[Federal Register Volume 70, Number 157 (Tuesday, August 16, 2005)]
[Notices]
[Pages 48201-48210]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E5-4403]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice Applications and Amendments to Facility Operating
Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 22, 2005, to August 4, 2005. The last
biweekly notice was published on August 2, 2005 (70 FR 44400).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or
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fact. Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: June 20, 2005.
Description of amendments request: The proposed change would revise
the Technical Specification Surveillance Requirement 3.6.1.6.2 of
3.6.1.6, ``Suppression Chamber-to-Drywell Vacuum Breakers'' for the
frequency of functionally testing the suppression chamber-to-drywell
vacuum breakers.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change revises Surveillance Requirement [SR]
3.6.1.6.2 to require performance of functional testing of each
suppression chamber-to-drywell vacuum breaker every 92 days, within
12 hours after any discharge of steam to the suppression chamber
from the safety/relief valves, and within 12 hours following an
operation that causes any of the vacuum breakers to open.
The proposed change does not involve physical changes to any
plant structure, system, or component. The suppression chamber-to-
drywell vacuum breakers only provide an accident mitigation
function. As such, the probability of occurrence for a previously
analyzed accident is not impacted by the change to the surveillance
frequency for these components. The consequences of a previously
analyzed accident are dependent on the initial conditions assumed
for the analysis, the behavior of the fuel during the analyzed
accident, the availability of successful functioning of the
equipment assumed to operate in response to the analyzed event, and
the setpoints at which these actions are initiated. No physical
change to suppression chamber-to-drywell vacuum breakers is being
made as a result of the proposed change, nor does the change alter
the manner in which the vacuum breakers operate. As a result, no new
failure modes of the suppression chamber-to-drywell vacuum breakers
are being introduced. The proposed quarterly surveillance frequency
for the suppression chamber-to-drywell vacuum breakers is consistent
with the American Society of Mechanical Engineers (ASME) Code
frequency for testing these valves, will avoid unnecessary cycling
and wear of the vacuum breakers, and will improve the reliability of
the vacuum breakers. Based on this evaluation, there is no
significant increase in the consequences of a previously analyzed
event.
Therefore, the proposed change to the surveillance frequency for
the suppression chamber-to-drywell vacuum breakers does not involve
a significant increase in the probability or consequences of an
accident previously analyzed.
2. Does not create the possibility of a new or different type of
accident from any accident previously evaluated.
The proposed change to the surveillance frequency for the
suppression chamber-to-drywell vacuum breakers does not involve any
physical alteration of plant systems, structures, or components. No
new or different equipment is being installed. No installed
equipment is being operated in a different manner. There is no
alteration to the parameters within which the plant is normally
operated or in the setpoints that initiate protective or mitigative
actions. As a result no new failure modes are being introduced.
Therefore, the proposed change to the surveillance frequency for the
suppression chamber-to-drywell vacuum breakers does 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 the margin of
safety.
The proposed change revises SR 3.6.1.6.2 to require performance
of functional testing of each vacuum breaker every 92 days, within
12 hours after any discharge of the steam to the suppression chamber
from the safety/relief valves, and within 12 hours following an
operation that causes any of the vacuum breakers to open. The
operability and functional characteristics of the suppression
chamber-to-drywell vacuum breakers remains unchanged. The margin of
safety is established through the design of the plant structures,
systems, and components, through the parameters within which the
plant is operated, through the establishment
[[Page 48203]]
of the setpoints for the actuation of equipment relied upon to
respond to an event, and through the margins contained within the
safety analyses. The proposed change to the surveillance frequency
for the suppression chamber-to-drywell vacuum breakers does not
impact the condition or performance of structures, systems,
setpoints, and components relied upon for accident mitigation. As
previously noted, the proposed quarterly surveillance frequency for
the suppression chamber-to-drywell vacuum breakers is consistent
with the ASME Code for frequency for testing these vacuum breakers,
will avoid unnecessary cycling and wear of the vacuum breakers, and
will improve the reliability of the vacuum breakers. The proposed
change does not impact any safety analysis assumptions or results.
Therefore, the proposed change does not result in 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: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Jr.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: June 29, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) to revise Surveillance
Requirements (SR) 3.6.1.3.11 and 3.6.1.3.12 in TS 3.6.1.3, ``Primary
Containment Isolation Valves (PCIVs).'' Specifically, the proposed
amendment would revise the combined secondary containment bypass
leakage rate limit for all bypass leakage paths in SR 3.6.1.3.11 from
0.05 to 0.10 La and the combined main steam isolation valve
(MSIV) leakage rate limit for all four main steam lines in SR
3.6.1.3.12 from 150 to 250 standard cubic feet per hour (scfh).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The increase in the allowed secondary containment bypass leakage
limit in SR 3.6.1.3.11 and the increase in the total Main Steam
Isolation Valve (MSIV) leakage rate limit have been evaluated in a
revision to the analysis of the Loss of Coolant Accident (LOCA).
Based on the results of the analysis, it has been demonstrated that,
with the requested change, the dose consequences of this limiting
Design Basis Accident (DBA) are within the regulatory guidance
provided by the NRC [Nuclear Regulatory Commission] for use with the
AST [alternative source term]. This guidance is presented in 10 CFR
50.67, Regulatory Guide 1.183, ''Alternative Radiological Source
Terms For Evaluating Design Basis Accidents At Nuclear Power
Reactors,'' and Standard Review Plan (SRP) Section 15.0.1. The
proposed change also updates the design basis value for the Control
Room Envelope (CRE) unfiltered inleakage based on actual test
results. This is acceptable because the assumed value in the
analysis is more than three times the worst case test value. The
proposed change does not affect the normal design or operation of
the facility before the accident; rather, it affects leakage limit
assumptions that constitute inputs to the evaluation of the
consequences. The radiological consequences of the analyzed LOCA
have been evaluated using the plant licensing basis for this
accident. The results conclude that the control room and offsite
doses remain within applicable regulatory limits. Therefore, the
proposed 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 accident previously
evaluated.
The change in leakage limits does not affect the design,
functional performance or normal operation of the facility.
Similarly, it does not affect the design or operation of any
component in the facility such that new equipment failure modes are
created. As such the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
This proposed license amendment involves changes in leakage rate
limits for the secondary containment bypass leakage and MSIV
leakage. The revised leakage rate limits are used in the LOCA
radiological analysis in conjunction with the revised CRE unfiltered
inleakage limit. The analysis has been performed using conservative
methodologies. Safety margins and analytical conservatisms have been
evaluated and have been found acceptable. The analyzed LOCA event
has been carefully selected and margin has been retained to ensure
that the analysis adequately bounds postulated event scenario. The
dose consequences of this limiting event are within the acceptance
criteria presented in 10 CFR 50.67, Regulatory Guide 1.183 and SRP
Section 15.0.1. The margin of safety is that provided by meeting the
applicable regulatory limits. The effect of the revision to the
Technical Specification requirements has been analyzed and doses
resulting from the pertinent design basis accident have been found
to remain within the regulatory limits. The change continues to
ensure that the doses at the exclusion area and low population zone
boundaries, as well as the control room, are within the
corresponding regulatory limits. Therefore, the proposed change will
not involve a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Section Chief: L. Raghavan.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of amendment request: June 8, 2005.
Description of amendment request: The proposed change allows a
delay time for entering a supported system Technical Specification (TS)
when the inoperability is due solely to an inoperable snubber, if risk
is assessed and managed consistent with the program in place for
complying with the requirements of 10 CFR 50.65(a)(4). Limiting
Condition for Operation (LCO) 3.0.8 is added to the TS to provide this
allowance and define the requirements and limitations for its use.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
November 24, 2004 (69 FR 68412), on possible amendments concerning
TSTF-372, including a model safety evaluation and model no significant
hazards consideration (NSHC) determination, using the consolidated line
item improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on May 4, 2005 (70 FR 23252). The
licensee affirmed the applicability of the following NSHC determination
in its application dated June 8, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change allows a delay time for entering a supported
system TS when the
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inoperability is due solely to an inoperable snubber if risk is
assessed and managed. The postulated seismic event requiring
snubbers is a low-probability occurrence and the overall TS system
safety function would still be available for the vast majority of
anticipated challenges. Therefore, the probability of an accident
previously evaluated is not significantly increased, if at all. The
consequences of an accident while relying on allowance provided by
proposed LCO 3.0.8 are no different than the consequences of an
accident while relying on the TS required actions in effect without
the allowance provided by proposed LCO 3.0.8. Therefore, the
consequences of an accident previously evaluated are not
significantly affected by this change. The addition of a requirement
to assess and manage the risk introduced by this change will further
minimize possible concerns. Therefore, this change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in Regulatory Guide 1.177. A bounding risk assessment
was performed to justify the proposed TS changes. The proposed LCO
3.0.8 defines limitations on the use of the provision and includes a
requirement for the licensee to assess and manage the risk
associated with operation with an inoperable snubber. The net change
to the margin of safety is insignificant. Therefore, this change
does not involve a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: May 31, 2005.
Description of amendment request: The proposed change allows entry
into a mode or other specified condition in the applicability of a
Technical Specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of Title 10 of the Code of Federal
Regulations (10 CFR), part 50, section 50.65(a)(4). Limiting Condition
for Operation (LCO) 3.0.4 exceptions in individual TSs would be
eliminated, several notes or specific exceptions are revised to reflect
the related changes to LCO 3.0.4, and Surveillance Requirement (SR)
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated May 31, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
[[Page 48205]]
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: May 24, 2005.
Description of amendment request: The proposed amendment would
delete the Technical Specification (TS) temperature limit for the
safety relief valve (SRV) discharge pipe and the requirements for NRC
approval of the associated engineering evaluation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. This proposed change deletes an administrative
requirement for NRC approval of an engineering evaluation to resolve
a non-conforming and degraded condition that is required by NRC
Generic Letter 91-18 (GL), Rev. 1, ``Information to Licensees
Regarding NRC Inspection Manual Section on Resolution of Degraded
and Nonconforming Conditions''. The SRVs will be maintained
operable, inspected, and tested to perform their safety function as
required by the current Specifications and any SRV non-conforming or
degraded condition will be addressed in accordance with GL 91-18.
The proposed change also deletes a Note regarding installed two-
stage Target Rock SRVs. The deletion of an administrative
requirement and the Note does not change the plant response to the
design basis accident and does not increase the probability of
inadvertent SRV operation. Therefore, the proposed change does not
significantly increase the probability or consequences of any
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 safety function of the SRVs is to provide
over-pressure protection of the primary coolant pressure boundary
and also for the automatic functions to rapidly depressurize the
primary system to a pressure at which low-pressure cooling systems
can provide makeup. The proposed change deletes an administrative
requirement and a Note related to installed two-stage Target Rock
SRVs, and does not introduce any new modes of equipment operation or
failure. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The ability of the SRVs to perform their safety
function is maintained during operation and will continue to be
tested as required in accordance with TS 3/4.13, Inservice Code
Testing. The proposed change deletes an administrative requirement
that is adequately addressed by following GL 91-18, Rev. 1. Deletion
of an administrative requirement does not reduce the margin of
safety. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: Darrell Roberts.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: May 24, 2005.
Description of amendment request: The proposed amendment would
delete the main steam isolation valve (MSIV) twice per week partial
stroke testing surveillance specified in Technical specification (TS)
4.7.A.2.b.1.c.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. This proposed change deletes the requirement to
exercise the MSIV's twice per week at power. The MSIVs will continue
to be full stroke tested by the Inservice Testing Program. The MSIVs
will continue to be able to perform their accident mitigation
function. The plant response to the design basis accident will not
change and the probability of inadvertent MSIV closure will not be
increased. Therefore, the proposed change does not significantly
increase the probability or consequences of any 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 safety function of the MSIVs is to isolate the
main steam lines in case of design basis accidents to limit the loss
of reactor coolant and/or limit the release of radioactive
materials. The proposed change does not introduce any new modes of
equipment operation or failure. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The ability of the MSIVs to perform their safety
function is tested during the MSIV full stroke fast closure test in
accordance with TS 3.13, Inservice Testing Program. The proposed
change deletes a high-risk surveillance. Deletion of the high-risk
surveillance does not reduce the margin of safety. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: Darrell Roberts.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: March 7, 2005.
Description of amendment request: The proposed amendment request
will add two NRC approved topical report references to the list of
analytical methods in Technical Specification 5.6.5, ``Core Operating
Limits Report (COLR),'' that can be used to determine core operating
limits. The proposed changes are:
1. Add a NRC previously approved Siemans Power Corporation (SPC)
topical report reference for determination of fuel assembly critical
power for previously loaded Global Nuclear Fuel (GNNF) GE14 fuel
which will be co-resident with reloaded Framatome ANP ATRIUM-10
fuel.
2. Add a NRC previously approved Framatome Advanced Nuclear
Power, Inc. (FRA-ANP) topical report reference for an uprated
methodology for evaluation of loss coolant accident (LOCA)
conditions.
The proposed changes are the result of a redesign to untilize
Framatome ANP ATRIUM-10 fuel during the Unit 1 Refueling Outage 11
currently scheduled for February 2006.
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:
[[Page 48206]]
Criterion 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The proposed changes will add two additional NRC approved
topical report references to the list of administratively controlled
analytical methods in Technical Specification (TS) 5.6.5, ``Core
Operating Limits Report (COLR),'' that can be used to determine core
operating limits. TS 5.6.5 lists NRC approved analytical methods
used at LaSalle County Station (LSCS) to determine core operating
limits.
LSCS Unit 1 is scheduled to reload Framatome ANP ATRIUM-10 fuel
during the Unit 1 Refueling Outage 11currently scheduled for
February 2006. The proposed changes to TS Section 5.6.5 will add
FRA-ANP methodologies to determine overall core operating limits for
future core configurations. This change will require the listing of
additional analytical methods for evaluating LOCA conditions and
determining the critical power performance of the GE14 fuel. Thus,
the proposed changes will allow LSCS to use the most recent FRA-ANP
LOCA methodology for evaluation of ATRIUM-10 fuel and SPC critical
power correlations to determine the critical power for the GE14
fuel.
The addition of approved methods to TS Section 5.6.5 has no
effect on any accident initiator or precursor previously evaluated
and does not change the manner in which the core is operated. The
methods have been reviewed to ensure that the output accurately
models predicted core behavior, have no effect on the type or amount
of radiation released, and have no effect on predicted offsite doses
in the event of an accident. Additionally the methods do not change
any key core parameters that influence any accident consequences.
Thus, the proposed changes do not have any effect on the probability
of an accident previously evaluated.
The methodology conservatively establishes acceptable core
operating limits such that the consequences of previously analyzed
events are not significantly increased.
The proposed changes in the administratively controlled
analytical methods do not affect the ability of LSCS to successfully
respond to previously evaluated accidents and does not affect
radiological assumptions used in the evaluations. Thus, the
radiological consequences of any accident previously evaluated are
not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--Does the proposed change create the possibility of
a new or different kind of accident from any previously evaluated?
Response: No.
The proposed changes involve TS 5.6.5 do not affect the
performance of any LSCS structure, system, or component credited
with mitigating any accident previously evaluated. The insertion of
fuel, which has been analyzed with NRC approved methodologies, will
not affect the control parameters governing unit operation or the
response of plant equipment to transient conditions. The proposed
changes do not introduce any new modes of system operation or
failure mechanism.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Criterion 3--Do the proposed changes involve a significant
reduction in the margin of safety.
Response: No.
The proposed changes will add two additional references to the
list of administratively controlled analytical methods in TS 5.6.5
that can be used to determine core operating limits. The proposed
changes do not modify the safety limits or setpoints at which
protective actions are initiated and do not change the requirements
governing operation or availability of safety equipment assumed to
operate to preserve the margin of safety. Therefore, LSCS has
determined that the proposed changes provide an equivalent level of
protection as that currently provided.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: July 5, 2005.
Description of amendment request: The proposed amendment would
modify the existing Technical Specification (TS) 3.3.1.3, ``Oscillation
Power Range Monitor (OPRM) Instrumentation,'' Surveillance Requirement
(SR) 3.3.1.3.5. Specifically, the thermal power level at which the
OPRMs are ``not bypassed'' (enabled to perform their design function)
will be changed from > 28.6 percent rated thermal power to >= 23.8
percent rated thermal power.
Plant-specific stability calculations are now required as part of
the resolution to several generic issues associated with OPRM
operability. One of the outcomes from this resolution was a change in
the OPRM enabled region of the power to flow map. The thermal power
level for enabling the OPRMs for Cycle 10 became > 27.2 percent rated
thermal power. Since the current TS SR requirement is > 28.6 percent,
the new TS SR thermal power level value is considered a non-
conservative TS. The Perry Nuclear Power Plant (PNPP) is currently
requiring the OPRMs to be enabled at >= 23.8 percent thermal power
level through administrative controls. These controls will remain in
place until such time that this license amendment is approved
(reference NRC Administrative Letter 98-10, ``Dispositioning of
Technical Specifications That Are Insufficient to Assure Plant
Safety,'' dated December 12, 1998).
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 involves the use of a revised thermal power
level to establish the OPRM enabled region. The OPRM enabled region
is that area on the power to flow map where the OPRM System is
activated to detect and suppress potential instability events. If
reactor operations result in entrance into this region and a core
instability is detected, the OPRM System will automatically initiate
a reactor scram. The revised enabled region provides assurance that
the requirements of 10CFR50, Appendix A, General Design Criteria 10
and 12 remain satisfied for current and future core designs. Though
the initiation of instability events are dependent upon thermal
power levels and core flows, the revision to the enabled region
thermal power level value does not increase the possibility of such
an event. Once the OPRMs are enabled, the OPRM System would still
mitigate an instability event, if detected. The revised enabled
region does not impact any OPRM detection or mitigation actions for
instability events.
The OPRMs are designed to detect and suppress potential
instability events. As such, the OPRMs are not credited to provide
any type of detection or mitigation actions for transients or
accidents described within the PNPP Updated Final Safety Analysis
Report (USAR) other than instability events. Hence, revising the
OPRMs enabled region will not impact the transients or accidents
described within the PNPP Updated Safety Analysis Report (USAR)
other than instability events.
Since the OPRMs will be enabled at a thermal power lower than
analytically required, the potential for additional scrams exists.
However, since the possibility of an instability event occurring in
the range between the revised thermal power level and the analytical
value is remote, the probability of an additional scram from
occurring is not significantly increased.
[[Page 48207]]
Therefore, since no significant changes are being made to the
plant or its design, the probability or the consequences of an
accident have not increased over those 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 involves the use of a revised thermal power
level to establish the OPRM enabled region. The use of a revised
thermal power level to establish the OPRM enabled region does not
involve a physical modification to any plant system or component,
including the fuel. The revised enabled region provides assurance
that the requirements of 10CFR50, Appendix A, General Design
Criteria 10 and 12 remain satisfied for current and future core
designs. Though the initiation of instability events are dependent
upon thermal power levels and core flows, the revision to the
enabled region thermal power level value does not increase the
possibility of such an event, or introduce any new or different
events. Once the OPRMs are enabled, the OPRM System detects and
mitigates an instability event if detected. The revised enabled
region does not impact any mitigation actions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change involves the use of a revised thermal power
level to establish the OPRM enabled region. Once the OPRMs are
enabled, the OPRM System mitigates an instability event if detected.
The revised enabled region does not impact any mitigation actions.
The use of a revised thermal power level to establish the OPRM
enabled region does not involve a physical modification to any plant
system or component, including the fuel. The revised enabled region
provides assurance that the requirements of 10CFR50, Appendix A,
General Design Criteria 10 and 12 remain satisfied for current and
future core designs. The revised enabled region restores the margin
of protection provided by the OPRMs, which had been reduced as fuel
and core designs have evolved since 1994. 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: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: May 25, 2005.
Description of amendment request: The proposed change allows entry
into a mode or other specified condition in the applicability of a
Technical Specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of Title 10 of the Code of Federal
Regulations (10 CFR), part 50, section 50.65(a)(4). Limiting Condition
for Operation (LCO) 3.0.4 exceptions in individual TSs would be
eliminated, several notes or specific exceptions are revised to reflect
the related changes to LCO 3.0.4, and Surveillance Requirement (SR)
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated May 25, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and
[[Page 48208]]
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Evangelos C. Marinos.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: July 5, 2005.
Description of amendment request: The proposed changes to the
Technical Specifications (TS) would add a reference in TS 5.65.b,
``Core Operating Limits Report (COLR),'' to permit the use of an
alternate methodology, VIPRE-D/BWU code/correlation (Virginia Electric
and Power Company version of the Electric Power Research Institute
(EPRI) computer code VIPRE [Versatile Internals and Components Program
for Reactors--EPRI] with the BWU Critical Heat Flux (CHF)
correlations), to perform thermal-hydraulic analysis to predict CHF and
Departure from Nucleate Boiling Ratio (DNBR) for the AREVA Advanced
Mark-BW (AMBW) fuel in the North Anna cores.
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 probability of occurrence or the consequences of an
accident previously evaluated are not significantly increased.
Neither the code/CHF correlation pair nor the Statistical DNBR
Evaluation Methodology make any contribution to the potential
accident initiators and thus cannot increase the probability of any
accident. Further, since both the deterministic and statistical DNBR
limits meet the required design basis of avoiding DNB with 95%
probability at a 95% confidence level, the use of the new code/
correlation and Statistical DNBR Evaluation Methodology do not
increase the potential consequences of any accident. Finally the
addition of a full core DNB design limit provides increased
assurance that the consequences of a postulated accident which
included radioactive release would be minimized because the overall
number of rods in DNB would not exceed the 0.1% level. All the
pertinent evaluations to be performed as part of the cycle specific
reload safety analysis to confirm that the existing safety analyses
remain applicable have been performed with VIPRE-D/BWU and found to
be acceptable. The use of a different code/correlation pair will not
increase the probability of an accident because plant systems will
not be operated in a different manner, and system interfaces will
not change. The use of the VIPRE-D/BWU code/correlation pair will
not result in a measurable impact on normal operating plant
releases, and will not increase the predicted radiological
consequences of accidents postulated in the UFSAR [Updated Final
Safety Analysis Report]. Therefore, neither the probability of
occurrence nor the consequences of any accident previously evaluated
is significantly increased.
2. The possibility for a new or different type of accident from
any accident previously evaluated is not created.
The use of VIPRE-D/BWU and its applicable fuel design limits for
DNBR does not impact any of the applicable design criteria and all
pertinent licensing basis criteria will continue to be met.
Demonstrated adherence to these standards and criteria precludes new
challenges to components and systems that could introduce a new type
of accident. Setpoint safety analysis evaluations have demonstrated
that the use of VIPRE-D/BWU is acceptable. All design and
performance criteria will continue to be met and no new single
failure mechanisms will be created. The use of VIPRE-D/BWU code/
correlation or the Statistical DNBR Evaluation Methodology does not
involve any alteration to plant equipment or procedures that would
introduce any new or unique operational modes or accident
precursors. Therefore, the possibility for a new or different kind
of accident from any accident previously evaluated is not created.
3. The margin of safety is not significantly reduced. North Anna
Technical Specification 2.1 specifies that any DNBR limit
Established by any used code/correlation must provide at least 95%
non-DNB probability at a 95% confidence level. The use of VIPRE-D/
BWU with the SDLs [Statistical Design Limits] listed in this package
provides that protection, just as LYNXT/BWU [LYNXT thermal-hydraulic
computer code with the AREVA BWU CHF correlations] and applicable
SDLs did. The required DNBR margin of safety for the North Anna
Nuclear units, which in this case is the margin between the 95/95
DNBR limit and clad failure, is therefore not reduced. Therefore,
the margin of safety as defined in the Bases to the North Anna Units
1 and 2 Technical Specifications is not significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: Evangelos C. Marinos.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: July 14, 2005.
Description of amendment request: The proposed changes to the
Technical Specifications (TS) would correct two errors in the units of
measure used to determine the Overtemperature [Delta]T Function
Allowable Value.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do changes involve a significant increase in the probability
or consequences of an accident previously evaluated?
The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the UFSAR [Updated Final Safety Analysis Report]. The proposed
changes correct errors in the unit designations used in the
f1([Delta]I) equation. The actual numerical values of
f1([Delta]I) calculated by the equation remain the same,
only the units applied to the value are changed. The Overtemperature
[Delta]T function allowable values are utilized by the Reactor Trip
System (RTS) instrumentation to prevent reactor operation in
conditions outside the range considered for accident analyses. The
proposed changes will not alter the allowable values used by the RTS
instrumentation. The Overtemperature [Delta]T allowable value is not
an initiator to any accident previously evaluated. As a result, the
probability of any accident previously evaluated is not
significantly increased. As the Overtemperature [Delta]T allowable
value is not changed, the probability or consequences of an accident
previously evaluated is not significantly increased.
2. Do changes create the possibility of a new or different kind
of accident from any accident previously evaluated?
The proposed changes do not create the possibility of a new or
different kind of accident from any accident already evaluated in
the UFSAR. The proposed changes correct errors in the unit
designations used in the f1([Delta]I) equation. Changes
do not introduce a new mode of plant operation and do not involve
any physical modifications to the plant. The changes will not
introduce new accident initiators. Therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Do changes involve a significant reduction in the margin of
safety?
The proposed changes do not involve a significant reduction in a
margin of safety. The proposed changes correct errors in the unit
designations used in the f1([Delta]I) equation. This will
eliminate the possibility of an error resulting from incorrect
interpretation of the equation and potential subsequent errors in
the application of the equation. The allowable value of the
Overtemperature [Delta]T function is unaffected. Therefore, the
proposed changes will not significantly reduce the margin of safety
as defined in the Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 48209]]
review, it appears that the three standards of 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: Evangelos C. Marinos.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: December 14, 2004.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.3.G, ``Scram Discharge Volume,'' for the condition
of having one or more SDV vent or drain lines with inoperable valves.
Date of issuance: July 29, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 216.
Facility Operating License No. DPR-35: The amendment revised the
TSs.
Date of initial notice in Federal Register: May 24, 2005 (70 FR 29792).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 29, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: April 8, 2004.
Brief description of amendments: These amendments relocated several
Technical Specifications (TSs) from Section 6, ``Administrative
Controls,'' requirements to the Quality Assurance Topical Report.
Specifically, the amendments relocated (1) the Plant Operations Review
Committee and Nuclear Review Board requirements, (2) the program/
procedure review and approval requirements, and (3) the record-
retention requirements.
Date of issuance: July 25, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 176 and 138.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the TSs.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34701).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 25, 2005.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al. Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: February 22, 2005.
Brief description of amendments: The amendments revise Technical
Specifications by eliminating the requirements to provide the NRC
monthly operating reports and annual occupational radiation exposure
reports.
Date of issuance: July 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 266 and 148.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24651).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 28, 2005.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: July 29, 2004.
Brief description of amendment: The amendment deleted the
requirements from the technical specifications to maintain a hydrogen
dilution system, a hydrogen purge system, and hydrogen monitors.
Date of issuance: August 1, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 265.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7764).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 1, 2005.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: October 15, 2004.
Brief description of amendment: The amendment revises surveillance
requirements related to the reactor
[[Page 48210]]
coolant pump flywheel inspections to extend the allowable inspection
interval to 20 years.
Date of issuance: July 27, 2005.
Effective date: July 27, 2005.
Amendment No.: 218.
Facility Operating License No. DPR-72: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9992).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 27, 2005.
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: May 11, 2004.
Brief description of amendments: The amendments revise Technical
Specification (TS) Surveillance Requirement 3.1.7.7 acceptance criteria
from 1224 psig to 1395 psig in TS 3.1.7, ``Standby Liquid Control
System.''
Date of issuance: July 25, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 221, 198.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 6, 2004 (69 FR
40678).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 25, 2005.
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 8, 2004.
Brief description of amendments: The amendments revised Technical
Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain
Valves,'' for the condition of having one or more SDV vent or drain
lines with one or both valves inoperable.
Date of issuance: July 26, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 222 and 199.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70721).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 26, 2005.
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 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: September 8, 2004.
Brief description of amendments: The amendments revised SSES 1 and
2 Technical Specification (TS) Surveillance Requirement 3.6.1.3.6 of TS
3.6.1.3, ``Primary Containment Isolation Valves,'' to reduce the
frequency of performing leakage rate testing for each primary
containment purge valve with resilient seals from 184 days to 24
months.
Date of issuance: August 4, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 223 and 200.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9995).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 4, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259 Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendment: August 2, 2004 (TS-435).
Brief description of amendment: The amendment modifies the
Technical Specification (TS) 3.6.3.1 required action to provide 7 days
of continued operation with two Containment Atmosphere Dilution
subsystems inoperable.
Date of issuance: July 18, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 255.
Facility Operating License Nos. DPR-33: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 2004 (69 FR
64991).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 18, 2005.
No significant hazards consideration comments received: No.
Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power
Station (YNPS) Franklin County, Massachusetts
Date of amendment request: November 24, 2003, and supplemented by
letters dated December 10, 2003, December 16, 2003, January 19, 2004,
January 21, 2004, February 10, 2004, March 4, 2004, April 27, 2004,
August 3, 2004, September 2, 2004, September 2, 2004, September 30,
2004, November 19, 2004, December 10, 2004, and April 7, 2005.
Supplemental letters provided additional clarifying information that
did not expand the scope of the application as originally noticed and
did not change the staff's original proposed no significant hazards
consideration determination.
Description of amendment request: The amendment revises the license
to incorporate a new license condition addressing the license
termination plan (LTP). This amendment documents the approval of the
LTP, documents the criteria for making changes to the LTP which will
and will not require pre-approval by the NRC, and documents the
conditions imposed with the approval of the LTP.
Date of issuance: July 28, 2005.
Effective date: Effective as of the date of issuance and shall be
implemented within 30 days from the date of issuance.
Amendment No.: 158.
Facility Operating License No. DPR-3: Amendment revises the
license.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7823).
The Commission's related evaluation of the amendment, state
consultation, and final NSHC determination are contained in a safety
evaluation dated July 28, 2005.
No significant hazards consideration comments received: No.
NRC Section Chief: Claudia Craig.
Dated at Rockville, Maryland, this 8th day of August, 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. E5-4403 Filed 8-15-05; 8:45 am]
BILLING CODE 7590-01-P