[Federal Register Volume 70, Number 233 (Tuesday, December 6, 2005)]
[Notices]
[Pages 72667-72681]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-23553]
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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 to 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 November 9, 2005 to November 21, 2005. The
last biweekly notice was published on November 22, 2005 (70 FR 70641).
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 72668]]
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 72669]]
4209, (301) 415-4737 or by e-mail to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: July 13, 2005.
Description of amendments request: The proposed amendment would
revise Technical Specification (TS) 1.1, ``Definitions,'' TS 3.4.13,
``RCS [reactor coolant system] Operational Leakage,'' TS 5.5.9, ``Steam
Generator Tube Surveillance Program,'' and TS 5.6.9, ``Steam Generator
Tube Inspection Report,'' and add a new specification (TS 3.4.18) for
Steam Generator (SG) Tube Integrity. The proposed changes are necessary
in order to implement the guidance for the industry initiative on
Nuclear Energy Institute (NEI) 97-06, ``Steam Generator Program
Guidelines.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting Technical Specification Task Force Change Traveller 449,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on May 6, 2005 (70 FR 24126). The
licensee affirmed the applicability of the following NSHC determination
in its application dated July 13, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change requires a SG Program that includes
performance criteria that will provide reasonable assurance that the
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.
A SGTR [steam generator tube rupture] 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 operational LEAKAGE rate limits
in the licensing basis plus the LEAKAGE rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as MSLB [main steam line
break], rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. 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 to the TS 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 SG Program required by the proposed change to the TS. 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 proposed
changes do not, therefore, significantly increase the probability of
an accident previously evaluated.
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 [500 gallons per
day or 720 gallons per day] in any one SG, and that the reactor
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS
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 TSs 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 TSs.
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 an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG 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 SG 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 [kind] of accident from any accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety
The 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
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical 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 SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff proposes to determine that the amendments request
involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
[[Page 72670]]
NRC Branch Chief: Richard J. Laufer.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: October 31, 2005.
Description of amendment request: The proposed amendment change
would add Technical Specification (TS) Limiting Condition for Operation
(LCO) 3.0.8, to allow a delay time for entering a supported system TS
when the inoperability is due solely to an inoperable snubber, if risk
is assessed and managed consistent with the program in place for
complying with the requirements of 10 CFR 50.65(a)(4). In addition, a
proposed change to LCO 3.0.1 is required to reference the addition of
LCO 3.0.8.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated as TSTF-372, Revision 4. The NRC
staff issued a notice of opportunity for comment in the Federal
Register on November 24, 2004 (69 FR 68412), on possible amendments
concerning TSTF-372, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on May 4, 2005
(70 FR 23252). The licensee affirmed the applicability of the following
NSHC determination in its application dated October 31, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low-probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident while relying on allowance provided by proposed LCO 3.0.8
are no different than the consequences of an accident while relying
on the TS required actions in effect without the allowance provided
by proposed LCO 3.0.8. Therefore, the consequences of an accident
previously evaluated are not significantly affected by this change.
The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's performance of a risk assessment and
the management of plant risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Branch Chief: L. Raghavan.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: October 3, 2005.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) Surveillance Requirements (SRs) to reflect
changes to the Emergency Core Cooling System throttle valves. The
proposed amendment will add the modified throttle valves to the
surveillance, remove existing throttle valves that are now locked
closed from the surveillance, and add existing valves to the
surveillance that are used in a throttle position when open.
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 to Surveillance Requirement (SR) 3.5.2.6
adds nine valves and removes two valves in the High Head Safety
Injection (HHSI) system discharge lines. The SR requires
verification that identified ECCS [emergency core cooling system]
throttle valves position stops are in the correct position. The
change reflects a stretch power uprate (SPU) modification that added
throttle valves SI-2165, 2166, 2168, 2169, 2170, 2171, and 2172, and
locked closed valves Sl-856A and 856F. This amendment is adding to
the SR those throttle valves which are now under administrative
control and deletes the valves which no longer perform a throttle
function. The amendment also adds hot leg valves Sl-856B and 856G
which are used as throttle valves but never included in the SR.
Valve Sl-856G still performs a throttle function and valve SI-856B
can still be considered a throttle valve when used to trim system
resistance. Verification of valve position has no effect on the
probability of an accident previously evaluated since the HHSI
system is not associated with the initiation of any accident. The
verification of valve positions that will be required by the revised
SR provides additional assurance that the HHSI throttle valves are
in the position that is established by flow testing. Providing
assurance of required valve positions does not increase the
consequences of an accident previously evaluated.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 to Surveillance Requirement 3.5.2.6 adds
nine valves and removes two valves in the High Head Safety Injection
(HHSI) system discharge lines. The SR requires verification that
identified ECCS throttle valves position stops are in the correct
position. The change corrects a deficient surveillance and does not
affect the function of the valves or otherwise affect the design and
operation of plant systems and components and therefore no new
accident
[[Page 72671]]
scenarios would be created. Therefore, no new failure modes are
being introduced that could lead to different accidents.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to Surveillance Requirement 3.5.2.6 adds
nine valves and removes two valves in the High Head Safety Injection
(HHSI) system discharge lines. The SR requires verification that
identified ECCS throttle valves position stops are in the correct
position. The change reflects a stretch power uprate (SPU)
modification that added throttle valves SI-2165, 2166, 2168, 2169,
2170, 2171, and 2172, and locked closed valves Sl-856A and 856F. The
proposed amendment also adds valves SI-856B and 856G which are used
as throttle valves but never included in the SR. Valve Sl-856G still
performs a throttle function and valve Sl-856B can still be
considered a throttle valve when used to trim system resistance. The
frequency for verification of throttle valve stop positions is not
altered by this amendment so this has no effect on the margin of
safety. The valves for which verification of positions stops is
required reflect the manner in which the system is currently
analyzed and configured so the proposed change serves to maintain
the required margin of safety by adding to the Technical
Specifications the surveillances presently being administratively
controlled. Therefore, the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Richard J. Laufer.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: June 29, 2005.
Description of amendment request: Entergy Operations, Incorporated
(Entergy) proposes to relocate the License Condition associated with
the Shutdown Cooling (SDC) Open Permissive Interlock (OPI) to the
Technical Requirements Manual (TRM). The Nuclear Regulatory Commission
(NRC) approved Standard Technical Specifications, Combustion
Engineering Plants (NUREG-1432) include a surveillance requirement for
this function due to the complexity and differences of plant designs,
which would not support complete removal of the function from the
NUREG. For Arkansas Nuclear One, Unit 2 (ANO-2), however, the OPI is
not an assumed function that supports the accident analysis and does
not meet the criteria in Section 50.36 of Title 10 of the Code of
Federal Regulations (10 CFR) for inclusion in the technical
specifications.
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 OPI function is not required to ensure continued safe
operation of the ANO-2 facility. The OPI function is not considered
an accident precursor or relied upon as a means of accident
mitigation. The proposed change has no affect on plant design or
operation.
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 relocation of the OPI function to the TRM does not require
any physical alteration to the plant or alter plant design. The OPI
function is not considered an accident initiator nor is this
function credited in any safety analyses for the prevention or
mitigation of any accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The OPI function is not credited in a margin of safety analysis
for any accident previously evaluated. Relocation of the OPI
function requirements will not result in a credible increase in
nuclear safety risk. Appropriate alarm, design features, and
administrative controls continue to ensure proper isolation of the
SDC system during plant operations with elevated RCS [reactor
cooling system] pressures. In addition, the OPI function will be
relocated to the TRM, which is part of the Safety Analysis Report
(SAR) and controlled by 10 CFR 50.59.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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: The proposed change will modify
the Surveillance Requirements related to Arkansas One, Unit 2,
technical specification (TS) 3.1.1.4, Moderator Temperature Coefficient
(MTC), and will allow the use of topical report WCAP-16011-P-A,
``Startup Test Activity Reduction Program.'' A change to NUREG-1432,
``Standard Technical Specifications Combustion Engineering Plants,''
has been proposed in Technical Specification Task Force (TSTF) Improved
Standard Technical Specification Change Traveler TSTF-486 to
incorporate the allowance to use WCAP-16011-P-A. The traveler was
submitted for Nuclear Regulatory Commission (NRC) approval in June
2005.
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 MTC is not an initiator of any previously evaluated
accidents. As an input into accident analyses, the MTC is used to
predict plant behavior in the event of an accident. It was
demonstrated in WCAP-16011-P-A that the modified MTC verification
(i.e., measured RCS [reactor coolant system] boron concentration) is
adequate to ensure that the MTC remains within the limits provided
the STAR applicability requirements are met. It was also
demonstrated in WCAP-16011-P-A that the elimination of the EOC
[emergency operations center] MTC measurement is acceptable when the
applicability requirements given in WCAP-16011-P-A are met and the
result of the MTC determination performed prior to reaching a Rated
Thermal Power equilibrium boron concentration of 800 ppm is within a
tolerance of 0.16 x 10-4 Dk/k/
[deg]F from the corresponding design value.
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
[[Page 72672]]
accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant (i.e., no new or different type of structure, system, or
component will be installed).
The methods governing normal plant operations are not altered by
the proposed change.
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 will not affect the margin of safety. The
MTC limits are unaffected and an acceptable method will be used to
demonstrate that MTC is within its limits.
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 Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2 (ANO-2), Pope County, Arkansas
Date of amendment request: September 19, 2005.
Description of amendment request: The proposed change will modify
the ANO-2 technical specification (TS) 3.1.1.5, Minimum Temperature for
Criticality. Specifically, the proposed change will raise the minimum
temperature for criticality from the current value of 3 525
[deg]F to 3 540 [deg]F. Changes are also proposed to the
Action statement and Surveillance Requirements to support the increase
in temperature. The change is needed to support core design.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
There are no accident analyses that dictate the minimum
temperature for criticality. The minimum temperature for criticality
is not an accident initiator. It is used in the reload analyses as a
limiting temperature at which the core design is verified to satisfy
the limit of the positive moderator temperature coefficient (MTC)
specified in the ANO-2 TS and Core Operating Limits Report (COLR).
The minimum temperature for criticality is one of many input
parameters used in the reload design analytical calculation that
confirms the core design satisfies the MTC TS and COLR.
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 to increase the minimum temperature for
criticality does not result in any plant design changes. In
addition, the minimum temperature at which the reactor is taken
critical is not an accident initiator. The nominal average reactor
coolant system temperature during an approach to criticality is
several degrees higher than the limit proposed for the minimum
temperature for criticality.
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 increase of the minimum temperature for criticality in
conjunction with the use of a sufficient number of burnable absorber
rods, which will be incorporated into the core design, will ensure
the current TS limits, as reflected in the COLR, for the most
positive MTC will continue to be satisfied.
The current transient analysis results are bounding and remain
applicable.
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 Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 25, 2005.
Description of amendment request: The proposed change will modify
the Waterford 3 Technical Specification (TS) 3.1.1.4, Minimum
Temperature for Criticality. Specifically, the proposed change will
raise the minimum temperature for criticality from the current value of
>=520[deg]F to >=533[deg]F. Changes are also proposed to the Action
statement and Surveillance Requirements to support the increase in
temperature.
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 minimum temperature for criticality is not an accident
initiator. It is used in the reload analyses as a limiting
temperature at which the core design is verified to satisfy the
limit of the positive moderator temperature coefficient (MTC)
specified in the Waterford 3 TS and Core Operating Limits Report
(COLR). The minimum temperature for criticality is one of many input
parameters used in the reload design analytical calculation that
confirms the core design satisfies the MTC TS and COLR.
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 to increase the minimum temperature for
criticality does not result in any plant design changes. In addition
the minimum temperature at which the reactor is taken critical is
not an accident initiator. The nominal average reactor coolant
system temperature during an approach to criticality is several
degrees higher than the limit proposed for the minimum temperature
for criticality.
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 increase of the minimum temperature for criticality in
conjunction with the appropriate core designs will ensure the
current TS limits, as reflected in the COLR, for the most positive
MTC will continue to be satisfied.
The current transient analysis results are bounding and remain
applicable.
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
[[Page 72673]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 25, 2005.
Description of amendment request: The proposed change will modify
the Surveillance Requirements (SRs) related to Waterford 3 Technical
Specification (TS) 3.1.1.3, Moderator Temperature Coefficient (MTC) and
will allow the use of the Startup Test Activity Reduction Program
(WCAP-16011-P-A).
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 MTC is not an initiator of any previously evaluated
accidents. As an input into accident analyses, the MTC is used to
predict plant behavior in the event of an accident. It was
demonstrated in WCAP-16011-P-A that the modified MTC verification
(i.e., measured RCS [reactor coolant system] boron concentration) is
adequate to ensure that the MTC remains within the limits, provided
the STAR applicability requirements are met. It was also
demonstrated in WCAP-16011-P-A that the elimination of the EOC [end-
of-cycle] MTC measurement is acceptable when the applicability
requirements given in WCAP-16011-P-A are met and the result of the
MTC determination performed at greater than 15 percent of Rated
Thermal Power and prior to reaching 40 EFPD [effective full power
days] is within a tolerance of 0.16 x 10-4
[Delta]k/k/[deg]F from the corresponding design value.
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 to
the plant (i.e., no new or different type of structure, system, or
component will be installed). The methods governing normal plant
operations are not altered by the proposed TS change.
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 will not affect the margin of safety. The
MTC limits are unaffected and an acceptable method will be used to
demonstrate that MTC is within its limits.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 25, 2005.
Description of amendment request: The proposed change to Technical
Specification 6.9.1.11, Core Operating Limits Report, will result in
the addition of a methodology that will allow the use of zirconium
diboride (ZrB2) burnable absorber coating on fuel pellets.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will add topical report WCAP-16072-P-A to
the NRC reviewed and approved analytical methods used to determine
the core operating limits. The topical report has been previously
approved by the NRC for use in Combustion Engineering core designs
and as such, the proposed change is administrative in nature and has
no impact on any plant configurations or on system performance that
is relied upon to mitigate the consequences of an accident. In
addition, prior to the use of the ZrB2 burnable absorber
coating, fuel design will be analyzed with applicable NRC staff
approved codes and methods.
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 adds a reference to the topical report that
allows the use of ZrB2 as a burnable absorber coating on
the fuel pellet. The topical report has been previously approved by
the NRC for use in Combustion Engineering core designs and as such,
the proposed change is administrative in nature and has no impact on
any plant configurations or on system performance that is relied
upon to mitigate the consequences of an accident. In addition, prior
to the use of the ZrB2 burnable absorber coating, fuel
design will be analyzed with applicable NRC staff approved codes and
methods. This change is administrative in nature and does not create
a new or different type of accident than previously evaluated
because the design requirements for the facility remain the same.
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 will add WCAP-16072-P-A to the list of
referenced topical reports. The topical report has been previously
approved by the NRC for use in Combustion Engineering core designs
and as such, the proposed change is administrative in nature and has
no impact on any plant configurations or on system performance that
is relied upon to mitigate the consequences of an accident. In
addition, prior to the use of the ZrB2 burnable absorber
coating, fuel design will be analyzed with applicable NRC staff
approved codes and methods.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: July 29, 2005.
Description of amendment requests: The proposed amendments would
delete requirements from the Technical Specifications (TSs) to submit
monthly operating reports and annual occupational radiation exposure
reports. The changes are consistent with
[[Page 72674]]
Revision 1 of Nuclear Regulatory Commission (NRC) approved Industry/
Technical Specifications Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-369, ``Removal of Monthly Operating
and Occupational Radiation Exposure Report.'' The availability of this
TS improvement was announced in the Federal Register (69 FR 35067) on
June 23, 2004, as part of the Consolidated Line Item Improvement
Process (CLIIP).
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 April 29,
2004 (69 FR 23542). The licensee affirmed the applicability of the
model NSHC determination in its application dated July 29, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC (which was previously published in 69 FR 23542) 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
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 Technical Specification 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.
Based on the reasoning presented above and the previous discussion
of the amendment request, the NRC staff proposes to determine that the
requested change does not involve a significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106.
NRC Branch Chief: L. Raghavan.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: August 10, 2005.
Description of amendment requests: The proposed amendments would
delete the power range neutron flux high negative rate trip function
from each unit's Technical Specifications. The licensee's proposed
changes are based on the methodology presented in Westinghouse Topical
Report WCAP-11394-P-A, ``Methodology for the Analysis of the Dropped
Rod Event,'' which had been previously accepted by the Nuclear
Regulatory Commission staff.
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The removal of the power range neutron flux high negative rate
trip function from technical specifications does not increase the
probability or consequences of reactor core damage accidents
resulting from dropped Rod Cluster Control Assembly (RCCA) events
previously analyzed. The safety functions of other safety-related
systems and components, which are related to mitigation of these
events, [will] not [be] altered. All other Reactor Trip System and
Engineered Safety Features Actuation Systems protection functions
are not impacted by the elimination of the trip function. The
dropped RCCA accident analysis does not rely on the negative flux
rate trip to safely shut down the plant. The safety analysis of the
plant is unaffected by the proposed change. Since the safety
analysis is unaffected, the calculated radiological releases
associated with the analysis are not affected.
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 adversely alter the design
assumptions, conditions, or configuration of the facility or the
manner in which the plant is operated. No new accident scenarios,
failure mechanisms, or limiting single failures are introduced as a
result of the proposed change. The proposed change does not
challenge the performance or integrity of any safety-related systems
or components. Nuclear Regulatory Commission (NRC)-approved
Westinghouse Topical Report WCAP-11394-P-A, ``Methodology for the
Analysis of the Dropped Rod Event,'' dated January 1990 has
demonstrated that the negative flux rate trip function can be
eliminated.
Therefore, the proposed changes does not created the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. It has been demonstrated that the
negative flux rate trip function can be eliminated by the NRC-
approved methodology described in WCAP-11394-P-A. Donald C. Cook
Nuclear Plant cycle-specific analyses have confirmed that for a
dropped RCCA(s) event, limits on departure from nucleate boiling are
not exceeded by eliminating the negative flux rate trip. The
proposed change will have no [e]ffect on the availability,
operability, or performance of safety-related systems and
components.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106.
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: August 11, 2005.
Description of amendment request: The proposed change allows a
delay time for entering a supported system Technical Specification (TS)
when the inoperability is due solely to an inoperable snubber, if risk
is assessed and managed consistent with the program in place for
complying with the requirements of Paragraph 50.65(a)(4) of Title 10 of
the Code of Federal
[[Page 72675]]
Regulations. Limiting Condition for Operation (LCO) 2.0.1(3) is added
to the TS to provide this allowance and define the requirements and
limitations for its use.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
November 24, 2004 (69 FR 68412), on possible amendments concerning
TSTF-372, including a model safety evaluation and model no significant
hazards consideration (NSHC) determination, using the consolidated line
item improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on May 4, 2005 (70 FR 23252). The
licensee affirmed the applicability of the following NSHC determination
in its application dated August 11, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low-probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident while relying on allowance provided by proposed LCO 3.0.8
[LCO 2.0.1(3) for Fort Calhoun Station] are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.8 [LCO
2.0.1(3)]. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG [Regulatory Guide] 1.177. A bounding risk
assessment was performed to justify the proposed TS changes. This
application of LCO 3.0.8 is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
[The proposed LCO 2.0.1(3) defines limitations on the use of the
provision and includes a requirement for the licensee to assess and
manage the risk associated with operation with an inoperable
snubber.] The net change to the margin of safety is insignificant.
Therefore, this change does not involve a significant reduction in a
margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Branch Chief: David Terao.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 8, 2005.
Description of amendment request: The proposed amendment will
modify Fort Calhoun Technical Specification (TS) 4.2.1, ``Fuel
Assemblies,'' to permit the use of AREVA (Framatome ANP)
M5TM advanced alloy for fuel rod cladding and structural
components such as guide tubes, intermediate spacer grids, end plugs,
and guide thimble tubes, beginning with Cycle 24. In addition, Omaha
Public Power District proposes to modify TS 5.9 to include the
Framatome ANP Topical Report evaluating the impact of M5TM
material properties on NRC-approved methodology. M5TM is a
proprietary, zirconium-based alloy that is a variant of Zr1Nb to
replace zircaloy-4 in the construction of fuel assembly components.
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 NRC[-]approved topical report BAW-10[2]27P-A (Reference 8.1
[of amendment request]) that provides the licensing basis for
M5TM cladding and structural material, has shown that the
M5TM alloy exhibits superior properties to the currently
used zircaloy-4 material. The cladding by itself does not initiate
an accident and therefore does not affect accident probability. It
has been determined that M5TM cladding will not
significantly affect the consequences of an accident.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously analyzed.
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. Topical report BAW-10227P-
A (Reference 8.1) demonstrated that the M5TM alloy will
perform similar to or better than zircaloy-4, thus precluding the
possibility of the fuel becoming an accident initiator and causing a
new or different type of accident.
Since the material properties of M5TM alloy are
similar to or better than zircaloy-4, there will not be any
significant change in the types of effluents that may be released
off-site. There will not be any significant increase in occupational
or public radiation exposure.
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 generic LOCA [loss-of-coolant accident] and
non-LOCA evaluations and demonstrated the use of the M5TM
material will have only a small, or beneficial, impact on the event
consequences.
Plant-specific analyses using NRC-approved methodology for the
mixed core will demonstrate that the reactor core safety limits will
continue to be met.
[[Page 72676]]
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.
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: November 3, 2005.
Description of amendment requests: The proposed amendment revises
Technical Specification (TS) Section 5.5.2.11 to modify the definitions
of steam generator tube ``Repair Limit'' and ``Tube Inspection.'' The
purpose of these changes is to define the extent of the required tube
inspections and repair criteria within the tubesheet regions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This proposed change revises the San Onofre [Nuclear Generating
Station,] Units 2 and 3 Technical Specifications (TS) by revising
the definitions of steam generator ``Repair Limit'' and ``Tube
Inspection[,]'' as contained in TS items 5.5.2.11.f.1.f and
5.5.2.11.f.1.h, respectively. This proposed change also adds words
in the ``Operability determination'' requirement (item 5.5.2.11.f.2)
to provide consistency with the proposed change in the definition of
``Repair Limit.'' These revisions maintain existing design limits
and would not increase the probability or consequences of an
accident involving tube burst or primary to secondary accident-
induced leakage, as previously analyzed in the San Onofre [Nuclear
Generating Station,] Units 2 and 3 Updated Final Safety Analysis
Report (UFSAR). Also, the NEI 97-06 steam generator tube performance
criterion for structural integrity and accident-induced leakage will
continue to be satisfied.
Tube burst is precluded for a tube with defects within the
tubesheet region because of the constraint provided by the
tubesheet. As such, tube pullout resulting from the axial forces
induced by primary to secondary differential pressures would be a
prerequisite for tube burst to occur. An industry test program
(WCAP-16208-P Revision 1), and follow-on San Onofre site-specific
analysis (WCAP-16208-P Revision 1, Supplement 1) defined the non-
degraded hot leg tube to tubesheet joint length and cold leg tube to
tubesheet joint length required to preclude tube pullout and
maintain acceptable primary to secondary accident-induced leakage,
assuming that 100% [percent] of the steam generator tubes
experienced complete circumferential separation (360 degree through
wall crack) immediately below both the hot leg recommended
inspection length (C*) and the cold leg C*. Any degradation below C*
is shown by empirical test results and analyses to be acceptable,
thereby precluding an event with consequences similar to a
postulated tube rupture event.
WCAP-16208-P Revision 1, with Supplement 1 includes a total 0.2
gpm [gallons per minute]/steam generator assumed value for primary
to secondary accident-induced leakage. Inspection to the C* lengths
will ensure that the postulated accident-induced leakage will remain
below the current primary to secondary leakage assumption utilized
in the UFSAR accident analyses (Chapter 15).
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.
Steam generator tube leakage and structural integrity will be
maintained during all plant conditions upon implementation of the
proposed inspection scope and repair limit changes to the San Onofre
[Nuclear Generating Station,] Unit 2 and 3 Technical Specifications.
These changes do not introduce any new mechanisms that might result
in a different kind of accident from those previously evaluated.
Even with the limiting circumstances of complete circumferential
separation (360 degree through wall crack) of all of the tubes below
the C* length, [a] tube pullout is precluded and leakage is
predicted to be maintained within accident analysis assumptions.
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.
Operation with potential tube degradation below the C*
inspection length within the tubesheet region of the steam generator
tubing meets the intent of the inspection guidance of Regulatory
Guide Number 1.83, Revision 1, titled Inservice Inspection of
Pressurized Water Reactor Steam Generator Tubes, the requirements of
General Design Criteria 14, 15, 31 and 32 of 10 CFR 50, and the
recommendations of the Nuclear Energy Institute in NEI 97-06, titled
Steam Generator Program Guidelines.
The total leakage from an undetected flaw population below the
C* inspection length under postulated accident conditions is
accounted for to assure that it is within the bounds of the accident
analysis assumptions. Adequate margin remains for other possible
steam generator tube leak sources.
The proposed changes also maintain the structural and accident-
induced leakage integrity of the steam generator tubes as required
by NEI 97-06.
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.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: November 3, 2005.
Description of amendment request: The amendment would revise the
Technical Specifications (TS) to adopt NRC-approved Revision 4 to
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity.'' The proposed amendment includes changes to the TS
definition of Leakage, TS 3.4.13, ``RCS [Reactor Coolant System]
Operational Leakage,'' TS 5.5.9, ``Steam Generator (SG) Program,'' TS
5.6.9, ``Steam Generator Tube Inspection Report,'' and adds TS 3.4.17,
``Steam Generator (SG) Tube Integrity.'' The proposed changes are
necessary in order to implement the guidance for the industry
initiative on NEI 97-06, ``Steam Generator Program Guidelines.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the following
NSHC determination in its application dated November 3, 2005.
[[Page 72677]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change requires an SG Program that includes
performance criteria that will provide reasonable assurance that the
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.
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 an SGTR event, a bounding
primary to secondary LEAKAGE rate equal to the operational LEAKAGE
rate limits in the licensing basis plus the LEAKAGE rate associated
with a double-ended rupture of a single tube is assumed.
For other design-basis accidents such as a main steamline break
(MSLB), rod ejection, and reactor coolant pump locked rotor, the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs are 1 gallon per minute or
increases to 1 gallon per minute as a result of accident-induced
stresses. 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 to the TS 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 SG Program required by the proposed change to the TS. 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 proposed
changes do not, therefore, significantly increase the probability of
an accident previously evaluated.
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 [500 gallons per
day or 720 gallons per day] in any one SG, and that the reactor
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS
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 TSs 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 TSs.
Therefore, the proposed change does not affect the consequences
of an SGTR accident, and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance-based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG 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 SG 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.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
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
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical 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 SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff proposes to determine that the amendments 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
[[Page 72678]]
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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: July 9, 2004.
Brief description of amendments: The amendments revise the
Operating Licenses and Technical Specifications (TSs) to allow
operation of Palo Verde Nuclear Generating Station (PVNGS), Units 1 and
3 up to a maximum reactor core power level of 3990 Megawatts thermal
(MWt), an increase of 2.94 percent above the current licensed power
level of 3876 MWt. The proposed amendments would also make
administrative changes to the PVNGS Unit 2 TSs so that the changed
pages would apply to the three PVNGS units. Operation at the uprated
power level with replacement steam generators has been approved for
PVNGS Unit 2.
Date of issuance: November 16, 2005.
Effective date: November 16, 2005, and shall be implemented within
90 days of the date of issuance.
Amendment Nos.: Unit 1-157, Unit 2-157, Unit 3-157.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revise the Operating Licenses for Units 1 and 3 and the
Technical Specifications for all three units.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57980). The June 2, June 3 (two letters), June 17, July 9 (two
letters), July 19 (two letters), August 3, September 29, October 21,
and November 1, 2005, supplemental letters provided additional
clarifying information, 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 November 16, 2005.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 31, 2005, as supplemented by
letter dated September 13, 2005.
Brief description of amendment: The amendment permitted a one-time
change to Technical Specification Table 3.3.8.1-1 to provide a one-time
relaxation of the Loss of Power instrumentation requirements.
Date of issuance: September 15, 2005.
Effective date: As of the date of issuance to be implemented
immediately.
Amendment No.: 147.
Facility Operating License No. NPF-47: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. The NRC published a public notice of the proposed
amendment, issued a proposed finding of no significant hazards
consideration, and requested that any comments on the proposed no
significant hazards consideration be provided to the NRC staff by the
close of business on September 9, 2005. The notice was published in The
St. Francisville Democrat (in St. Francisville) on September 8, 2005,
and The Advocate (in Baton Rouge) on September 7, 2005. No public
comments were received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of Louisiana, and
final no significant hazards consideration determination are contained
in a Safety Evaluation dated September 15, 2005.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: November 1, 2004, as
supplemented by letters dated April 12, July 22, and September 26,
2005.
Brief description of amendment: The amendment authorizes the use of
a single-failure-proof gantry crane for spent fuel cask handling
operations up to 110 tons in weight.
Date of issuance: November 21, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 244.
Facility Operating License No. DPR-26: The amendment allows use of
the gantry crane for spent fuel cask handling operations up to 110 tons
in weight.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70716). The April 12, July 22, and September 26, 2005, supplements
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 21, 2005.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: November 3, 2004, and its
supplements dated February 24, June 23, and September 30, 2005.
Brief description of amendments: The amendments allow installation
and use of a temporary cask pit spent fuel storage rack for Units 1 and
2. The cask pit rack would allow the storage of an additional 154 spent
fuel assemblies for each unit. The total spent fuel pool storage
capacity for each unit would be increased from the current 1324 spent
fuel assemblies to 1478 assemblies for Cycles 14-16.
Date of issuance: November 21, 2005.
Effective date: As of the date of issuance, and shall be
implemented upon installation of the temporary cask pit spent fuel
rack.
Amendment Nos.: Unit 1--183; Unit 2-185.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 2004 (69
FR 76481). The February 24, June 23, and September 30, 2005,
supplemental letters provided additional clarifying information, did
not expand the scope of the application as originally noticed, and did
not change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a
[[Page 72679]]
Safety Evaluation dated November 21, 2005.
No significant hazards consideration comments received: Yes. The
comments are addressed in the enclosure of the above Safety Evaluation.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
[[Page 72680]]
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by email to
[email protected]. A copy of the request for hearing and 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 or 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).
PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric
Station, Unit 1 (SSES-1), Luzerne County, Pennsylvania
Date of amendment request: October 14, 2005, as supplemented on
October 21 and November 2, 2005.
Description of amendment request: The amendment changed the SSES-1
Technical Specifications (TSs) by revising the SSES-1 Cycle 14 Minimum
Critical Power Ratio Safety Limit in TS Section 2.1.1.2 from 1.08 to
1.09.
Date of issuance: November 10, 2005.
Effective date: November 10, 2005.
Amendment No.: 227.
Facility Operating License No. NPF-14: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. October 24, 2005 (70 FR 61475). The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by December 22, 2005, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment. The
Commission's related evaluation of the amendment, finding of exigent
circumstances, state consultation, and final NSHC determination are
contained in a safety evaluation dated November 10th 2005. The
supplemental letters dated October 21 and November 2, 2005, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the Nuclear Regulatory Commission staff's original proposed no
significant hazards consideration determination.
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. Laufer.
Virginia Electric and Power Company, Docket No. 50-338, North Anna
Power Station, Unit No. 1 (North Anna 1), Louisa County, Virginia
Date of amendment request: November 3, 2005, as supplemented by
letter dated November 4, 2005.
Description of amendment request: This amendment allows a temporary
7-day Completion Time to repair a weld leak that was discovered on the
low-head safety injection (LHSI) suction pump piping. This change is
needed to prevent an unnecessary plant transient and unscheduled
shutdown of North Anna 1.
Date of issuance: November 4, 2005.
Effective date: As of the date of issuance and is applicable until
the ``A'' train of the Unit 1 LHSI system is returned to operable
status or until November 9, 2005, at 0330 hours, whichever occurs
first.
[[Page 72681]]
Amendment No.: 246.
Renewed Facility Operating License No. NPF-4: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a Safety Evaluation dated
November 4, 2005.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: Evangelos Marinos.
Dated at Rockville, Maryland, this 28th day of November, 2005.
For the Nuclear Regulatory Commission.
Catherine Haney, Director,
Division of Operating Reactor Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 05-23553 Filed 12-5-05; 8:45 am]
BILLING CODE 7590-01-P