Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 49568-49586 [E7-16766]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 166 (Tuesday, August 28, 2007)]
[Notices]
[Pages 49568-49586]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-16766]



[[Page 49567]]

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Part III





Nuclear Regulatory Commission





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Biweekly Notice; Applications and Amendments to Facility Operating 
Licenses Involving No Significant Hazards Considerations; Notice

Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / 
Notices

[[Page 49568]]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 2, 2007, to August 15, 2007. The last 
biweekly notice was published on August 14, 2007 (72 FR 45454).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing 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, https://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 basis 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

[[Page 49569]]

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, HearingDocket@nrc.gov; 
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 
OGCMailCenter@nrc.gov. 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, https://
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 pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, PA

    Date of amendment request: June 29, 2007.
    Description of amendment request: The proposed license amendment 
would revise the TMI-1 Technical Specifications 3.3.1.3, 3.3.2.1 and 
4.1, to reflect a change to the Reactor Building spray system buffering 
agent from sodium hydroxide to trisodium phosphate dodecahydrate. This 
proposed change is designed to minimize the potential for exacerbating 
sump screen blockage under post loss of coolant event conditions by 
limiting potential adverse chemical interactions between the buffering 
agent and certain insulation materials used in the TMI-1 containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    For the proposed change, trisodium phosphate dodecahydrate (TSP) 
will be used as a buffer for post-accident pH control and will 
replace the existing buffer. The buffer material and means of 
storage and delivery are not initiators for previously analyzed 
accidents. The accident mitigation function of the replacement 
buffer is the same as the existing buffer. The pH of the water in 
the emergency sump following a loss of coolant accident (LOCA) will 
be adjusted with TSP rather than sodium hydroxide (NaOH) to be 
within a range that will reduce the potential for elemental iodine 
re-evolution and long term stress corrosion during the recirculation 
mode of emergency core cooling system (ECCS) operation. In addition, 
the replacement buffer will reduce the formation of precipitates 
resulting from chemical reactions between the recirculating spray 
solution and insulating materials in the Reactor Building (RB), thus 
reducing the potential for ECCS emergency sump intake screen 
blockage. The proposed sump pH range will not result in an increase 
in post-LOCA hydrogen generation. The proposed isolation of the 
sodium hydroxide tank, and the installation of TSP in baskets has 
been evaluated for impacts on accident effects and the safety 
functions of required systems, structures, and components (SSCs). 
The RB emergency sump solution pH profile resulting from the 
proposed change has been evaluated for impacts on environmental 
qualification of SSCs. The accident mitigation functions of required 
SSCs will not be affected by the proposed change.
    As a part of the proposed change, the radiological consequences 
of a postulated LOCA have been reanalyzed using Standard Review Plan 
(SRP) 6.5.2, ``Containment Spray as a Fission Product Cleanup 
System,'' and the Alternate Source Term (AST) guidance in Regulatory 
Guide 1.183. The analysis considered the use of a plain borated 
water spray during the post-LOCA injection phase and a spray mixture 
with a minimum pH of 7.3 during the recirculation phase. The results 
of the reanalysis show that the consequences of the accident are not 
increased. The calculated doses at the Exclusion Area Boundary, Low 
Population Zone boundary, and in the Control Room remain within 10 
CFR 50.67 AST dose limits.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change will replace the existing spray additive 
design using sodium hydroxide solution stored in a tank with TSP 
contained in baskets located on the floor of the RB. The TSP storage 
and delivery method is passive. The baskets are constructed of 
stainless steel to resist corrosion and are seismically qualified. 
The existing sodium hydroxide tank, associated piping, and valves 
will no longer be used and will be permanently isolated, but their 
structural integrity will be maintained. The RB spray system will 
perform the same function and operate in the same manner for the 
proposed change; however, the sodium hydroxide tank isolation valves 
will no longer be required to open on an engineered safeguards 
actuation

[[Page 49570]]

signal. The accident mitigation function of TSP will be the same as 
the existing buffer, sodium hydroxide. The TSP will act as a 
buffering agent to raise the pH of the water in the containment 
emergency sump to greater than 7.3 for long-term post-LOCA RB spray 
recirculation. The SSCs required for post-LOCA accident mitigation 
have been evaluated for the proposed change including the effects of 
the modified emergency sump solution pH profile. No new accident 
scenarios, failure mechanisms, or single failures are introduced as 
a result of the proposed change.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change from sodium hydroxide to TSP will not reduce 
the effectiveness of the post-LOCA pH control buffer. The TSP will 
buffer the sump water sufficiently to assure that the resulting 
mixture pH is > 7.3 and < 8.0. This pH level will be effective in 
reducing the potential for iodine re-evolution during the 
recirculation phase of a LOCA, preventing long-term stress corrosion 
cracking of austenitic stainless steel, and minimizing post-LOCA 
hydrogen generation. In addition, the use of TSP will reduce the 
formation of precipitates resulting from chemical reactions between 
the recirculating spray solution and insulating materials in the RB, 
thus reducing the potential for ECCS emergency sump intake screen 
blockage. The proposed use of SRP 6.5.2 guidance, which is an NRC-
approved methodology, for post-LOCA dose calculations does not 
result in a reduction in a margin of safety. The proposed change 
does not adversely affect the performance of SSCs required for post-
LOCA mitigation, and does not affect an operating parameter or 
setpoint used in the accident analyses to establish a margin of 
safety. Also, the proposed change does not affect a margin of safety 
associated with containment functional performance.
    Therefore, the proposed change does not involve a significant 
reduction in any 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. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, SC

    Date of amendment request: July 17, 2007.
    Description of amendment request: A change is proposed to the 
standard technical specifications (STS) (NUREGs 1430 through 1434) and 
plant specific technical specifications (TS), to strengthen TS 
requirements regarding control room envelope (CRE) habitability by 
changing the action and surveillance requirements associated with the 
limiting condition for operation operability requirements for the CRE 
emergency ventilation system, and by adding a new TS administrative 
controls program on CRE habitability. Accompanying the proposed TS 
change are appropriate conforming technical changes to the TS Bases. 
The proposed revision to the Bases also includes editorial and 
administrative changes to reflect applicable changes to the 
corresponding STS Bases, which were made to improve clarity, conform 
with the latest information and references, correct factual errors, and 
achieve more consistency among the STS NUREGs. The proposed revision to 
the TS and associated Bases is consistent with STS as revised by TSTF-
448, Revision 3.
    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:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident From Any Accident Previously 
Evaluated

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
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 the Margin of Safety

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, WI

    Date of amendment request: June 12, 2007.

[[Page 49571]]

    Description of amendment request: The proposed amendment would 
revise the nuclear instrumentation system permissive setpoints in 
Technical Specification (TS) Table 3.5-2, ``Instrument Operation 
Conditions for Reactor Trip,'' revise the Table format, and revise TS 
2.3, ``Instrumentation System,'' to make consistent with other proposed 
changes to the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not change the probability or 
consequences of any previously evaluated accidents in the KPS 
[Kewaunee Power Station] updated safety analysis report (USAR). The 
proposed amendment would modify the TS setpoint values for the P-7 
and P-10 permissives. The actual plant settings will continue to be 
approximately 10% of rated reactor power. The reactor protection 
system (RPS) is designed to monitor various plant parameters and 
initiate a reactor trip in the event these parameters are outside 
predetermined limits. The RPS is not an accident initiator and 
therefore, changing the setpoints for these permissives will not 
increase the probability of an accident previously evaluated.
    The proposed amendment would add a setpoint band to the current 
TS required settings for permissive P-7 and P-10 to accommodate 
proper setting of the permissives. The only previously evaluated 
accident that is potentially affected by the proposed changes is the 
Uncontrolled Rod Cluster Assembly Rod Withdrawal At-Power (RWAP) 
accident analysis. The effects of these setpoint changes have been 
evaluated and determined not to have a significant effect on the 
consequences of the RWAP accident analysis results. The acceptance 
criteria for the RWAP accident analysis continue to be met. 
Therefore the proposed changes would not increase the consequences 
of an accident previously evaluated.
    Therefore the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment modifies the TS setpoint values for 
permissives P-7 and P-10. The actual plant settings will continue to 
be approximately 10% power. The proposed changes affect the power 
level at which RPS trip functions are enabled or blocked to ensure 
proper operation of the RPS. The changes do not add any new systems, 
structures or components (SSCs) or physically modify any existing 
SSCs with the possibility of creating a new accident.
    The proposed amendment does not functionally affect the 
operation of any SSC important to safety or its ability to perform 
its design function. Additionally, the proposed amendment does not 
create the possibility of a new or different kind of accident due to 
credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing bases.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment would add a setpoint band to the current 
TS required settings for permissivies P-7 and P-10 to accommodate 
proper setting of the permissives. The safety function of the 
nuclear instrumentation system and the affected permissives are not 
affected by this proposed change.
    The only safety analysis in the KPS USAR potentially affected by 
these proposed changes is the Uncontrolled Rod Cluster Assembly Rod 
Withdrawal At-Power (RWAP) event analysis. Evaluation of the RWAP 
event analysis results demonstrated that the RWAP would not have a 
significant effect on a margin of safety.
    The effects of the proposed change have been evaluated and all 
safety analysis acceptance criteria will continue to be met.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Acting Branch Chief: Travis L. Tate.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin

    Date of amendment request: July 2, 2007.
    Description of amendment request: The proposed amendment would 
delete operating license (OL) condition 2.C (5), ``Fuel Burnup,'' which 
restricts maximum rod average burnup to 60 giga-watt days per metric 
ton uranium (GWD/MTU). Deletion of the OL condition will provide the 
opportunity to increase maximum rod average burnup to as high as 62 
GWD/MTU and allow fuel management flexibility.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Deletion of KPS OL condition 2.C (5) does not add, delete, or 
modify any KPS systems, structures, or components (SSCs). The 
proposed amendment would effectively allow future increases in the 
KPS maximum rod average burnup limit using currently existing fuel 
management methods and models that have been reviewed and approved 
by the NRC [Nuclear Regulatory Commission].
    Maximum average rod burnup limits will continue to be maintained 
within safe and acceptable limits using these fuel management 
methods and models. Nuclear fuel is the only plant component 
potentially affected by increasing the maximum rod average burnup 
limit. Increasing the KPS maximum rod average burnup limit does not 
affect the thermal hydraulic response or the radiological 
consequences of any previously evaluated accident. The fuel rod 
design criteria will continue to be met at the maximum burnup limits 
allowed under the current fuel management and evaluation processes. 
An increase to the maximum rod average burnup limit will not 
increase the likelihood of a malfunction of nuclear fuel since the 
fuel currently used at KPS has been designed to support a maximum 
rod average burnup well in excess of 62 GWD/MTU.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment would delete a KPS OL condition that 
limits maximum rod average burnup. The proposed amendment would 
effectively allow future increases in the KPS maximum rod average 
burnup limit using currently existing fuel management methods and 
models that have been reviewed and approved by the NRC. Nuclear fuel 
is the only component potentially affected by changes to the maximum 
rod average burnup limit. The proposed amendment does not change the 
design function of the nuclear fuel or create any credible new 
failure mechanisms or malfunctions for nuclear fuel. Fuel rod design 
criteria will continue to be met at the maximum burnup limits 
allowed under the fuel management methods and models that have been 
previously reviewed and approved by the NRC. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?

[[Page 49572]]

    Response: No.
    The proposed amendment deletes a KPS OL condition that limits 
maximum rod average burnup. The proposed amendment would effectively 
allow future increases in the KPS maximum rod average burnup limit 
using currently existing methods and models that have been reviewed 
and approved by the NRC. The proposed amendment does not result in 
altering or exceeding a design basis or safety limit for the plant. 
All current fuel design criteria will continue to be satisfied, and 
the safety analysis of record, including evaluations of the 
radiological consequences of design basis accidents, will remain 
applicable.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Acting Branch Chief: Travis L. Tate.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, WA

    Date of amendment request: July 26, 2007, as superseded by letter 
dated August 8, 2007.
    Description of amendment request: The proposed changes revise the 
requirements of Technical Specification (TS) 3.3.5.2, ``Reactor Core 
Isolation Cooling (RCIC) System Instrumentation,'' and TS 3.5.2, ``ECCS 
[Emergency Core Cooling System]--Shutdown,'' to increase the Condensate 
Storage Tank (CST) level.
    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 operation of Columbia in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated. Neither of 
these changes affects the probability of any accident previously 
evaluated as they do not involve or impact accident initiators.
    The proposed change to TS 3.3.5.2 would ensure that the 
consequences would remain the same as that previously evaluated for 
during any event in which the RCIC pump was utilized. Adequate 
volume would be maintained in the CST whenever the RCIC pump was 
aligned to it to ensure that it did not experience loss of suction 
due to vortexing.
    The proposed changes to TS 3.5.2.2 would ensure that the 
previously assumed volume of water in the CST would still be 
available to inject into the reactor vessel during Modes 4 and 5 
should the suppression pool not meet minimum volume requirements. 
Therefore, operation of Columbia in accordance with the proposed 
amendment will 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 operation of Columbia in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
change will not create a new or different kind of accident since it 
only affects the amount of water held in reserve to support reactor 
vessel inventory loss. The proposed change does not introduce any 
credible mechanisms for unacceptable radiation release nor does it 
require physical modification to the plant. The plant has operated 
well within the existing allowable values. The increased margin 
provided by the increased level will assure no new or different 
kinds of accidents result from the proposed change. Therefore, the 
operation of Columbia in accordance with the proposed amendment will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The operation of Columbia in accordance with the proposed 
amendment will not involve a significant reduction in the margin of 
safety. The proposed amendment provides assurance that the RCIC pump 
suction will be transferred without loss of suction and that 135,000 
gallons of CST inventory will continue to be available for injection 
into the RPV [reactor pressure vessel] under worst case conditions. 
Therefore, operation of Columbia in accordance with the proposed 
amendment will 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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, WA

    Date of amendment request: July 30, 2007.
    Description of amendment request: The proposed changes revise 
Technical Specifications (TSs) 1.4, ``Frequency,'' 3.1.5, ``Control Rod 
Scram Accumulators,'' 3.4.1, ``Recirculation Loops Operating,'' 3.5.1, 
``ECCS [Emergency Core Cooling System]--Operating,'' 3.5.2, ``ECCS--
Shutdown,'' 3.7.1, ``Standby Service Water (SW) System and Ultimate 
Heat Sink (UHS),'' 3.8.1, ``AC [Alternating Current] Sources--
Operating,'' 3.8.2, ``AC Sources--Shutdown,'' and 5.5.6, ``Inservice 
Testing Program.'' The proposed changes include updates to adopt 
approved TS Task Force (TSTF) Travelers 284, Revision 3, ``Add `Met' 
vs. `Perform' to Specification 1.4, Frequency,'' TSTF-479, Revision 0, 
``Changes to Reflect Revision of 10 CFR 50.55a,'' and TSTF-485, 
Revision 0, ``Correct Example 1.4-1.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment is administrative in nature and does not 
affect analysis inputs or mitigation of analyzed accidents and 
transients. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    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. The 
proposed change does not introduce any new modes of plant operation 
or make any changes to system setpoints. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment is administrative in nature and does not 
involve physical changes to plant SSCs [structures, systems, or 
components], or the manner in which these SSCs are operated, 
maintained, modified, tested, or inspected. The proposed amendment 
does not involve a change to any

[[Page 49573]]

safety limit, limiting safety system setting, limiting condition for 
operation, or design parameters for any SSC. The only minor 
alteration to the plant design basis is relative to the application 
of TS 3.4.1. However, as discussed in Section 4 [of the licensee's 
submittal], this alteration biases the operation of the plant in the 
direction of safety. The proposed amendment does not impact any 
safety analysis assumptions and does not involve a change in initial 
conditions, system response times, or other parameters affecting any 
accident analysis. For these reasons, the proposed amendment does 
not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, WA

    Date of amendment request: July 30, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to establish more effective 
and appropriate action, surveillance, and administrative TS 
requirements related to ensuring the habitability of the control room 
envelope (CRE) in accordance with Nuclear Regulatory Commission (NRC)-
approved TS Task Force (TSTF) Standard Technical Specification change 
traveler TSTF-448, Revision 3, ``Control Room Habitability.'' 
Specifically, the proposed amendment would modify TS 3.7.3, ``Control 
Room Emergency Filtration (CREF) System,'' and add new TS 5.5.14, 
``Control Room Envelope Habitability Program,'' to Section 5.5, 
``Programs and Manuals.''
    The NRC staff issued a ``Notice of Availability of Technical 
Specification Improvement to Modify Requirements Regarding Control Room 
Envelope Habitability Using the Consolidated Line Item Improvement 
Process'' associated with TSTF-448, Revision 3, in the Federal Register 
on January 17, 2007 (72 FR 2022). The notice included a model safety 
evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request. In its 
application dated July 30, 2007, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee 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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE [control room 
envelope] emergency ventilation system, which is a mitigation system 
designed to minimize unfiltered air leakage into the CRE and to 
filter the CRE atmosphere to protect the CRE occupants in the event 
of accidents previously analyzed. An important part of the CRE 
emergency ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident From Any Accident Previously 
Evaluated

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation.
    The proposed change does not alter any safety analysis 
assumptions and is consistent with current plant operating practice. 
Therefore, this 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 the Margin of Safety

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based upon this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendment involves NSHC.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.

Energy Northwest, Docket No.50-397, Columbia Generating Station, Benton 
County, WA

    Date of amendment request: July 30, 2007.
    Description of amendment request: The proposed changes revise 
Technical Specifications (TSs) 3.3.3.1, ``Post Accident Monitoring 
(PAM) Instrumentation,'' 3.3.6.1, ``Primary Containment Isolation 
Instrumentation,'' 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs),'' and 3.6.4.2, ``Secondary Containment Isolation Valves 
(SCIVs).'' The proposed changes adopt the following TS Task Force 
(TSTF) Travelers that have been previously approved by the Nuclear 
Regulatory Commission (NRC): TSTF-45-A, Revision 2, ``Exempt 
Verification of CIVs [containment isolation valves] that are Not 
Locked, Sealed or Otherwise Secured,'' TSTF-46-A, Revision 1, ``Clarify 
the CIV Surveillance to Apply Only to Automatic Isolation Valves,'' 
TSTF-207-A, Revision 5, ``Completion Time for Restoration of Various 
Excessive Leakage Rates,'' TSTF-269-A, Revision 2, ``Allow 
Administrative Means of Position Verification for Locked or Sealed 
Valves,'' TSTF-295-A, Revision 0, ``Modify Note 2 to Actions of PAM 
Table to Allow Separate Condition Entry for Each Penetration,'' TSTF-
306-A, Revision 2, ``Add Action to LCO

[[Page 49574]]

[limiting condition for operation] 3.3.6.1 to Give Option to Isolate 
the Penetration,'' and TSTF-323-A, Revision 0, ``EFCV [excess flow 
check valve] Completion Time to 72 Hours.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below. The licensee addressed each 
proposed TSTF separately in its analysis:

TSTF-45-A, Revision 2

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change would exempt manual isolation valves and 
blind flanges located inside and outside the primary containment and 
in the secondary containment that are locked, sealed, or otherwise 
secured in position from the periodic verification of valve position 
required by SRs [surveillance requirements] 3.6.1.3.2 and 3.6.1.3.3, 
and SR 3.6.4.2.1. The exempted valves are verified to be in the 
correct position upon being locked, sealed, or secured. Because the 
valves are in the condition assumed in the accident analysis, the 
proposed change will not affect the initiators or mitigation of any 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change replaces the periodic verification of valve 
position with verification of valve position followed by locking, 
sealing, or otherwise securing the valve in position. Periodic 
verification is also effective in detecting valve mispositioning. 
However, verification followed by securing the valve in position is 
effective in preventing valve mispositioning.

TSTF-46-A, Revision 1

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change would revise the verification of PCIV and 
SCIV closure time to clarify that only power operated, automatic 
valves are required to be tested. PCIVs and SCIVs are not an 
initiator of any accident previously evaluated; rather, they serve 
to mitigate the consequences of evaluated accidents. The proposed 
change does not change the requirement to verify that power 
operated, automatic PCIVs and SCIVs close within the time assumed in 
the accident analysis, but rather, clarifies that non-automatic 
valves, which the accident analysis does not assume close within a 
specified time, are not required to be tested to verify the closure 
time. As a result, the mitigating action of the PCIVs and SCIVs is 
not affected by this change.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change would revise the verification of PCIV and 
SCIV closure time to clarify that only power operated, automatic 
valves are required to be tested, and not all power operated valves. 
There is no closure time assumed in the accident analysis for power 
operated PCIVs and SCIVs that are not automatic.

TSTF-207-A, Revision 5

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change revises the Actions of TS 3.6.1.3 to make 
the presentation consistent with similar Conditions in the ISTS 
[Improved Standard TSs]. Part of this change would extend the CT 
[completion time] for hydrostatically tested lines on a closed 
system to 72 hours for
    Condition D. Most of the proposed changes do not affect the 
requirements in the TS and have no effect on the initiation or 
mitigation of any accident previously evaluated. Leakage of 
hydrostatically tested lines on a closed system is not an initiator 
of any accident previously evaluated. The consequences of a 
previously evaluated accident during the extended CT are the same as 
the consequences during the existing CT.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes are editorial in nature and do not affect 
the requirements of the TS. Extension of the CT for hydrostatically 
tested lines on a closed system to 72 hours does not represent a 
significant reduction in safety given the reliability of closed 
systems. Nonetheless, leakage can be isolated restored by isolating 
the penetration with a valve not exceeding the leakage limits.

TSTF-269-A, Revision 2

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The proposed change modifies TS 3.6.1.3 and TS 
3.6.4.2. Both TS 3.6.1.3 and TS 3.6.4.2 require penetrations with an 
inoperable isolation valve to be isolated and periodically verified 
to be isolated. A Note is added to TS 3.6.1.3, Actions A and C, and 
TS 3.6.4.2, Action A, to allow isolation devices that are locked, 
sealed, or otherwise secured to be verified by use of administrative 
means. The proposed change does not affect any plant equipment, test 
methods, or plant operation, and are not initiators of any analyzed 
accident sequence. The inoperable containment penetrations will 
continue to be isolated, and hence perform their isolation function. 
Operation in accordance with the proposed TS will ensure that all 
analyzed accidents will continue to be mitigated as previously 
analyzed.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The PCIVs and SCIVs will continue to be operable or will 
be isolated as required by the existing specifications.

TSTF-295-A, Revision 0

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change clarifies the separate condition entry Note 
in TS 3.3.3.1 for Function 7, ``PCIV Position.'' The proposed change 
does not affect any plant equipment, test methods, or plant 
operation, and are not initiators of any analyzed accident sequence. 
The actions taken for inoperable PAM channels are not changed. 
Operation in accordance with the proposed TS will ensure that all 
analyzed accidents will continue to be mitigated as previously 
analyzed.

[[Page 49575]]

    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The PAM channels will continue to be operable or the 
existing, appropriate actions will be followed.

TSTF-306-A, Revision 2

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change revises TS 3.3.6.1 by adding an Actions Note 
that would allow penetration flow paths to be unisolated 
intermittently under administrative controls. Furthermore, the TIP 
[traversing incore probe] isolation system is segregated into a 
separate Function, allowing 24 hours to isolate the penetration. The 
proposed change does not affect any plant equipment, test methods, 
or plant operation, and are not initiators of any analyzed accident 
sequence. The allowance to unisolate a penetration flow path will 
not have a significant effect on the mitigation of any accident 
previously evaluated because the penetration flow path can be 
isolated, if needed, by a dedicated operator. The option to isolate 
a TIP penetration will ensure the penetration will perform as 
assumed in the accident analysis. Operation in accordance with the 
proposed TS will ensure that all analyzed accidents will continue to 
be mitigated as previously analyzed.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The allowance to unisolate a penetration flow path will 
not have a significant effect on a margin of safety because the 
penetration flow path can be isolated manually, if needed. The 
option to isolate a TIP penetration will ensure the penetration will 
perform as assumed in the accident analysis.

TSTF-323-A, Revision 0

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change would revise Action C of TS 3.6.1.3 to 
provide a 72-hour CT instead of a 12 hour CT to isolate an 
inoperable EFCV. PCIVs are not an initiator of any accident 
previously evaluated. The consequences of a previously evaluated 
accident during the extended CT are the same as the consequences 
during the existing CT.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The PCIVs serve to mitigate the potential for radioactive 
release from the primary containment following an accident. The 
design and response of the PCIVs to an accident are not affected by 
this change. The revised CT is appropriate given the EFCVs are on 
penetrations that have been found to have acceptable barrier(s) in 
the event that the single isolation valve failed.

    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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, MI

    Date of amendment request: May 22, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 5.5.7, ``Inservice Testing 
Program'' to: (1) Delete reference to American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code), Section 
XI and incorporate reference to the ASME Code for Operation and 
Maintenance of Nuclear Power Plants (ASME OM Code), and (2) address the 
applicability of Surveillance Requirement (SR) 3.0.2 to other normal 
and accelerated frequencies specified as two years or less in the 
inservice testing (IST) program.
    The proposed amendment incorporates changes based on U.S. Nuclear 
Regulatory Commission (NRC)--approved Technical Specification Task 
Force (TSTF) TSTF-479-A, ``Changes to Reflect Revision of 10 CFR 
50.55a,'' Revision 0, as modified by NRC-approved TSTF-497, ``Limit 
Inservice Testing Program SR 3.0.2 Application to Frequencies of Two 
Years or Less,'' Revision 0. The proposed changes include two 
deviations from the NRC-approved TSTFs that are administrative in 
nature: (1) Addition of ``ASME'' to TS 5.5.7 to make references to 
``ASME OM Code'' and (2) use of the term ``intervals'' instead of 
``frequencies.'' Basis for proposed no significant hazards 
consideration determination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed changes do not have any impact on the 
integrity of any plant system, structure, or component that 
initiates an analyzed event. The proposed changes would not alter 
the operation of, or otherwise increase the failure probability of 
any plant equipment that initiates an analyzed accident. Thus, the 
probability of any accident previously evaluated is not 
significantly increased.
    The prop