[Federal Register Volume 71, Number 1 (Tuesday, January 3, 2006)]
[Notices]
[Pages 145-159]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-24669]
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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 December 9, 2005 to December 21, 2005. The
last biweekly notice was published on December 20, 2005 (70 FR 75489).
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
[[Page 146]]
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by 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-
[[Page 147]]
4209, (301) 415-4737 or by e-mail to [email protected].
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: September 19, 2005.
Description of amendment request: Pursuant to 10 CFR 50.90, Entergy
Operations, Inc. hereby requests an Operating License amendment for
Arkansas Nuclear One, Unit 2, to replace the existing steam generator
(SG) tube surveillance program with that being proposed by the
Technical Specifications Task Force (TSTF) in TSTF 449, Revision 4.
Specifically, Technical Specification (TS) 1.1, Definitions; TS 3/
4.4.5, Steam Generators; TS 3.4.6.2, Reactor Coolant System Leakage; TS
6.5.9, Steam Generator Tube Surveillance Program; and TS 6.6.7, Steam
Generator Tube Surveillance Reports are being revised to incorporate
the new Steam Generator Program of TSTF 449, Revision 4. The proposed
changes are consistent with the Consolidated Line Item Improvement
Process provided in the May 6, 2005, Federal Register Notice (70 FR
24126).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change requires a Steam Generator Program that
includes performance criteria that will provide reasonable assurance
that the steam generator (SG) tubing will retain integrity over the
full range of operating conditions (including startup, operation in
the power range, hot standby, cooldown and all anticipated
transients included in the design specification). The SG performance
criteria are based on tube structural integrity, accident induced
leakage, and operational leakage.
The structural integrity performance criterion is:
Structural integrity performance criterion: All in-service steam
generator tubes shall retain structural integrity over the full
range of normal operating conditions (including startup, operation
in the power range, hot standby, and cool down and all anticipated
transients included in the design specification) and design basis
accidents. This includes retaining a safety factor of 3.0 against
burst under normal steady state full power operation primary to
secondary pressure differential and a safety factor of 1.4 against
burst applied to the design basis accident primary to secondary
pressure differentials. Apart from the above requirements,
additional loading conditions associated with the design basis
accidents, or combination of accidents in accordance with the design
and licensing basis, shall also be evaluated to determine if the
associated loads contribute significantly to burst or collapse. In
the assessment of tube integrity, those loads that do significantly
affect burst or collapse shall be determined and assessed in
combination with the loads due to pressure with a safety factor of
1.2 on the combined primary loads and 1.0 on axial secondary loads.
The accident induced leakage performance criterion is:
The primary to secondary accident induced leakage rate for any
design basis accidents, other than a SG tube rupture, shall not
exceed the leakage rate assumed in the accident analysis in terms of
total leakage rate for all SGs and leakage rate for an individual
SG. Leakage is not to exceed 1 gpm through any one SG.
The operational leakage performance criterion is:
The RCS operational primary to secondary leakage through any one
SG shall be limited to <=150 gallons per day per SG.
A steam generator tube rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis of a SGTR event, a bounding primary to
secondary leakage rate equal to the leakage rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as main steam line break
(MSLB) and control element assembly (CEA) ejection, the tubes are
assumed to retain their structural integrity (i.e., they are assumed
not to rupture). The accident induced leakage criterion introduced
by the proposed changes accounts for tubes that may leak during
design basis accidents. The accident induced leakage criterion
limits this leakage to no more than the value assumed in the
accident analysis.
The SG performance criteria proposed change identify the
standards against which tube integrity is to be measured. Meeting
the performance criteria provides reasonable assurance that the SG
tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design basis accident. The performance criteria are only a part of
the Steam Generator Program required by the proposed change. The
program, defined by NEI 97-06, Steam Generator Program Guidelines,
includes a framework that incorporates a balance of prevention,
inspection, evaluation, repair, and leakage monitoring.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in the plant technical specifications
for operational leakage and for DOSE EQUIVALENT I-131 in primary
coolant to ensure the plant is operated within its analyzed
condition. The typical analysis of the limiting design basis
accident assumes that primary to secondary leak rate after the
accident is 1 gallon per minute with no more than 720 gallons per
day in any one SG, and that the reactor coolant activity levels of
DOSE EQUIVALENT I-131 are at the technical specification values
before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current technical specifications and
enhances the requirements for SG inspections. The proposed change
does not adversely impact any other previously evaluated design
basis accident and is an improvement over the current technical
specifications.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of other design basis events.
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 performance based requirements are an improvement
over the requirements imposed by the current technical
specifications.
Implementation of the proposed Steam Generator Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the Steam Generator Program will be
an enhancement of SG tube performance. Primary to secondary leakage
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical
[[Page 148]]
condition of the tube. The proposed change does not affect tube
design or operating environment. The proposed change is expected to
result in an improvement in the tube integrity by implementing the
Steam Generator Program to manage SG tube inspection, assessment,
and plugging. The requirements established by the Steam Generator
Program are consistent with those in the applicable design codes and
standards and are an improvement over the requirements in the
current technical specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: September 19, 2005.
Description of amendment request: Entergy Operations, Inc.,
proposes to amend Technical Specification (TS) 3.6.2.1, ``Containment
Spray System,'' to allow a one-time extension of the allowable outage
time (AOT) for the Containment Spray System (CSS) from 72 hours to a
maximum of 7 days, to be used once for each train or, at most, two
times during fuel cycles 18 and 19. The proposed change is intended to
provide flexibility in scheduling CSS maintenance activities, reduce
refueling outage duration, and improve the availability of CSS
components important to safety during plant shutdowns.
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 TS change does not affect the design, operational
characteristics, function or reliability of the CSS.
The CSS is primarily designed to mitigate the consequences of a
Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). The
requested change does not affect the assumption used in the
deterministic LOCA or MSLB analyses.
The duration of a TS AOT is determined considering that there is
a minimal possibility that an accident will occur while a component
is removed from service. A risk informed assessment was performed
which concluded that the increase in plant risk is small and
consistent with the guidance contained in Regulatory Guide 1.177
[``An Approach for Plant-Specific Risk-Informed Decisionmaking:
Technical Specifications''].
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 previously evaluated?
Response: No.
The proposed change does not involve a change in the design,
configuration, or method of operation of the plant that could create
the possibility of a new or different kind of accident. The proposed
change extends the AOT currently allowed by the TS to 7 days.
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 Containment Heat Removal System (CHRS) consists of the CSS
and the Containment Cooling System (CCS). The CHRS functions to
rapidly reduce the containment pressure and temperature after a
postulated LOCA or MSLB accident by removing thermal energy from the
containment atmosphere. The CHRS also assists in limiting off-site
radiation levels by reducing the pressure differential between the
containment atmosphere and the outside atmosphere, thereby reducing
the driving force for leakage of fission products from the
containment.
The CHRS is designed so that either both trains of the CSS, or
one train of CSS and one train of CCS will provide adequate heat
removal to attenuate the post-accident pressure and temperature
conditions imposed upon the containment following a LOCA or MSLB.
The proposed change includes administrative controls that will
be established to ensure one train of CSS and one train of CCS will
be available during the extended CSS AOT.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: October 18, 2005.
Description of amendment request: The proposed amendment would
revise applicability requirements related to single control rod
withdrawal allowances in shutdown modes.
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 special operation allowances do not
involve the modification of any plant equipment or affect basic
plant operation. The relevant design basis analyses are associated
with refueling operations. The refueling interlocks are designed to
back up procedural core reactivity controls during refueling
operations to prevent an inadvertent criticality during refueling
operations. The relaxations proposed in relocating and revising
single controlrod withdrawal allowances during the Refueling MODE
with the reactor vesselhead fully tensioned, to the proposed special
operations allowances consistent with NUREG-1433 recommendations,
will not increase the probability of an accident compared to a
withdrawal of a rod while in Refueling MODE with the reactor vessel
head removed. This is because the proposed special operations will
allow the withdrawal of only one control rod at a time while
requiring the one-rod-out interlock to be OPERABLE and other
requirements imposed to ensure that all other rods remain fully
inserted. This requirement coupled with the reactivity margin
requirement for the most reactive rod fully withdrawn or removed, is
adequate to prevent inadvertent criticality when a single rod is
withdrawn for maintenance or testing. As such, there is no
significant increase in the probability of an accident previously
evaluated. Since no criticality is assumed to occur, the
consequences of analyzed events are therefore not affected.
Therefore, the proposed change does not involve a significant
increase in the 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 physical
alteration of existing plant equipment or the installation of new
equipment. The basic operation of installed equipment is unchanged
and no new accident initiators or failure modes are introduced as a
result of these changes. The methods governing plant operation and
[[Page 149]]
testing remain consistent with current safety analysis assumptions.
These changes do not adversely affect existing plant safety margins
or the reliability of the equipment assumed to operate in the safety
analysis. The requirements imposed during these Special Operations
ensure the existing analyses and equipment operating conditions
remain bounding. 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 margin of safety is not reduced because the
proposed requirements offer similar protection to those imposed
during normal refueling activities. The proposed special operation
allowances do not involve the modification of any plant equipment or
affect basic plant operation. The proposed allowances limit the
withdrawal of only one control rod at a time. This allowance is
controlled by the reactor mode switch in the refuel position, or
other precautions to prevent the withdrawal or removal of more than
one rod and the requirement that adequate reactivity margin be
maintained. These requirements are adequate to prevent an
inadvertent criticality. These changes do not adversely affect
existing plant safety margins or the reliability of the equipment
assumed to operate in the safety analysis. As such, there are no
changes being made to safety analysis assumptions, safety limits or
safety system settings that would adversely affect plant safety as a
result of the proposed change. 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 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 Branch Chief: Richard Lauder.
Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of amendment request: January 10, 2005.
Description of amendment request: The proposed change will delete
the License Conditions concerning emergency core cooling system pump
suction strainers from Appendix C of the Limerick Generating Station,
Unit No. 1 Facility Operating License that were added by Amendment No.
128.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed change is administrative in nature.
The proposed change does not involve the modification of any plant
equipment nor does it affect basic plant operation. The proposed
change will have no impact on any safety related structures, systems
or components. The License Conditions proposed for deletion pertain
to actions that have been completed and are obsolete, or involve
activities that are controlled in accordance with other regulatory
processes, i.e., 10 CFR 50.59 and 10 CFR 50.65.
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 is administrative in nature.
The proposed change has no impact on the design, function or
operation of any plant structure, system or component and does not
affect any accident analyses. The License Conditions in Appendix C
can be deleted because they are obsolete or involve activities that
are controlled in accordance with other regulatory processes.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed change is administrative in nature,
does not negate any existing requirement, and does not adversely
affect existing plant safety margins or the reliability of the
equipment assumed to operate in the safety analysis. As such, there
is no change being made to safety analysis assumptions, safety
limits or safety system settings that would adversely affect plant
safety as a result of the proposed change. Margins of safety are
unaffected by deletion of the License Conditions.
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. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Darrell J. Roberts.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 29, 2005.
Description of amendment request: The proposed amendment would
eliminate operability requirements for Secondary Containment, Secondary
Containment Isolation Valves, the Standby Gas Treatment System, and
Secondary Containment Isolation Instrumentation when handling
irradiated fuel that has decayed for 24 hours since critical reactor
operations and when performing Core Alterations. Similar technical
specification relaxations are proposed for the Control Room Emergency
Filter System and its initiation instrumentation after a decay period
of 7 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment involves implementation of the
Alternative Source Term (AST) for the fuel handling accident (FHA)
at Cooper Nuclear Station (CNS). There are no physical design
modifications to the plant associated with the proposed amendment.
The FHA AST calculation does not impact the initiators of an FHA in
any way.
The changes also do not impact the initiators for any other
design[-]basis accident (DBA) or events. Therefore, because DBA
initiators are not being altered by adoption of the AST analyses the
probability of an accident previously evaluated is not affected.
With respect to consequences, the only previously evaluated
accident that could be affected is the FHA. The AST is an input to
calculations used to evaluate the consequences of the accident, and
does not, in and of itself, affect the plant response or the actual
pathways to the environment utilized by the radiation/activity
released by the fuel. It does, however, better represent the
physical characteristics of the release, so that appropriate
mitigation techniques may be applied. For the FHA, the AST analyses
demonstrate acceptable doses that are within regulatory limits after
24 hours of radioactive decay since reactor shutdown, without credit
for Secondary Containment, the Standby Gas Treatment System,
Secondary Containment Isolation Valves, or Secondary Containment
Isolation Instrumentation, and that the Control Room Emergency
Filter System (CREFS) and CREFS Instrumentation need not be credited
after a 7[-]day period of decay. Therefore, the consequences of an
[[Page 150]]
accident previously evaluated are not significantly increased.
Based on the above conclusions, this proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment does not involve a physical alteration of
the plant. No new or different types of equipment will be installed
and there are no physical modifications to existing equipment
associated with the proposed changes. The proposed changes to the
control of Engineered Safety Features during handling of irradiated
fuel do not create new initiators or precursors of a new or
different kind of accident. New equipment or personnel failure modes
that might initiate a new type of accident are not created as a
result of the proposed amendment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The proposed amendment is associated with the implementation of
a new licensing basis for the CNS FHA. Approval of this change from
the original source term to an AST derived in accordance with the
guidance of Regulatory Guide (RG) 1.183 is being requested. The
results of the FHA analysis, revised in support of the proposed
license amendment, are subject to revised acceptance criteria. The
AST FHA analysis has been performed using conservative
methodologies, as specified in RG 1.183. Safety margins have been
evaluated and analytical conservatism has been utilized to ensure
that the analysis adequately bounds the postulated limiting event
scenario. The dose consequences of the limiting FHA remain within
the acceptance criteria presented in 10 CFR 50.67, the Standard
Review Plan, and RG 1.183.
The proposed changes continue to ensure that the doses at the
Exclusion Area Boundary (EAB) and Low Population Zone (LPZ)
boundary, as well as the Control Room, are within the corresponding
regulatory limits. For the FHA, RG 1.183 conservatively sets the EAB
and LPZ limits below the 10 CFR 50.67 limit, and sets the Control
Room limit consistent with 10 CFR 50.67.
Since the proposed amendment continues to ensure the doses at
the EAB, LPZ and Control Room are within corresponding regulatory
limits, the proposed license amendment 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. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: David Terao.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 12, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3.4.9, ``RCS [reactor
coolant system] Pressure and Temperature (P/T) Limits,'' curves 3.4.9-
1, ``Pressure/Temperature Limits for Non-Nuclear Heatup or Cooldown
Following Nuclear Shutdown,'' 3.4.9-2, ``Pressure/Temperature Limits
for Inservice Hydrostatic and Inservice Leakage Tests, and 3.4.9-3,
``Pressure/Temperature Limits for Criticality,'' to remove the cycle
operating restriction and replace it with a limitation of 30 effective
full-power years (EFPY).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed revisions to the Cooper Nuclear Station (CNS) P/T
curves are based on the recommendations in Regulatory Guide (RG)
1.99, Revision 2, and are, therefore, in accordance with the latest
Nuclear Regulatory Commission (NRC) guidance. The fluence evaluation
for the P/T curves for 30 EFPY was performed using the NRC-approved
Radiation Analysis Modeling Application (RAMA) fluence methodology.
The curves generated from this method provide guidance to ensure
that the P/T limits will not be exceeded during any phase of reactor
operation. Accordingly, the proposed revision to the CNS P/T curves
is based on an NRC accepted means of ensuring protection against
brittle reactor vessel fracture, and compliance with 10 CFR 50
Appendix G. The curves are the same as approved in Amendment Number
204, CNS is only requesting to remove the one cycle limitation and
limit their use to 30 EFPY based on the shift in the Adjusted
Reference Temperature (ART) using the new fluence values. Therefore,
this proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Based on the above, NPPD [Nebraska Public Power District]
concludes that the proposed TS change to TS 3.4.9[,] P/T curves,
Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 does not significantly
increase the probability or consequences of an accident previously
evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change updates existing P/T operating limits to
correspond to the current NRC guidance. The proposed TS change
extends the use of the current, NRC-approved P/T curves beyond the
end of Cycle 23 to 30 EFPY. The proposed change does not involve a
physical change to the plant, add any new equipment or any new mode
of operation. These TS changes demonstrate compliance with the
brittle fracture requirements of 10 CFR 50 Appendix G and,
therefore, do not create the possibility for a new or different kind
of accident from any accident previously evaluated.
Based on the above, NPPD concludes that the proposed TS change
to TS 3.4.9[,] P/T curves, Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the existing CNS P/T curves to limit
their use to 30 EFPY based on fluence calculation using the NRC-
approved Radiation Analysis Modeling Application (RAMA) fluence
methodology. The curves have not been recalculated. Limiting the use
of the P/T curves to 30 EFPY, based on the recalculation of the
fluence per the NRC-approved (RAMA) fluence methodology does not
affect a margin of safety. These changes do not affect any system
used to mitigate accidents or transients.
Based on the above, NPPD concludes that the proposed TS change
to TS 3.4.9[,] P/T curves, Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: David Terao.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: September 16, 2005.
Description of amendment request: The proposed amendment would
revise the surveillance requirements (SRs) for the emergency Diesel
Generators (EDGs) to provide more margin to the acceptance criterion.
The new SR
[[Page 151]]
acceptance criterion will allow the EDG frequency to be within 2 percent of the rated value. The current acceptance limit is
nominally 1 percent of rated frequency.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change. The EDG are not an initiator of any
accident previously evaluated. As a result, the probability of any
accident previously evaluated is not significantly increased. The
consequences of any accident previously evaluated are not increased,
as the EDG will continue to meet their safety function, as specified
in the accident analysis, in a highly reliable manner.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. The changes do not alter assumptions made in the safety
analysis for the EDG performance. The proposed changes remain
consistent with the safety analysis assumptions (e.g., UFSAR Section
8.3.1.4).
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises the acceptance criterion for EDG
Surveillances to match that in the NRC's guidelines (Safety Guide 9)
and the Improved Standard Technical Specifications (NUREG-1433, Rev
3). Because the EDG can perform to the specified acceptance
criterion as stated in the UFSAR Section 8.3.1.4; the EDG will
continue to meet their specified safety function in the safety
analysis, in a highly reliable manner.
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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: November 21, 2005.
Description of amendment request: The proposed amendments to
Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2 Operating
Licenses, would allow extension of the Completion Time associated with
Technical Specification (TS) 3.8.1 Required Action B4, from 7 days to
14 days and for concomitant TS changes. The proposed amendment would
also allow online performance of emergency diesel generator maintenance
activities that are currently performed during refueling outages, to
provide additional flexibility.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes Technical Specification
changes to extend the Technical Specification 3.8.1, ``AC Sources-
Operating,'' Completion Time for an inoperable emergency diesel
generator to 14 days. These changes allow an emergency diesel
generator to be inoperable for 7 days more than Technical
Specification 3.8.1 currently provides. A minor format correction on
the Technical Specification 3.8.1 Actions Table is also proposed.
The emergency diesel generators are safety related components
which provide backup electrical power supply to the onsite
Safeguards Distribution System. The emergency diesel generators are
not accident initiators, thus allowing an emergency diesel generator
to be inoperable for an additional 7 days for performance of
maintenance or testing does not increase the probability of a
previously evaluated accident.
Deterministic and probabilistic risk assessments evaluated the
effect of the proposed Technical Specification changes on the
availability of an electrical power supply to the plant emergency
safeguards features systems. These assessments concluded that the
proposed Technical Specification changes do not involve a
significant increase in the risk of power supply unavailability.
The plant emergency safeguards features systems consist of two
trains for 100% redundancy within each unit. Accident analyses
demonstrate that only one emergency safeguards features train is
required for accident mitigation. Thus, with one train inoperable
the other train is capable of performing the required safety
function. Design basis analyses are not required to be performed
assuming extended loss of all power supplies to the plant emergency
safeguards features systems. Thus this change does not involve a
significant increase in the consequences of a previously analyzed
accident.
The Technical Specification format correction is an
administrative change and does not involve a significant increase in
the probability or consequences of an accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This license amendment request proposes Technical Specification
changes to extend the Technical Specification 3.8.1, ``AC Sources-
Operating,'' Completion Time for an inoperable emergency diesel
generator to 14 days. These changes allow an emergency diesel
generator to be inoperable for 7 days more than Technical
Specification 3.8.1 currently provides. A minor format correction on
the Technical Specification 3.8.1 Actions Table is also proposed.
The proposed Technical Specification changes do not involve a
change in the plant design, system operation, or procedures involved
with the emergency diesel generators. The proposed changes allow an
emergency diesel generator to be inoperable for additional time.
There are no new failure modes or mechanisms created due to plant
operation for an extended period to perform emergency diesel
generator maintenance or testing. Extended operation with an
inoperable emergency diesel generator does not involve any
modification in the operational limits or physical design of plant
systems. There are no new accident precursors generated due to the
extended allowed Completion Time.
The Technical Specification format correction is an
administrative change and does not create the possibility of a new
or different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
This license amendment request proposes Technical Specification
changes to extend the Technical Specification 3.8.1, ``AC Sources-
Operating,'' Completion Time for an inoperable emergency diesel
generator to 14 days. These changes allow an emergency diesel
generator to be inoperable for 7 days
[[Page 152]]
more than Technical Specification 3.8.1 currently provides. A minor
format correction on the Technical Specification 3.8.1 Actions Table
is also proposed.
Currently, if an inoperable emergency diesel generator is not
restored to operable status within 7 days, Technical Specification
3.8.1 will require unit shutdown to MODE 3 within 6 hours and MODE 5
within 36 hours. The proposed Technical Specification changes will
allow steady state plant operation at 100% power for an additional 7
days.
There is some risk associated with continued operation for an
additional 7 days with one emergency diesel generator inoperable.
This risk is judged to be small and reasonable consistent with the
risk associated with operations for 7 days with one emergency diesel
generator inoperable as allowed by the current Technical
Specifications. Specifically, the remaining operable emergency
diesel generator and paths are adequate to supply electrical power
to the onsite Safeguards Distribution System. An emergency diesel
generator is required to operate only if both offsite power sources
fail and there is an event which requires operation of the plant
emergency safeguards features such as a design basis accident. The
probability of a design basis accident occurring during this period
is low.
Deterministic and probabilistic risk assessments evaluated the
effect of the proposed Technical Specification changes on the
availability of an electrical power supply to the plant emergency
safeguards features systems. These assessments concluded that the
proposed Technical Specification changes do not involve a
significant increase in the risk of power supply unavailability.
There is also some risk associated with the Technical
Specification unit shutdown evolutions. Plant load change evolutions
require additional plant operations activities which introduce
equipment challenges, increase the risk of plant trip and increase
the risk for operational errors. Also unit shutdown does not remove
the desirability of having emergency diesel generator backup for the
4 kV safeguards buses, but rather places dependence on the operable
4 kV bus by requiring operation of the residual heat removal system.
Thus, possible additional risk associated with continuing operation
an additional 7 days with an inoperable emergency diesel generator
may be offset by avoiding the additional risk associated with unit
shutdown.
Therefore, based on the considerations given above, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: September 30, 2005.
Description of amendment request: Omaha Public Power District
(OPPD) proposes to change the licensing basis by replacing EMF-
2087(P)(A), Revision 0, ``SEM/PWR-98: ECCS [Emergency Core Cooling
System] Evaluation Model for PWR [pressurized-water reactor] LBLOCA
[large break loss-of-coolant accident] Applications,'' Siemens Power
Corporation, June 1999, with the AREVA Topical Report EMF-2103(P)(A),
``Realistic Large Break LOCA Methodology,'' Framatome ANP, Inc. in the
Fort Calhoun Station, Unit 1 (FCS) Core Operating Limit Report (COLR).
Currently, fuel for the FCS is supplied by AREVA. AREVA has performed
an FCS-specific LBLOCA analysis using their realistic LBLOCA
methodology for Cycle 24 and beyond.
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 amendment replaces EMF-2087(P)(A), Revision 0,
``SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications,''
Siemens Power Corporation, June 1999 (Reference 8.6 [of the
licensee's amendment request]), with the AREVA Topical Report EMF-
2103(P)(A), ``Realistic Large Break LOCA Methodology,'' Framatome
ANP, Inc. (Reference 8.1 [of the licensee's amendment request]) in
the FCS COLR. AREVA Topical Report EMF-2103(P)(A) will also replace
EMF-2087(P)(A) in OPPD topical report OPPD-NA-8303 (Reference 8.5
[of the licensee's amendment request]). This amendment will allow
the use of the RLBLOCA [realistic large break loss-of-coolant
accident] methodology to perform the FCS LBLOCA analysis. The
proposed amendment will not affect any previously evaluated
accidents because they are analyzed using applicable NRC[-]approved
methodologies to ensure all required safety limits are met.
The proposed amendment does not affect any acceptance criteria
for any postulated accidents or anticipated operational occurrences
(AOOs) analyzed and listed in the FCS Updated Safety Analysis Report
(USAR). The proposed change will not increase the likelihood of a
malfunction of a structure, system or components (SSC) since the
change does not involve operation of SSCs in a manner or
configuration different from those previously evaluated.
The results from the FCS RLBLOCA analysis have demonstrated the
adequacy of the ECCS, and these results satisfy the regulatory
criteria set forth in 10 CFR 50.46(b).
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not result in changes in the operation
or overall configuration of the facility. The proposed amendment
does not involve a change in the design function or the operation of
SSCs involved. The proposed amendment does not involve the operation
or configuration of the SSCs different from those previously
analyzed. The proposed amendment to add the RLBLOCA methodology to
the FCS COLR and OPPD topical report OPPD-NA-8303 (Reference 8.5 [of
the licensee's amendment request]) does not create any new or
different kind of 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?
Response: No.
AREVA has performed the RLBLOCA analysis for FCS and
demonstrated that the Emergency Core Cooling System (ECCS) is
adequate to mitigate the consequences of a[n] LBLOCA. The analysis
has concluded that the acceptance criteria for the ECCS are met with
significantly increased margins.
All required safety limits will continue to be analyzed using
methodologies approved by the Nuclear Regulatory Commission.
Therefore, the proposed amendment 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: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Branch Chief: David Terao.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: October 5, 2005.
Description of amendment request: The proposed amendment would
delete
[[Page 153]]
requirements from the Technical Specifications (TSs) to maintain
hydrogen recombiners and hydrogen and oxygen monitors. A notice of
availability for this TS improvement using the consolidated line item
improvement process was published in the Federal Register on September
25, 2003 (68 FR 55416).
Licensees were generally required to implement upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide (RG) 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2 in 1979. Requirements
related to combustible gas control were imposed by order for many
facilities and were added to, or included in, the TSs for nuclear power
reactors currently licensed to operate. The revised Title 10 of the
Code of Federal Regulations (10 CFR) Section 50.44, ``Combustible gas
control for nuclear power reactors,'' eliminated the requirements for
hydrogen recombiners and relaxed safety classifications and licensee
commitments to certain design and qualification criteria for hydrogen
and oxygen monitors.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
availability of a model no significant hazards consideration (NSHC)
determination for referencing license amendment applications in the
Federal Register on September 25, 2003 (68 FR 55416). The licensee
affirmed the applicability of the model NSHC determination in its
application dated October 5, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The NRC has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG 1.97 Category 1, is intended for key variables that
most directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen and oxygen monitors no
longer meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44, the NRC found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents. Also, as part of the
rulemaking to revise 10 CFR 50.44, the NRC found that Category 2, as
defined in RG 1.97, is an appropriate categorization for the oxygen
monitors, because the monitors are required to verify the status of
the inert containment.
The regulatory requirements for the hydrogen and oxygen monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3,
[classification of the oxygen monitors as Category 2,] and removal
of the hydrogen and oxygen monitors from TSs will not prevent an
accident management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and site survey monitoring that support modification of
emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TSs, does
not involve a significant increase in the probability or the
consequences of any accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TSs, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen monitor equipment are not considered accident precursors,
nor does their existence or elimination have any adverse impact on
the pre-accident state of the reactor core or post accident
confinement of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TSs, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The NRC has found that this hydrogen
release is not risk-significant because the design-basis LOCA
hydrogen release does not contribute to the conditional probability
of a large release up to approximately 24 hours after the onset of
core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Category 2 oxygen monitors are adequate to verify the status of
an inserted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TSs will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Lauder.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: October 5, 2005.
Description of amendment request: The requested change will delete
Technical Specification (TS) 5.6.1,
[[Page 154]]
``Occupational Radiation Exposure Report,'' and TS 5.6.4, ``Monthly
Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated October 5, 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Lauder.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: November 7, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.9.3, ``Containment Penetrations,'' to
allow an emergency egress door, access door, or roll up door, as
associated with the equipment hatch penetration, to be open, but
capable of being closed, during core alterations or movement of
irradiated fuel within containment.
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 change has no impact on the probability of a FHA [fuel-
handling accident] inside containment. It merely allows the transfer
of equipment and personnel through the equipment hatch, and allows
parallel activities. The refueling operations have spatial
separation from the open hatch precluding interaction with
refueling. Having the equipment hatch open will not impact the
operation or operability of refueling equipment or the performance
of the refueling crew.
Per [Regulatory Guide 1.183, ``Alternative Radiological Source
Terms for Evaluating Design Basis Accidents at Nuclear Power
Reactors''], the analysis was performed assuming a two hour release
of radioactivity with the hatch open for the entire duration. An
analysis assuming a closed hatch was not performed for comparison.
This change merely allows plant conditions to exist that are assumed
in the analysis. The relatively small off-site dose values shown in
Section 4 [of the November 7 application], and the additional
conservatism provided by the requirement for administrative closure
capability, demonstrates that any consequence to the public
resulting from this change would be minimal.
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 change more closely aligns the allowed plant conditions with
those conditions assumed in an existing (analyzed) accident.
Allowing movement of equipment through the equipment hatch during
core alterations does not create any new accident initiators. Given
the plant conditions, it does not affect system operation or the
functions they perform. 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 change does not create conditions different from or less
conservative than, those assumed in the analysis, and is consistent
with the regulatory guidance for performing that analysis.
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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Branch Chief: Richard J. Lauder.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: November 18, 2005.
Description of amendment request: The proposed amendment would
revise the frequency in Technical Specification Surveillance
Requirement (SR) 3.6.6.15, which verifies that each containment spray
nozzle is unobstructed. The frequency would be changed from ``10
years'' to ``following maintenance which could result in nozzle
blockage.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the SR to verify that the
Containment Spray System nozzles are unobstructed after maintenance
that could introduce material that could result in nozzle blockage.
The spray nozzles are not assumed to be initiators of any previously
analyzed accident. Therefore, the change does not increase the
probability of
[[Page 155]]
any accident previously evaluated. The spray nozzles are assumed in
the accident analyses to mitigate design basis accidents. The
revised SR to verify system OPERABILITY following maintenance is
considered adequate to ensure OPERABILITY of the Containment Spray
System. Since the system will still be able to perform its accident
mitigation function, the consequences of accidents previously
evaluated are not increased. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the SR to verify that the
Containment Spray System nozzles are unobstructed after maintenance
that could result in nozzle blockage. The change does not introduce
a new mode of plant operation and does not involve physical
modification to the plant. The change will not introduce new
accident initiators or impact the assumptions made in the safety
analysis. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the frequency for performance of the
SR to verify that the Containment Spray System nozzles are
unobstructed. The frequency is changed from every 10 years to
following maintenance that could result in nozzle blockage. This
requirement, along with foreign material exclusion programs and the
remote physical location of the spray nozzles, provides assurance
that the spray nozzles will remain unobstructed. As the spray
nozzles are expected to remain unobstructed and able to perform
their post-accident mitigation function, plant safety is not
significantly affected. 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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Branch Chief: Richard J. Lauder.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: December 6, 2005.
Description of amendment requests: The proposed amendment will
delete Technical Specification (TS) Limiting Condition for Operation
(LCO) 3.3.10, ``Fuel Handling Isolation Signal (FHIS),'' and TS LCO
3.7.14, ``Fuel Handling Building Post-Accident Cleanup Filter System,''
and their associated Surveillance Requirements. The proposed amendment
will also delete the Fuel Handling Building Post-Accident Cleanup
Filter Systems from the Ventilation Filter Testing Program in
administrative TS 5.5.2.12.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Fuel Handling Building (FHB) Post-Accident Cleanup Filter
System (PACFS) and its initiating radiation monitors are not
involved in the initiation of any accidents. The PACFS is not
credited with providing any supplemental filtration of releases from
an accident occurring in the FHB. The PACFS was designed to provide
an accident mitigation function by isolating the system and
filtering the radioiodines that may be released from a damaged fuel
assembly in the event of a Fuel Handling Accident (FHA). The
charcoal adsorber was the primary component that supported this
filtration function. However, the FHA dose consequences analysis has
demonstrated that doses due to the FHA, to both the public and the
control room operators, remain well within regulatory acceptance
limits even assuming no credit for either isolation or filtration.
The charcoal filtration function is not required and need not be
tested. Thus, there is no required safety function provided by
either the ventilation system or the airborne radiation monitor in
the event of a fuel handling accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The FHB PACFS and its initiating radiation monitors do not
initiate any accidents. The PACFS was designed to provide an
accident mitigation function by isolating the system and filtering
the radioiodines that may be released from a damaged fuel assembly
in the event of a Fuel Handling Accident. Analysis shows that the
isolation and filtration functions are not required. The charcoal
adsorber cannot influence any accident initiators. The deletion of
the Technical Specification requirements does not impact this
conclusion and does not influence any new potential accident
scenarios in any way.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The FHB PACFS and its initiating radiation monitors were
designed to provide an accident mitigation function by filtering the
radioiodines that may be released from a damaged fuel assembly in
the event of a Fuel Handling Accident. Analysis of the FHA in the
FHB demonstrates that the margin of safety provided by the Technical
Specification requirement will not change. Since the control room
charcoal adsorber is capable of accommodating the design[-]basis
loss[-]of[-]coolant accident fission product halogen loadings, which
are more limiting than the fuel handling accident loadings, [a] more
than adequate design margin is available with respect to postulated
FHA releases. The margin of safety, in terms of the dose limitations
of 10 CFR part 100 and 10 CFR part 50[,] Appendix A, General Design
Criterion 19, has not been significantly reduced.
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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: David Terao.
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: July 21, 2005.
Description of amendment request: The proposed change would revise
the accident monitoring instrumentation listing, the allowed outage
times (AOTs) to be consistent with the requirements of the Improved
Technical Specifications (ITS) for post accident monitoring
instrumentation. TS 3.7E, TS Table 3.7-6, and TS Table 4.1-2 would be
affected by this change.
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 156]]
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change revises the [AOTs] and requirements for
accident monitoring instrumentation. The proposed change expands the
instrumentation listing in the Technical Specifications to include
the Category 1 RG [Regulatory Guide] 1.97 variables and deletes the
Category 2 RG 1.97 variables, which are addressed in a licensee
controlled document. The revise requirements continue to require the
accident monitoring instrumentation to be operable. The required
operability will continue to ensure that sufficient information is
available on selected unit parameters to monitor and assess unit
status and response during and following an accident. Accident
monitoring instrumentation is not an initiator of any accident
previously evaluated. The consequences of an accident during the
extended [AOTs] would be the same as the consequences during the
current [AOTs]. Therefore, the proposed change does not involve a
significant increase in either 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 identified.
The proposed change involves no physical changes to the plant,
nor is there any impact on the design of the plant or the accident
monitoring instrumentation. There is also no impact on the
capability of the instrumentation to provide post accident data for
plant operator use, the accident monitoring instrumentation
initiates no automatic action, and there is no change in the
likelihood that the instrumentation will fail since surveillance
tests will continue to be performed. Therefore, the proposed change
does not introduce any new failures that could create the
possibility of a new or different kind of accident from any accident
previously identified.
3. Involve a significant reduction in a margin of safety.
The proposed change provides more appropriate times to restore
inoperable accident monitoring instrumentation to operable status
and does not impact the level of assurance that the instrumentation
will be available to perform its function. Accident monitoring
instrumentation has been screened out of the probabilistic risk
analysis (PRA) model due to its low risk significance, so the
proposed change has no risk impact from a PRA perspective. The
proposed change does not alter the condition or performance of
equipment or systems used in accident mitigation or assumed in any
accident analysis. Therefore, this 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: 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 Branch Chief: Evangelos C. Marinos.
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: September 13, 2005.
Description of amendment request: The proposed change would change
the exclusion area boundary (EAB), reduce the design-basis accident
(DBA) Atmospheric Dispersion Factor (X/Q), and reduce the calculated
EAB dose consequences for accidents described in Chapter 14 of the
Updated Final Safety Analysis Report (UFSAR).
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 redefinition of the EAB will significantly reduce
the design basis accident X/Q, which will result in an increase in
margin to the dose consequence limits for future accident analyses.
The dose consequence accident analyses were not reanalyzed with this
change because the EAB results currently documented in the UFSAR are
conservative with respect to consequences that would be calculated
using this redefined EAB. The EAB redefinition is not an initiator
of any accident previously evaluated and has no impact on radiation
levels, airborne activity, DBA source terms, or releases.
Therefore, the proposed change does not involve a significant
increase in either 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 identified.
The proposed change involves no physical changes to the plant,
nor is there any impact on the design or operation of the plant.
There is also no impact on any equipment relied upon to mitigate an
accident. Therefore, the proposed change does not introduce any new
failures that could create the possibility of a new or different
kind of accident from any accident previously identified.
3. Involve a significant reduction in a margin of safety.
The proposed change does not alter the condition or performance
of equipment or systems used in accident mitigation or assumed in
any accident analysis. The EAB redefinition has no impact on
radiation levels, airborne activity, DBA source terms, or releases.
Therefore, this proposed change does not involve a significant
reduction in the [a] margin of safety. However, the proposed
redefinition of the EAB will significantly reduce the design basis
accident X/Q, which will result in an increase in margin to the dose
consequence limits for future accident analyses.
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: 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 Branch Chief: Evangelos C. Marinos.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 27, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16,
``RCS [reactor coolant system] Specific Activity.'' The revisions would
replace the current Limiting Condition for Operation (LCO) 3.4.16 limit
on RCS gross specific activity with limits on RCS Dose Equivalent I-131
and Dose Equivalent XE-133 (DEX). The conditions and required actions
for LCO 3.4.16 not being met, and surveillance requirements for LCO
3.4.16, are being revised. The modes of applicability for LCO 3.4.16
would be extended. The current definition of [Emacr]--Average
Disintegration Energy in TS 1.1 would be replaced by the definition of
DEX. In addition, the current definition of Dose Equivalent I-131 in TS
1.1 would be revised to allow alternate, NRC-approved thyroid dose
conversion factors.
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.
Response: No.
The proposed changes would add new thyroid dose conversion
factor reference[s] to
[[Page 157]]
the definition of DOSE EQUIVALENT I-131, eliminate the definition of
[Emacr]-AVERAGE DISINTEGRATION ENERGY, add a new definition of DOSE
EQUIVALENT XE-133, replace the Technical Specification (TS) 3.4.16
limit on reactor coolant system (RCS) gross specific activity with a
limit on noble gas specific activity in the form of a Limiting
Condition for Operation (LCO) on DOSE EQUIVALENT XE-133, replace TS
Figure 3.4.16-1 with a maximum limit on DOSE EQUIVALENT I-131,
extend the Applicability of LCO 3.4.16, and make corresponding
changes to TS 3.4.16 to reflect all of the above. The proposed
changes are not accident initiators and have no impact on the
probability of occurrence of any design[-]basis accidents.
The proposed changes will have no impact on the consequences of
a design[-]basis accident because they will limit the RCS noble gas
specific activity to be consistent with the values assumed in the
radiological consequence analyses. The changes will also limit the
potential RCS [radio]iodine concentration excursion to the value
currently associated with full power operation, which is more
restrictive on plant operation than the existing allowable RCS
[radio]iodine specific activity at lower power levels.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Response: No.
The proposed changes do not alter any physical part of the plant
nor do they affect any plant operating parameters besides the
allowable specific activity in the RCS. The changes which impact the
allowable specific activity in the RCS are consistent with the
assumptions assumed in the current radiological consequence
analyses. [The proposed changes are also not 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. The proposed change does not involve a significant reduction
in a margin of safety.
Response: No.
The acceptance criteria related to the proposed changes involve
the allowable control room and offsite radiological consequences
following a design[-]basis accident. The proposed changes will have
no impact on the radiological consequences of a design[-]basis
accident because they will limit the RCS noble gas specific activity
to be consistent with the values assumed in the radiological
consequence analyses. The changes will also limit the potential RCS
[radio]iodine specific activity excursion to the value currently
associated with full power operation, which is more restrictive on
plant operation than the existing allowable RCS [radio]iodine
specific activity at lower power levels.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
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].
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: October 12, 2004, as
supplemented by March 4 and August 4, 2005.
Brief description of amendment: The license amendment changes the
Final Safety Analysis Report (FSAR) to reflect that the reactor core
isolation cooling (RCIC) system is not required to mitigate the
consequences of the control rod drop accident (CRDA). The FSAR revision
clarifies that although the RCIC system is designed to initiate and
inject into the reactor pressure vessel (RPV) at a low water level
(L2), the additional RPV inventory is not required to prevent the
accident or to mitigate the consequences of the CRDA.
Date of issuance: December 14, 2005.
Effective date: This license amendment is effective as of the date
of its issuance, and shall be implemented within 60 days.
Amendment No.: 196.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 2004 (69 FR
64987).
The supplemental letters dated March 4 and August 4, 2005, provided
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 December 14, 2005.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: August 17, 2005.
Brief description of amendment: The amendment allows a one-time
extension of the 72-hour Completion Time (CT) for the required action
of Condition B of Technical Specification (TS) 3.7.1, ``Standby Service
Water (SW) System and Ultimate Heat Sink (UHS),'' and of TS 3.8.1, ``AC
Sources--Operating.''
[[Page 158]]
Specifically, the proposed one-time extension request is for an
additional 72 hours to the CT and would result in a 144-hour CT for an
inoperable SW subsystem. This would allow extensive maintenance, not
capable of being completed in the current 72-hour CT, to be conducted
on the SW train B pump.
Date of issuance: December 8, 2005.
Effective date: The license amendment is effective as of its date
of issuance.
Amendment No.: 195.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 27, 2005 (70
FR 56501)
The November 15 and 30, 2005, supplemental letters provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 8, 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, Beaver County,
Pennsylvania; FirstEnergy Nuclear Operating Company, et al., Docket No.
50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio;
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of application for amendments: May 18 and June 1, 2005, as
supplemented by letters dated July 15 and October 31, 2005.
Brief description of amendments: The conforming amendments
implement the direct license transfers of the Facility Operating
Licenses for Beaver Valley Power Station, Units 1 and 2, Davis-Besse
Nuclear Power Station, Unit 1, and Perry Nuclear Power Plant, Unit 1,
to the extent held by Pennsylvania Power Company, Ohio Edison Company,
OES Nuclear, Inc., the Cleveland Electric Illuminating Company, and the
Toledo Edison Company, with respect to their current ownership
interests, to FirstEnergy Nuclear Generation Corporation, a new nuclear
generation subsidiary of FirstEnergy Corporation.
Date of issuance: December 16, 2005.
Effective date: As the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos. for License Nos. DPR-66 and NPF-73: 269 and 151.
Amendment Nos. for License No. NPF-3: 270.
Amendment Nos. for License No. NPF-58: 137.
Facility Operating License Nos. DPR-66, NPF-73, NPF-3, and NPF-58:
Amendments revised the Licenses.
Date of initial notice in Federal Register: August 2, 2005 (70 FR
44390-44395).
The supplements dated July 15 and October 31, 2005 clarified the
application, did not expand the scope of the application as originally
noticed.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 16, 2005.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 20, 2005.
Brief description of amendment: The amendment revises Cooper
Nuclear Station TS 5.3, Unit Staff Qualifications, to upgrade the
qualification standard for the shift manager, senior operator, licensed
operator, and shift technical engineer from Regulatory Guide 1.8,
``Qualification and Training of Personnel for Nuclear Power Plants,''
Revision 2, April 1987, to Regulatory Guide 1.8, Revision 3, May 2000.
It also clarifies qualification requirements applicable to the
operations manager position.
Date of issuance: December 15, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 214.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59085).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 15, 2005.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of application for amendment: July 9, 2004, as supplemented by
letters dated July 9, 2004, August 17, 2004, and June 3, 2005.
Brief description of amendment: The amendment authorizes the use of
the Holtec davit crane in the refueling building for cask handling
operations.
Date of issuance: December 15, 2005.
Effective date: December 15, 2005, and shall be implemented within
60 days of issuance.
Amendment No.: 37.
Facility Operating License No. DPR-7: This amendment revises the
licensing basis.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70721).
The July 9, 2004, August 17, 2004, and June 3, 2005, supplemental
letters provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff original no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 15, 2005.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: March 10, 2005, as supplemented
on June 8 and August 31, 2005.
Brief description of amendment: The amendment revises Technical
Specification 5.5.15, ``Containment Leakage Rate Testing Program,'' to
extend, on a one-time basis, the interval for completing the next
containment integrated leakage rate test, pursuant to Appendix J to
Part 50 of Title 10 of the Code of Federal Regulations, from 10 years
to 15 years since the last test. Therefore, the first test performed
after the May 31, 1996, test shall be performed by May 31, 2011.
Date of issuance: December 8, 2005.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 93.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33217).
The June 8 and August 31, 2005, letters provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a
[[Page 159]]
Safety Evaluation dated December 8, 2005.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of application for amendment: June 22, 2005.
Brief description of amendment: This amendment for Virgil C. Summer
replaces the current reactor coolant system pressure-temperature limits
for 32 effective full power years with the proposed limits for 56
effective full power years.
Date of issuance: December 13, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 174.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: September 27, 2005 (70
FR 56504).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 13, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: May 25, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications to adopt the provisions of Industry/TS Task
Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: December 13, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 246/190.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48207).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 13, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: April 26, 2004, as supplemented
by letters dated April 18 and July 22, 2005.
Brief description of amendments: The amendments revised the Units 1
and 2 Technical Specifications Limiting Condition for Operation 3.7.9,
``Ultimate Heat Sink (UHS),'' to allow plant operation with three fans
and four spray cells in the Nuclear Service Cooling Water system under
certain atmospheric conditions.
Date of issuance: December 2, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 140 and 119.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 2004 (69 FR
43462).
The supplements dated April 18 and July 22, 2005, provided
clarifying information that did not change the scope of the April 26,
2004, application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 2, 2005.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 23rd day of December, 2005.
For the Nuclear Regulatory Commission.
Edwin M. Hackett,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 05-24669 Filed 12-30-05; 8:45 am]
BILLING CODE 7590-01-P