previously
evaluated. Rather, the reactor vessel surveillance program,
corresponding material evaluations, and adjustment of a plant's P/T
limits, as necessary, protect against the possibility of reactor
vessel brittle fracture. Monitoring, evaluation, and adjustment of
CNS P/T limits to ensure adequate margin exists to brittle fracture
will continue. This change only replaces a plant-specific monitoring
and evaluation program with an integrated industry program, the
BWRVIP ISP. The NRC has reviewed this program and approved it for
implementation in a Safety Evaluation, dated February 1, 2002.
CNS's current P/T limits were established based on adjusted
reference temperatures developed in accordance with the procedures
described in RG 1.99, Revision 2. Calculation of adjusted reference
temperature by these procedures includes a margin term to ensure
conservative, upper-bound values are used for the calculation of the
P/T limits. This change does not affect the existing P/T limits in
the CNS Technical Specifications Figures 3.4.9-1, 2, and 3. This
change will not affect any plant safety limits or limiting
conditions of operation. The proposed change will not affect reactor
pressure vessel performance as no physical changes are involved
aside from changes related to surveillance capsule withdrawal, and
CNS vessel P/T limits will remain conservative in accordance with RG
1.99, Revision 2 criteria. The proposed change will not cause the
reactor pressure vessel or interfacing systems to be operated
outside of their design or testing limits. Also, the proposed change
will not alter any assumptions previously made in evaluating the
radiological consequences of accidents. 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 revises the CNS license basis to reflect
participation in the BWRVIP ISP. Participation in the BWRVIP ISP
will continue to ensure that the CNS reactor vessel materials are
monitored and evaluated as necessary to protect against brittle
fracture. This proposed change does not involve a modification of
the design of plant structures, systems, or components. The proposed
change will not impact the manner in which the plant is operated as
plant operating and testing procedures will not be affected by the
change. The proposed change will not degrade the reliability of
structures, systems, or components important to safety as equipment
protection features will not be deleted or modified, equipment
redundancy or independence will not be reduced, supporting system
performance will not be downgraded, the frequency of operation of
equipment will not be increased, and increased or more severe
testing of equipment will not be imposed. No new accident types or
failure modes will be introduced as a result of the proposed change.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from that previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Conformance with 10 CFR [part] 50 Appendix G defines the
accepted safety margin for Reactor Coolant Pressure Boundary
fracture toughness. The P/T limits are not derived from Design Basis
Accident (DBA) analyses. They are prescribed during normal operation
to avoid encountering pressure, temperature, and temperature rate of
change conditions that might cause undetected flaws to propagate and
cause nonductile failure of the reactor pressure vessel, a condition
that is unanalyzed. Since the P/T limits are not derived from any
DBA, there are no acceptance limits related to the P/T limits.
Rather the P/T limits are acceptance limits themselves since they
preclude operation in an unanalyzed condition.
This proposed change will not alter the required margins as
defined in 10 CFR [part] 50, Appendix G. This proposed change will
not affect any safety limits, limiting safety system settings, or
limiting conditions of operation. The proposed change does not
represent a change in initial conditions, or in a system response
time, or in any other parameter affecting the course of an accident
analysis supporting the Bases of any Technical Specification. The
proposed
[[Page 5679]]
change does not involve revision of the P/T limits. Rather, this
change involves a revision to the surveillance capsule withdrawal
schedule, a revision to the reactor vessel fluence calculational
methodology to achieve consistency within the BWRVIP ISP, and
participation in future BWRVIP ISP developments. The current P/T
limits were established based on adjusted reference temperatures for
vessel beltline materials calculated in accordance with RG 1.99,
Revision 2 which will continue to conform to 10 CFR [part] 50
Appendix G. Therefore, the proposed change does not involve a
significant reduction in any safety margins.
In summary, it is concluded that this License Amendment Request
does not involve significant hazards consideration results. NPPD has
researched the existing regulatory precedent and has identified five
BWR licensees with similar License Amendment Requests currently
under NRC staff review:
[sbull] Browns Ferry Units 2 and 3--Submittal date November 6, 2002.
[sbull] Monticello Generating Station--Submittal date September 19,
2002.
[sbull] River Bend--Submittal date August 15, 2002.
[sbull] Fermi Unit 2--Submittal date August 8, 2002.
[sbull] Susquehanna Units 1 and 2--Submittal date July 25, 2002.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: January 13, 2003.
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant (KNPP) operating license and
Technical Specifications to increase the licensed rated power by 1.4
percent to 1673 megawatts thermal (MWt) using measurement uncertainty
recapture.
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 Kewaunee Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
There are no changes as a result of the measurement uncertainty
recapture (MUR) power uprate to the design or operation of the plant
that could affect system, component, or accident mitigative
functions. All systems and components will function as designed and
the applicable performance requirements have been evaluated and
found to be acceptable.
The reduction in power measurement uncertainty allows for some
of the safety analyses to continue to be used without modification.
This is because the safety analyses were performed or evaluated at
either 102 percent of 1650 MWt or higher. Analyses at these power
levels support a core power level of 1673 MWt with a measurement
uncertainty of 0.6 percent. Radiological consequences of USAR
[updated safety analysis report] chapter 14 accidents were assessed
previously using the alternate source term (AST) methodology
(reference 7.1, TAC [technical assignment control] No. MB4596).
These analyses were performed at 102 percent of 1650 MWt and
continue to be bounding. The USAR chapter 14 analyses and accident
analyses submitted to the NRC [Nuclear Regulatory Commission] with
the fuel transition (reference 7.3, TAC No. MB5718) continue to
demonstrate compliance with the relevant accident analyses
acceptance criteria. Therefore, there is no significant increase in
the consequences of any accident previously evaluated.
The primary loop components (reactor vessel, reactor internals,
control rod drive mechanisms, loop piping and supports, reactor
coolant pumps, steam generators, and pressurizer) were evaluated at
an uprated core power level of 1772 MWt and continue to comply with
their applicable structural limits. These analyses also demonstrate
the components will continue to perform their intended design
functions. Changing the applicability of the heatup and cooldown
curves is based on uprated fluence values. This does not have a
significant effect on the reactor vessel integrity. Thus, there is
no significant increase in the probability of a structural failure
of the primary loop components.
All of the NSSS [Nuclear Steam Supply System] systems will
continue to perform their intended design functions during normal
and accident conditions. The auxiliary systems and components
continue to comply with the applicable structural limits and will
continue to perform their intended functions. The NSSS/BOP [balance
of plant] interface systems were evaluated at 1772 MWt and will
continue to perform their intended design functions. Plant
electrical equipment was also evaluated and will continue to perform
their intended functions. Therefore, there is no significant
increase in the probability or consequences of an accident
previously evaluated.
2. Operation of the Kewaunee Nuclear Power Plant in accordance
with the proposed amendments does not result in a new or different
kind of accident from any accident previously evaluated.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of the proposed change. All
systems, structures, and components previously required for the
mitigation of an event remain capable of fulfilling their intended
design function at the uprated power level. The proposed change has
no adverse effects on any safety-related systems or component and
does not challenge the performance or integrity of any safety-
related system. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Operation of the Kewaunee Nuclear Power Plant in accordance
with the proposed amendments does not result in a significant
reduction in a margin of safety.
Operation at the 1673 MWt core power does not involve a
significant reduction in the margin of safety. The current accident
analyses have been previously performed with a two percent power
measurement uncertainty or at uprated core powers that exceed the
MUR uprated core power. System and component analyses have been
completed at a core power in excess of the MUR uprated core power.
Analyses of the primary fission product barriers at uprated core
powers have concluded that all relevant design basis criteria remain
satisfied in regard to integrity and compliance with the regulatory
acceptance criteria. As appropriate, all evaluations have been
either reviewed and approved by the NRC, are in the process of being
approved by the NRC, or are in compliance with applicable regulatory
review guidance and standards. 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: John H. O'Neill, Jr., Esq., Shaw Pittman,
Potts & Trowbridge, 2300 N. Street, NW., Washington, DC 20037-1128.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: September 12, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.5.2, ``ECCS [Emergency Core
Cooling System]--Operating,'' and TS 3.5.3, ``ECCS-Shutdown,'' to add a
surveillance requirement to verify every 31 days that the ECCS piping
is full of water; consistent with NUREG-1431, Standard Technical
Specifications, Westinghouse Plants, Revision 2.
[[Page 5680]]
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 Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
Operation of this facility under the proposed Technical
Specifications will not create a significant increase in the
probability or consequences of an accident previously evaluated.
This license amendment request proposes to add a surveillance
requirement to verify the ECCS is full of water every 31 days while
operating in Modes 1, 2, 3 and 4.
This proposed change does not cause an increase in the
probabilities of any accidents previously evaluated, because the
change will not cause an increase in the probability of any
initiating events for accidents previously evaluated. In particular,
the change affects the ECCS, which serves to mitigate rather than
initiate accidents.
The consequences of the accidents previously evaluated in the
PBNP [Point Beach Nuclear Plant] Final Safety Analysis Report (FSAR)
are determined by the results of analyses that are based on initial
conditions of the plant, the type of accident, transient response of
the plant, and the operation and failure of equipment and systems.
The change proposed in this license amendment request provides an
appropriate surveillance requirement for the ECCS, and thus does not
increase the probability of failure of this equipment or its ability
to operate as required for the accidents previously evaluated in the
PBNP FSAR.
Therefore, the consequences of an accident previously evaluated
in the PBNP FSAR will not be significantly increased as a result of
the proposed change, because the factors that are used to determine
the consequences of accidents are not being changed.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
Equipment important to safety will continue to operate as
designed. The proposed change does not result in any event
previously deemed incredible being made credible. The change does
not result in more adverse conditions or result in any increase in
the challenges to safety systems. Therefore, operation of the Point
Beach Nuclear Plant in accordance with the proposed amendment will
not create the possibility of a new or different type of accident
from any accident previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences. The
proposed amendment will not otherwise affect the plant protective
boundaries and will not cause a release of fission products to the
public. Venting the piping associated with a train of ECCS will
render that ECCS train inoperable while it is being vented.
Performance of this surveillance will therefore affect the
availability of the associated ECCS train, but performance of the
surveillance requirement at the specified frequency is consistent
with the requirements of NUREG-1431, Standard Technical
Specifications for Westinghouse Plants, Revision 2. Additionally,
verifying the ECCS piping is full of water ensures that the system
will perform properly, injecting its full capacity into the RCS
[reactor coolant system], upon demand. Therefore, adopting a
surveillance requirement to verify the ECCS piping is full of water,
will not result in more than a minimal 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: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: L. Raghavan.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: September 26, 2002.
Description of amendment request: The proposed change would revise
the Steam Generator low-low level trip setpoint and allowable values
provided in the Salem Nuclear Generating Station, Unit Nos. 1 and 2,
Technical Specifications Table 2.2-1, ``Reactor Trip System
Instrumentation Trip Setpoints,'' and Table 3.3-4, ``Engineered Safety
Feature Actuation System Instrumentation Trip Setpoints.'' The changes
are necessary based on PSEG Nuclear's evaluation of a loss of feedwater
transient at Diablo Canyon. During the event, Diablo Canyon personnel
observed a flow induced pressure drop in the steam generator mid-deck
area. The proposed change accounts for a level measurement bias
resulting from the pressure drop that was not considered in the
previous Westinghouse analysis. This bias has the effect of providing
nonconservative level readings and setpoints.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change to Tables 2.2-1 and 3.3-4 changes both the
allowable trip setpoint and allowable value for the Steam Generator
Water Level-Low-Low from =9.0% to =14.0% and
from =8.0% to =13% respectively. The Steam
Generator Water Level Low-Low trip provides core protection by
preventing operation with the steam generator water level below the
minimum volume required for adequate heat removal capacity. The
signal is used as a primary protection signal for the design basis
loss of normal feedwater, loss of offsite power and feedwater line
break safety analysis. The specified setpoint provides allowance
that there will be sufficient water inventory in the steam
generators at the time of trip to allow for starting delays of the
auxiliary feedwater system. The change in the setpoint and allowable
value allows the trip to function as originally designed accounting
for the differential pressure created by steam flow past the mid-
deck plate in the moisture separator section of the steam generator.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes to the Steam Generator Water Level-Low-Low
trip setpoint and allowable values allow the trip to function as
originally designed. They do not alter the plant configuration in
any way, and do not replace or modify existing plant equipment, or
affect any plant operations. No additional failure mechanisms are
introduced as a result of the changes to the setpoints and allowable
values.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment would not involve a significant
reduction in the margin of safety.
The proposed changes to the allowable trip setpoint and
allowable value for the Steam Generator Water Level-Low-Low trip
maintains core protection by preventing operation with the steam
generator water level below the minimum volume required for adequate
heat removal capacity.
Therefore, it is concluded that the proposed changes to the
steam generator low low level trip setpoint and allowable value[s]
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
[[Page 5681]]
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: James W. Clifford.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: October 23, 2002.
Description of amendment request: The proposed change would revise
the Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2,
Technical Specification (TS) 6.12, ``High Radiation Area'' to be
consistent with the Standard TSs for Westinghouse Plants (NUREG-1431,
Revision 2) by updating the current reference to title 10 of the Code
of Federal Regulations (10 CFR), section 20.203 with the corresponding
reference to 10 CFR 20.1601.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not affect accident initiators or
precursors and do not alter the design assumptions, conditions,
configuration of the facility, or manner in which the plant is
operated. The proposed changes do not alter or prevent the ability
of structures, systems, or components to perform their intended
safety function to mitigate the consequences of an initiating event
within the acceptance limits assumed in the UFSAR. The proposed
changes are administrative in nature. Technical Specification (TS)
6.12 will be updated to include the new 10 CFR 20 (effective 06/20/
91) requirements. The proposed changes do not alter the conditions
or assumptions in any of the previous accident analyses, and as a
result, the radiological consequences associated with these analyses
remain unchanged.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter the design assumptions,
conditions, configuration of the facility, or the manner in which
the plant is operated.
The proposed changes are administrative in nature and the
relocated procedural details do not change the level of programmatic
controls and procedural details. Accordingly, the proposed changes
do not create any new failure modes or limiting single failures
associated with a plant structure, system, or component important to
safety. Also, there will be no change in the types or increase in
the amounts of any effluents released offsite.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment would not involve a significant
reduction in the margin of safety.
The proposed changes do not impact equipment design or
operation, nor do the changes affect any TS safety limits or safety
system settings that could adversely affect plant safety. The
proposed changes are administrative in nature. Technical
Specification (TS) 6.12 will be updated to include the new 10CFR20
requirements (effective 06/20/91) and are in conformance with NUREG-
1431. Furthermore, the proposed changes do not result in a change in
the types or an increase in the amounts of any effluents released
offsite.
Therefore, it is concluded that the proposed changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 4, 2002 (TS 02-07).
Description of amendment request: The proposed amendment would
revise Technical Specification 6.8.4.h, ``Containment Leakage Rate
Testing Program,'' to allow a one-time, 5-year extension to the current
10-year test interval for the performance-based leakage rate test
program for 10 CFR 50, Appendix J, Type A tests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change for extending Type A test frequency does not
significantly increase the probability of an accident previously
evaluated since the change is not a modification to plant systems,
nor a change to plant operation that could initiate an accident. TVA
performed an evaluation of the risk significance for the proposed
increase to the SQN Units 1 and 2 Type A test frequency. The results
of the TVA risk evaluation indicates that the increase in Large
Early Release Frequency (LERF) remains below the level of risk
significance defined in NRC Regulatory Guide (RG) 1.174, ``An
Approach for Using Risk Assessment In Risk-Informed Decisions On
Plant-Specific Changes to the Licensing Basis.'' TVA's evaluation
indicates that the increase in frequency for all releases (small,
large, early and late) and the increase in radiation dose to the
population is also non-risk significant. The proposed test interval
extension does not involve a significant increase in the
consequences of an accident. Research documented in NUREG-1493
determined that generically, very few potential containment leakage
paths fail to be identified by Type A tests. An analysis of 144 Type
A test results, including 23 failures, found that no failures were
due to containment liner breach. The NUREG concluded that reducing
the Type A test frequency to once per 20 years would lead to an
imperceptible increase in risk. Furthermore, the NUREG concluded
that Type B and C testing provides assurance that containment
leakage from penetration leak paths (i.e., valves, flanges,
containment air-locks) identify any leakage that would otherwise be
detected by the Type A tests. In addition to the NUREG conclusions,
TVA's American Society of Mechanical Engineers (ASME) IWE program
performs containment inspections in order to detect evidence of
degradation that may affect either the containment structural
integrity or leak tightness. In addition to the IWE examinations,
TVA will perform additional nondestructive examinations of the steel
containment vessel in the ice condenser region (inaccessible areas)
at various elevations. These additional non-destructive examinations
will provide added assurance of containment integrity during the 5-
year extended interval. Accordingly, TVA's proposed extension of the
Type A test interval does not increase the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change to extend the Type A test interval does not
create the possibility of a new or different type of accident
because there are no physical changes made to the plant or plant
equipment governing normal plant operation. There are no changes to
the operation of the plant that would introduce a new failure mode
creating the possibility of a new or different kind of accident. TVA
will perform additional non-destructive examinations of the steel
containment vessel in the ice condenser region (inaccessible areas)
at various elevations. These additional non-destructive examinations
will provide added assurance of containment integrity during the 5
year extended interval.
[[Page 5682]]
3. Does the proposed change not involve a significant reduction
in a margin of safety?
The proposed change to extend the Type A test interval will not
significantly reduce the margin of safety. A generic study
documented in NUREG-1493 indicates that extending the Type A leak
test interval to 20 years would result in an imperceptible increase
in risk to the public. The NUREG also found that, generically, the
containment leakage rate contributes a very small amount to the
individual risk and that the decrease in the Type A test frequency
would have a minimal affect on risk because most potential leakage
paths are detected by Type C testing. Previous Type A leakage tests
conducted on SQN Units 1 and 2 indicate that leakage from
containment have been less than the 10 CFR 50, Appendix J leakage
limit of 1.0 La. A review of the previous Type A test
results indicate a stable trend with a 10 percent margin below the
1.0 La leakage limit. Accordingly, these test results, in
conjunction with the research findings from NUREG-1493, provide
assurance that the proposed extension to the Type A test interval
would not significantly reduce the margin of safety. Based on the
above, TVA concludes that the proposed amendment presents no
significant hazards consideration under the standards set forth in
10 CFR 50.92(c), and accordingly, a finding of ``no significant
hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: November 15, 2002 (TS 02-06).
Brief description of amendments: The proposed amendments would
revise the Technical Specification (TS) 3.7.1.3, ``Condensate Storage
Water,'' Limiting Condition for Operation by increasing the required
minimum amount of stored water from 190,000 gallons to 240,000 gallons.
This change is being made to support the replacement steam generator
requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change does not change the physical design and
construction of the condensate storage tank (CST). The purpose of
the increased water volume is to ensure that the required volume of
water, preserved by the technical specification (TS), is sufficient
to meet Sequoyah Nuclear Plant (SQN) Licensing and Design Basis
after installation of the replacement steam generators. The change
in the administratively controlled inventory of the CST will not
increase the probability of an accident. Therefore, the proposed
change does not involve a significant increase in the probability of
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?
This change increases the minimum required volume of water in
the CST, thus ensuring that the auxiliary feedwater (AFW) system can
perform its required safety function, using a preferred water source
for plant transient mitigation. The maximum and normal water levels
in the CST are not being changed. Additionally, increasing the
minimum water volume requirement will not initiate any accident.
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?
This change does not reduce any margin associated with the CST
inventory available to AFW. The requirement for sufficient CST
volume to maintain hot standby and subsequent cooldown to hot
shutdown continues to be met by the minimum volume increase.
Additionally, the essential raw cooling water (ERCW) system still
provides the long-term supply of safety grade cooling water to the
AFW in the event that all inventory of the CST is lost. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: November 15, 2002 (TS 02-01).
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to revise the trip setpoint
column of the Reactor Protection System and Engineered Safety Features
(ESF) instrumentation tables to utilize a nominal setpoint value and
revise the associated Bases discussions.
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:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed revisions for the nominal trip setpoint
representation are administrative changes that will not impact the
application of the reactor trip or ESF actuation system
instrumentation requirements. This is based on the setpoint
requirements being applied without change, as well as the Avs
[allowable values], in accordance with the setpoint methodology. The
removal of the inequalities associated with the trip setpoint values
will be more appropriate for the use of nominal setpoint values but
will not differ in application from the setpoint methodology
utilized by TVA. The revision of the radiation monitoring
instrumentation table to use an Av will continue to provide
appropriate operability limits. Deletion of the nominal terminology
associated with overtemperature delta temperature average
temperature at rated thermal power (T') and reactor coolant system
power operated relief valve (PORV) lift settings provides a better
representation of the limits associated with these values. In
addition, this change will not alter plant equipment or operating
practices. Therefore, the implementation of these changes will not
increase the probability or consequences of an accident.
The revision of the reactor coolant pump (RCP) underfrequency
trip setpoint and the Avs for the RCP underfrequeny and undervoltage
and the containment purge radiation high has been evaluated and the
results are documented in approved calculations. These calculations
verify that the revised values are acceptable in accordance with
appropriate calculation methodologies and that they will continue to
support the accident analysis. This is based on margin being
available in the accuracy determinations that could be used without
impacting the intended functions of this instrumentation and
maintains the established safety limits. These revisions will not
require changes to the instrumentation settings currently being used
or the methods for maintaining them. The offsite dose potential will
not be impacted because this instrumentation will continue to
adequately provide the designed safety functions to limit the
release of radioactivity. Therefore, the proposed revision of these
values will not significantly increase the probability or
consequences of an accident.
The relocation and enhancement of current radiation monitoring
and loss of voltage
[[Page 5683]]
functions to new LCOs [limiting condition for operations] does not
alter the intended functions of these systems or physically alter
these systems. While some requirements have change[d] from current
limitations, these changes have provided more appropriate criteria
to ensure that the accident mitigation functions are maintained
properly and are available. Changes to Avs have been evaluated in
accordance with TVA setpoint methodology and have been verified to
acceptably protect the associated safety limits. Format changes
provide a clearer representation of the requirements and provide
more consistency with the standard TSs in NUREG-1431. These changes
continue to support or improve the required safety functions and
therefore, will not increase the possibility or consequence of an
accident.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The revision of the nominal trip setpoint representation and
elimination of the nominal nomenclature, as well as the revised
setpoint value and Avs, and the relocated LCOs will not alter the
plant configuration or functions. The revised setpoint and the
proposed operability limits will continue to provide acceptable
initiation of safety functions for the mitigation of postulated
accidents as required by the design basis. The primary function of
the reactor protection system, the ESF actuation system, and the new
actuation function LCOs is to initiate accident mitigation
functions. These functions are not considered to be initiators of
postulated accidents. The PORVs provide accident mitigation
functions and could be the source of a loss of coolant accident.
However, a clarification of how to apply the actuation setpoints
without a change to the setpoints will not impact accident
generation. The proposed changes do not create the possibility of a
new or different kind of accident because the design functions are
not altered and the proposed values meet the accident analysis
requirements for accident mitigation.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The setpoint and Av revisions proposed in this request were
evaluated and found to be acceptable based on operating margin
available in the accuracy determinations. The reassignment of this
excess margin to the setpoint and Av will not impact the safety
limits required for the associated functions. The nominal trip
setpoint representation change and the elimination of inappropriate
nominal indications does not alter the TS functions or their
application and will not require changes to design settings. The
relocated requirements to new LCOs provide appropriate limits and
enhancements to the actuation functions. Plant systems will continue
to be actuated for those plant conditions that require the
initiation of accident mitigation functions. The margin of safety is
not significantly reduced because the proposed changes to the Av and
setpoint representations will not change design functions and the
initiation of accident mitigation functions for appropriate plant
conditions will not be adversely impacted.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 5, 2002.
Description of amendment request: The proposed changes would delete
the monthly analog rod position test for the control rod bottom
bistables currently required by Technical Specification (TS) Table 4.1-
1, Item 9.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change deletes the monthly analog rod position test
that verifies the operation of the rod bottom bistables. However,
the TSs still require bistable action to be functionally verified to
ensure operability on an 18-month frequency as part of the overall
analog rod position indication system calibration. Furthermore, the
TS-required monthly rod bottom bistable action test was being
performed to address instrument drift in the rod bottom setpoint,
which will essentially be eliminated by the design of new digital-
based IRPI [Individual Rod Position Indication] electronics being
installed. Consequently, elimination of the monthly rod bottom
bistable action test will not result in the failure of any plant
structures, systems, or components and does not have a detrimental
impact on the integrity of any plant structure, system, or component
that initiates an analyzed event. The proposed change will not alter
the operation of or otherwise increase the failure probability of
any plant equipment that initiates an analyzed accident. As a
result, the probability of any accident previously evaluated is not
significantly increased.
Consequences of analyzed events are the result of the plant
being operated within assumed parameters at the onset of any event,
and the successful functioning of at least one train or division of
the equipment credited with mitigating the event. These changes do
not impact the capability of the credited equipment to perform, nor
is there any change in the likelihood that credited equipment will
fail to perform. Deletion of the monthly rod bottom bistable action
test does not affect the ability of the control rods to perform
their function. Surveillance tests to verify the operability of the
IRPI System are still being performed. Furthermore, the Rod Position
Demand Counter System provides redundant control rod position
indication. As a result, the consequences of any accident previously
evaluated are not significantly affected by the proposed change.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed change deletes the monthly surveillance of rod
bottom bistable action in the Individual Rod Position Indication
system. This change does not alter the methods governing normal
plant operation. The IRPI provides indication of rod position, is
one of two independent systems that are provided to detect a rod
drop and is the backup to detection by rapid reduction of ex-core
neutron flux. The dropping of a rod assembly can occur when the rod
drive mechanism is de-energized from the Rod Control System. This
accident has been evaluated in the UFSAR and in all cases the DNB
design bases is met by demonstration that the DNBR is greater than
the limiting value. Thus, this change deleting the monthly analog
rod position test does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The digital-based IRPI system continues to meet the design
function of providing reliable control rod position indication. The
proposed change and associated replacements with digital-based IRPI
system electronics provides enhanced testing through the automatic
self-testing diagnostic features. Consequently, the overall ability
to detect failures is not degraded. Therefore, the change deleting
the monthly analog rod position test 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: Ms. Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station,
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford,
Connecticut 06385.
NRC Section Chief: John A. Nakoski.
[[Page 5684]]
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendments request: December 19, 2002.
Description of amendments request: The proposed changes would
revise the Technical Specifications (TS) to Facility Operating License
Nos. DRP-32 and DRP-37 for Surry Power Station, Units 1 and 2,
respectively, to reflect changes in regulations, correct typographical
and editorial errors made in previous TS revisions, and to revise TS
cross-references to Updated Final Safety Analysis Report sections.
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:
Dominion has reviewed the requirements of 10 CFR 50.92 as they
relate to the proposed administrative change to the Surry Power
Station Units 1 and 2 Technical Specifications (TS) and Bases. The
proposed change to the Surry TS makes administrative revisions to
reflect changes in regulations, corrects editorial and typographical
errors from previous TS revisions, and revises TS cross-references
to Updated Final Safety Analysis Report (UFSAR) sections. Due to the
strictly administrative nature of the proposed TS change, we have
determined that a significant hazards consideration does not exist.
The basis for this determination is provided as follows:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed change is administrative in nature and as such does
not impact the condition or performance of any plant structure,
system or component. The proposed administrative change does not
affect the initiators of any previously analyzed event nor the
assumed mitigation of accident or transient events. As a result, the
proposed change to the Surry Technical Specifications does not
involve any increase in the probability [nor] the consequences of
any accident or malfunction of equipment important to safety
previously evaluated since neither accident probabilities or
consequences are being affected by this proposed administrative
change.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed change is administrative in nature, and therefore
does not involve any changes in station operation or physical
modifications to the plant. In addition, no changes are being made
in the methods used to respond to plant transients that have been
previously analyzed. No changes are being made to plant parameters
within which the plant is normally operated or in the setpoints,
which initiate protective or mitigative actions and no new failure
modes are being introduced. Therefore, the proposed administrative
change to the Surry Technical Specifications does not create the
possibility of a new or different kind of accident or malfunction of
equipment important to safety from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed change is administrative in nature, and does not
impact station operation or any plant structure, system or component
that is relied upon for accident mitigation. Furthermore, the margin
of safety assumed in the plant safety analysis is not affected in
any way by the administrative ``cleanup'' of the Surry Technical
Specifications. Therefore, the proposed administrative change to the
Surry Technical Specifications does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station,
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford,
Connecticut 06385.
NRC Section Chief: John A. Nakoski.
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 PDR, located at One White Flint North, Public File
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: December 19, 2001, as
supplemented July 30, 2002, and November 14, 2002.
Brief description of amendment: The amendment includes a revision
of the Technical Specification (TS) Limiting Conditions for Operation
3.4, ``Decay Heat Removal Capability,'' conforming changes to TS Table
3.5-2, ``Accident Monitoring Instruments,'' and TS 4.9.1.2, ``Decay
Heat Removal--Periodic Testing,'' and numerous editorial changes.
Date of issuance: January 16, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 242.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 19, 2002 (67 FR
12598).
The supplements dated July 30, and November 14, 2002, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a
[[Page 5685]]
Safety Evaluation dated January 16, 2003.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: July 2, 2002.
Brief description of amendments: These amendments change the
administrative controls in Technical Specification 5.7, ``High
Radiation Area.''
Date of issuance: January 13, 2003.
Effective date: January 13, 2003.
Amendment Nos.: 225 and 252.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50950).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 13, 2003.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 14, 2002, as supplemented by letter
dated December 17, 2002.
Brief description of amendment: The amendment revised Technical
Specification Table 3.3.8.1-1, ``Loss of Power Instrumentation,'' by
changing the degraded voltage--voltage basis and loss-of-coolant
accident time delay allowable values to reflect the results of new
calculations performed in association with a design basis
reconstitution.
Date of issuance: January 16, 2003.
Effective date: As of the date of issuance and shall be implemented
no later than November 30, 2003.
Amendment No.: 128.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42823).
The December 17, 2002, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 16, 2003.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington Date of application for amendment: October
22, 2002.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to change TS section 5.0, ``Administrative
Controls,'' and adopt Technical Specification Task Force (TSTF) -258,
Revision 4. The change revises: (1) Section 5.2.2, ``Unit Staff,'' to
delete details of staffing requirements and delete requirements for the
Shift Technical Advisor (STA) as a separate position while retaining
the function, (2) section 5.5.4, ``Radioactive Effluent Controls
Program,'' to be consistent with the intent of 10 CFR part 20, (3)
section 5.6.4, ``Monthly Operating Reports,'' to delete periodic
reporting requirements for main steam safety/relief valve challenges to
be consistent with Generic Letter 97-02, ``Revised Contents of the
Monthly Operating Report,'' and (4) section 5.7, ``High Radiation
Area,'' in accordance with 10 CFR 20.1601(c). TS section 5.3.2 is added
to incorporate regulatory definitions for the senior reactor operator
(SRO) and reactor operator (RO) positions.
Date of issuance: January 9, 2003.
Effective date: January 9, 2003, and shall be implemented within 60
days from the date of issuance.
Amendment No.: 182.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75870).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 9, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of application for amendment: March 28, 2002.
Brief description of amendment: The amendment revised Technical
Specification (TS) sections 3.7, ``Auxiliary Electrical Systems,'' and
4.6, ``Emergency Power System Periodic Tests,'' to relocate the
requirements for the gas turbine generators to the Updated Final Safety
Analysis Report (UFSAR) and the plans, programs and procedures that
document and control the credited functions of these systems,
structures, and components. The amendments also deleted TS 3.7.B.2.b.
to remove the option that allows power operation for up to 72 hours
with a gas turbine as the only available 13.8 kilovolt power source.
Date of issuance: January 17, 2003.
Effective date: This license amendment is effective as of the date
of its issuance and shall be implemented within 60 days and only after
incorporation of the required changes into the UFSAR and completion of
the necessary implementation and procedural changes.
Amendment No.: 236.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications and Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34484).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 17, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: July 5, 2002.
Brief description of amendment: The amendment relocates Technical
Specification (TS) 3/4.6.I to the Pilgrim Nuclear Power Station Updated
Final Safety Analysis Report. The affected TS contains snubber
operability and surveillance requirements. The associated Bases section
will also be relocated.
Date of issuance: January 14, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 195.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68735).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 14, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: September 27, 2002.
Brief description of amendments: The amendments revise Appendix B,
[[Page 5686]]
``Environmental Protection Plan (Non-Radiological),'' of the licenses
to remove a parenthetical reference to a superseded section of 10 CFR
part 51.
Date of issuance: January 21, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 211 & 205.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised Appendix B of the licenses.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66009).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 21, 2003.
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: October 8, 2002.
Brief description of amendment: The amendment relocates Technical
Specification 2.13, ``Nuclear Detector Cooling System,'' and its
associated Bases to the Fort Calhoun Station Updated Safety Analysis
Report.
Date of issuance: January 16, 2003.
Effective date: January 16, 2003, and shall be implemented within
120 days of the date of issuance. Implementation includes the
incorporation of changes to the Fort Calhoun Station Updated Safety
Analysis Report as described in the licensee's application dated
October 8, 2002.
Amendment No.: 214.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68741).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 16, 2003.
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: July 22, 2002, as supplemented by letter
dated October 8, 2002.
Brief description of amendment: The amendment revises Technical
Specification (TS) 5.19, Surveillance Requirement (SR) 3.0.4 and adds
TS 5.20 and SR 3.0.5 to extend the delay period before entering a
limiting condition for operation following a missed surveillance.
Date of issuance: January 16, 2003.
Effective date: January 16, 2003, and shall be implemented with 120
days from the date of issuance, including the incorporation of changes
to the technical specification Bases as described in the licensee's
application dated July 22, 2002, as supplemented by letter dated
October 8, 2002.
Amendment No.: 215.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56326).
The supplemental letter of October 8, 2002, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 16, 2003.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: September 23, 2002.
Brief description of amendments: These amendments revised Technical
Specification (TS) section 2.6.2.4, ``Residual Heat Removal [RHR]
Suppression Pool Cooling,'' to adopt TS Task Force (TF) change 230,
Revision 1 (TSTF-230, Revision 1). This change to Required Action B of
Limiting Condition for Operation 3.6.2.3 allows two RHR suppression
pool cooling subsystems to be inoperable for up to 8 hours.
Date of issuance: January 16, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 207, 181.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66012). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 16, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: December 17, 2002, as
supplemented December 31, 2002.
Brief description of amendment: The amendment provides a one-time
change to Technical Specification (TS) 4.8.1.1.2.h.14 to allow the
testing of Hope Creek's emergency diesel generator (EDG) lockout relays
to be performed at power until startup from its eleventh refueling
outage (spring 2003). The current TS surveillance requirement only
allows the EDG lockout relays to be tested during shutdown conditions.
PSEG requested that the TS change be issued on an exigent basis in
accordance with title 10 of the Code of Federal Regulations (10 CFR),
part 50, section 50.91(a)(6). Approval and implementation of the TS
change allows the testing that has been completed to be used to comply
with TS 4.8.1.1.2.h.14.
Date of issuance: January 10, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 141.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Public Comments Requested as to Proposed No Significant Hazards
Consideration: Yes (67 FR 79163) December 27, 2002. That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by January 27, 2003, but indicated that if the Commission makes
a final no significant hazards consideration determination, any such
hearing would take place after the issuance of the amendment. The
Commission's related evaluation of the amendment, finding of exigent
circumstances, final determination of no significant hazards
consideration, and state consultation are contained in a safety
evaluation dated January 10, 2003.
NRC Section Chief: James W. Clifford.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendments request: February 18, 2002, as supplemented in
letter dated July 23, 2002.
Brief description of amendments: The proposed amendments revised
[[Page 5687]]
Technical Specification 3/4.6.1.7, ``Containment Ventilation System,''
to extend the intervals between operability tests of the normal and
supplementary containment purge valves.
Date of issuance: January 7, 2003.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-147; Unit 2-135.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 19, 2002 (67 FR
12608). The July 23, 2002, supplemental letter provided clarifying
information that was within the scope of the original Federal Register
notice (67 FR 12608) and did not change the initial no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 7, 2003.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project. Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 22, 2002.
Brief description of amendments: The proposed amendments revised
Technical Specification 3/4.3.5, allowing the automatic operation of
the atmospheric steam relief valves during Mode 2 to maintain secondary
side pressure at or below an indicated steam generator pressure of 1225
psig during startup and shutdown of the reactors.
Date of issuance: January 13, 2003.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-148; Unit 2-136.
Facility Operating License Nos. NPF-76 and NPF-80: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45571).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 13, 2003.
No significant hazards consideration comments received: No.
Dated in Rockville, Maryland, this 28th day of January 2003.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-2415 Filed 2-3-03; 8:45 am]
BILLING CODE 7590-01-P