[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, <a href="/cdn-cgi/l/email-protection#aee6cbcfdcc7c0c9eac1cdc5cbdaeec0dccd80c9c1d8"><span class="__cf_email__" data-cfemail="d69eb3b7a4bfb8b192b9b5bdb3a296b8a4b5f8b1b9a0">[email protected]</span></a>; 
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 
<a href="/cdn-cgi/l/email-protection#a7e8e0e4eac6cecbe4c2c9d3c2d5e7c9d5c489c0c8d1"><span class="__cf_email__" data-cfemail="f7b8b0b4ba969e9bb49299839285b7998594d9909881">[email protected]</span></a>. 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 <a href="/cdn-cgi/l/email-protection#770713053719051459101801"><span class="__cf_email__" data-cfemail="2f5f4b5d6f415d4c01484059">[email protected]</span></a>.

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 <plus-minus> 0.16 x 10<SUP>-</SUP><SUP>4</SUP> 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 <SUP>3</SUP> 525 
[deg]F to <SUP>3</SUP> 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 <plus-minus> 0.16 x 10<SUP>-4</SUP> 
[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 (ZrB<INF>2</INF>) 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 ZrB<INF>2</INF> 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 ZrB<INF>2</INF> 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 ZrB<INF>2</INF> 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 ZrB<INF>2</INF> 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) 
M5<SUP>TM</SUP> 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 M5<SUP>TM</SUP> 
material properties on NRC-approved methodology. M5<SUP>TM</SUP> 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 
M5<SUP>TM</SUP> cladding and structural material, has shown that the 
M5<SUP>TM</SUP> 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 M5<SUP>TM</SUP> 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 M5<SUP>TM</SUP> 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 M5<SUP>TM</SUP> 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 M5<SUP>TM</SUP> 
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 
<a href="/cdn-cgi/l/email-protection#aadaced8eac4d8c984cdc5dc"><span class="__cf_email__" data-cfemail="b7c7d3c5f7d9c5d499d0d8c1">[email protected]</span></a>.

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 <a href="/cdn-cgi/l/email-protection#a8d8ccdae8c6dacb86cfc7de"><span class="__cf_email__" data-cfemail="3c4c584e7c524e5f125b534a">[email protected]</span></a>.
    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 <a href="/cdn-cgi/l/email-protection#1f6f7b6d5f716d7c31787069"><span class="__cf_email__" data-cfemail="01716573416f73622f666e77">[email protected]</span></a>. 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.
---------------------------------------------------------------------------

    \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, <a href="/cdn-cgi/l/email-protection#8ac2efebf8e3e4edcee5e9e1effecae4f8e9a4ede5fc"><span class="__cf_email__" data-cfemail="0c44696d7e65626b48636f6769784c627e6f226b637a">[email protected]</span></a>; 
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 
<a href="/cdn-cgi/l/email-protection#f4bbb3b7b9959d98b7919a809186b49a8697da939b82"><span class="__cf_email__" data-cfemail="f0bfb7b3bd91999cb3959e849582b09e8293de979f86">[email protected]</span></a>. 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


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