Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 61655-61669 [05-21180]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 205 (Tuesday, October 25, 2005)]
[Notices]
[Pages 61655-61669]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-21180]


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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 September 30 to October 13, 2005. The last 
biweekly notice was published on October 11, 2005 (70 FR 59082).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, 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.

[[Page 61656]]

    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, 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.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone 
Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: September 13, 2005.
    Description of amendment request: The amendment would reduce the 
temperature at which shutdown and control rod cluster control 
assemblies (RCCA) drop testing is done from greater than or equal to 
551 [deg]Fahrenheit (F) to greater than or equal to 500 [deg]F.
    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: Does the proposed amendment involve a significant 
increase in the probability or consequences of an accident previously 
evaluated?
    Response: No. DNC [Dominion Nuclear Connecticut, Inc.] is proposing 
to change the temperature at which the shutdown and control RCCA drop 
tests are performed from ``greater than or equal to 551 [deg]F,'' to 
``greater than or equal to 500 [deg]F.'' The proposed change does not 
modify any plant equipment and does not impact any failure modes that 
could lead to an accident. Additionally, the proposed change has no 
effect on the consequence of any analyzed accident since the change 
does not affect the function of any equipment credited for accident 
mitigation. Based on this discussion, the proposed amendment does not 
increase the probability or consequences of an accident previously 
evaluated.
    Criterion 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 does not modify any plant 
equipment and there is no impact on the capability of existing 
equipment to perform its intended functions. No system setpoints are 
being modified and no changes are being made to the method in which 
plant operations are conducted. No new failure modes are introduced by 
the proposed change. The proposed amendment does not introduce accident 
initiators or malfunctions that would cause a new or different kind of 
accident.
    As noted above, the proposed change does not affect the revisions 
to plant procedures, which were made to address Westinghouse Nuclear 
Safety Advisory Letter, NSAL-00-016 (Rod Withdrawal from Subcritical 
Protection in Lower Modes, issued in 2000).
    Therefore, the proposed amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?
    Response: No. The TS [technical specification] change does not 
involve a significant reduction in margin because the acceptance 
criterion for the RCCA drop time will not change. The proposed change 
will reduce the minimum RCCA drop test temperature from greater than or 
equal to 551 [deg]F to greater than or equal to 500 [deg]F. This will 
slightly increase the measured test

[[Page 61657]]

RCCA drop time. However, the measured test RCCA drop time is required 
to remain within the current TS limit of 2.7 seconds and the 2.19 
seconds for surveillance testing acceptance criteria (plant specific 
seismic allowance of 0.51 seconds). The proposed change does not affect 
any of the assumptions used in the accident analysis, nor does it 
affect any operability requirements for equipment important to plant 
safety. Therefore, the margin of safety is not impacted by the proposed 
amendment.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: Darrell J. Roberts.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone 
Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: September 13, 2005.
    Description of amendment request: The changes revise surveillance 
requirements for the recirculation spray system (RSS) to verify proper 
initiation of recirculation spray through actuation by the refueling 
water storage tank (RWST) low-low level signal instead of actuation by 
a timer.
    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: Does the proposed amendment involve a significant 
increase in the probability or consequences of an accident previously 
evaluated?
    Response: No. The RSS is only an accident mitigation system. As 
such, changes in the operation of RSS cannot have an impact on the 
probability of an accident. The delay in the start of the RSS pump is 
to assure there is sufficient water in the containment sump for 
adequate RSS pump NPSH [net positive suction head] and margin to 
suction pipe flashing in light of the debris analysis conducted in 
response to GL [Generic Letter] 2004-02. Containment analyses have been 
performed to demonstrate that there is no impact on the peak 
containment pressure and temperature following a LOCA [loss-of-coolant 
accident]. While there are some changes in the predicted post-LOCA 
environmental conditions, evaluations have been performed to show that 
there is no significant impact on the environmental qualification for 
equipment inside containment. The impact to piping and supports has 
been demonstrated to be acceptable without modification. Delay in RSS 
spray start will result in a reduction in diesel generator loading 
since the RSS pumps and the RHS pumps will no longer be running 
concurrently. The reduction in iodine removal efficiency during the 
delay period is more than offset by elimination of over-conservatisms 
in assumptions for long term iodine removal by the RSS system. The net 
impact is a reduction in the predicted offsite doses and control room 
doses following a design basis LOCA. Based on this discussion, the 
proposed amendment does not increase the probability or consequence of 
an accident previously evaluated.
    Criterion 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 modification alters the RSS pump 
circuitry by initiating the start sequence with an existing RWST low-
low level signal instead of a timer. The timer is now used to sequence 
pump starts. The pump function is not changed in any way. The proposed 
amendment does not introduce failure modes, accident initiators, or 
malfunctions that would cause a new or different kind of accident. 
Therefore, the proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?
    Response: No. The proposed change ensures that adequate margin to 
suction line flashing and NPSH margin exists for proper operation of 
the RSS pumps once the effects of debris are considered as required per 
GL 2004-02. Function of the pumps is not affected. Analyses have been 
performed that show the containment design basis limits are satisfied 
and the post-LOCA offsite and control room doses meet the required 
criteria. Therefore, based on the above, 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: Darrell J. Roberts.

Duke Energy Corporation, et al., Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina and 
Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, 
York County, South Carolina

    Date of amendment request: June 29, 2005.
    Description of amendment request: The amendment would revise 
Technical Specification Bases Section 3.6.11, ``Air Return System 
(ARS),'' and the Updated Final Safety Analysis Reports (UFSAR), Section 
6.2, ``Containment Systems,'' for McGuire Nuclear Station, Units 1 and 
2 and Catawba Nuclear Station, Units 1 and 2. The licensee proposes to 
implement an additional manual operator action to respond to NRC 
Bulletin 2003-01, ``Potential Impact of Debris Blockage on Emergency 
Sump Recirculation at Pressurized-Water Reactors.'' This amendment 
would allow plant operators to manually start one air return fan at a 
containment pressure of 1 psig prior to the automatic 9 minutes (+ 1 
minute) delayed start described in the UFSAR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

First Standard

    Does the change involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    No. The manual start of an Air Return System (ARS) fan will not 
result in a significant increase in the probability of an accident 
previously evaluated. The starting of an ARS fan is not considered to 
be an initiator of any accident or transient. This action is not taken 
during normal plant operation, but in response to an accident. The ARS 
fans do not operate to provide any normal ventilation requirement. The 
Containment Pressure Control System (CPCS) is provided to prevent 
excessive depressurization of the containment through inadvertent or 
excessive operation of certain engineered safety features. The CPCS 
prevents the

[[Page 61658]]

inadvertent actuation of an ARS fan during normal operation.
    This change is being requested in order to mitigate the 
consequences of a small break loss of coolant accident (SBLOCA) and 
help prevent or delay reaching the initiation pressure setpoint for 
containment spray, thereby reducing associated problems with possible 
sump debris buildup. SBLOCA events are bounded by the consequences of a 
design basis large break [loss of coolant accident] LOCA as addressed 
in Section 15 of the McGuire and Catawba [Updated Final Safety Analysis 
Report] UFSARs. Accordingly, this amendment will not involve a 
significant increase in the consequences of an accident previously 
evaluated.

Second Standard

    Does the change create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    No. The change proposed in this [license amendment request] LAR 
does not involve a physical alteration to the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing any normal plant operation. It does allow for the 
early start of one ARS fan during a SBLOCA event with containment 
pressure greater than 1 psig and less than 3 psig. This change will not 
affect or degrade the ability of the ARS to perform its specified 
safety functions.
    Accidents of a different type are credible accidents that the 
proposed amendment could create that are not bounded by UFSAR evaluated 
accidents. This amendment allows for the manual start of an ARS fan 
following a SBLOCA within the containment. No new failure modes are 
introduced due to the manual start of an ARS fan. The circuit used to 
manually start an ARS fan does not interfere with the automatic signal 
to start an ARS fan. This change does not require any modifications to 
the control circuitry for the ARS. The starting of an ARS fan is not 
considered to be an initiator of any accident or transient. This action 
(starting of an ARS fan) is not taken during normal operation, but in 
response to an accident. Previous accidents considered incredible are 
not made more likely by this change. A human performance error, such as 
starting the ARS fan too early, too late, or not at all, would not 
result in a substantial difference in the calculated differential 
pressure across the divider deck. Since no new malfunctions of 
equipment with a different result are introduced, all effects of any 
malfunctions are bounded by those already evaluated in the UFSAR. Thus 
it is concluded that the change contained in this LAR will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.

Third Standard

    Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The early manual start of an ARS fan for SBLOCA events will not 
reduce the ability of this system to perform its design functions to 
assure the rapid return of air from the upper to the lower containment 
compartment after the initial blowdown following a Design Basis 
Accident (DBA). The return of this air to the lower compartment and 
subsequent recirculation back up through the ice condenser assists in 
cooling the containment atmosphere and limiting post accident pressure 
and temperature in containment to less than design values. Limiting 
pressure and temperature also reduces the release of fission product 
radioactivity from containment to the environment in the event of a 
DBA. Therefore, there are no adverse dose effects from the early start 
of the ARS fan or from the delay of containment spray based on the 
current licensing basis.
    Analyses have shown that there will be no fan or damper malfunction 
due to the early manual start of a fan. The other functions of the 
system are not affected by the change proposed in this LAR. The manual 
start of the ARS during a SBLOCA will help maintain the margin of 
safety by forcing air and steam through the ice condenser with a 
subsequent reduction in the rate of pressure increase in the 
containment, and a delay in reaching the actuation setpoint for the 
containment spray system. The containment spray system will continue to 
be initiated at the normal setpoint pressure of the system (-3 psig). 
Therefore, the proposed changes listed above do not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Evangelos C. Marinos.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: August 17, 2005.
    Description of amendment request: The proposed changes will revise 
the Operating License Condition 2.C.(41), Fire Protection Program, to 
add a reference to the Nuclear Regulatory Commission (NRC) safety 
evaluation that allows the application of National Fire Protection 
Agency risk-informed, performance based fire protection methods and 
tools that have been approved by the NRC.
    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 activity involves the use of a risk-
informed, performance-based method to identify those circuits where a 
single fire could damage more than one safe shutdown train. These 
circuits would then be provided with one hour rated fire wrap. With the 
exception of the fire wrap itself, the proposed activity does not 
result in any physical changes to safety-related structures, systems, 
or components (SSCs), or the manner in which safety-related SSCs are 
operated, maintained, modified, tested, or inspected. The proposed 
activity does not degrade the performance or increase the challenges of 
any safety-related SSCs assumed to function in the accident analysis. 
As a result, the proposed activity does not introduce any new accident 
initiators. In addition, fires are not an accident that is previously 
evaluated in Chapter 15. Regardless, the proposed activity does not 
change the probability of a fire occurring since fire ignition 
frequency is independent of the presence of the fire wrap. The 
consequences of the proposed activity are bounded by the fire safe 
shutdown analysis, which assumes one train is free of fire damage.
    Therefore, providing one hour rated fire wrap for those circuits 
where a single fire could damage more than one safe shutdown train does 
not involve a significant increase in the probability or consequences 
of an accident previously evaluated.

[[Page 61659]]

    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 activity involves the use of a risk-
informed, performance-based method to identify those circuits where a 
single fire could damage more than one safe shutdown train. These 
circuits would then be provided with one hour rated fire wrap. With the 
exception of the fire wrap itself, the proposed activity does not 
result in any physical changes to safety-related structures, systems, 
or components (SSCs), or the manner in which safety-related SSCs are 
operated, maintained, modified, tested, or inspected. The proposed 
activity does not degrade the performance or increase the challenges of 
any safety-related SSCs assumed to function in the accident analysis. 
As a result, the proposed activity does not introduce nor increase the 
number of failure mechanisms of a new or different type than those 
previously evaluated. The fire safe shutdown analysis assumes one train 
is maintained free of fire damage.
    Therefore, providing one hour rated fire wrap for those circuits 
where a single fire could damage more than one safe shutdown train does 
not create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed activity involves the use of a risk-
informed, performance-based method to identify those circuits where a 
single fire could damage more than one safe shutdown train. These 
circuits would then be provided with one hour rated fire wrap. With the 
exception of the fire wrap itself, the proposed activity does not 
result in any physical changes to safety-related structures, systems, 
or components (SSCs), or the manner in which safety-related SSCs are 
operated, maintained, modified, tested, or inspected. The proposed 
activity does not degrade the performance or increase the challenges of 
any safety-related SSCs assumed to function in the accident analysis.
    The proposed activity does not impact plant safety since the 
conclusions of the fire safe shutdown analysis remain unchanged.
    Therefore, providing one hour rated fire wrap for those circuits 
where a single fire could damage more than one safe shutdown train 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 Section Chief: David Terao.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 21, 2005.
    Description of amendment request: The amendment proposes to replace 
the existing steam generator tube surveillance program with that being 
proposed by the Technical Specification Task Force (TSTF) in TSTF 449, 
Revision 4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No. The proposed change requires a Steam Generator 
Program that includes performance criteria that will provide reasonable 
assurance that the steam generator (SG) tubing will retain integrity 
over the full range of operating conditions (including startup, 
operation in the power range, hot standby, cooldown and all anticipated 
transients included in the design specification). The SG performance 
criteria are based on tube structural integrity, accident induced 
leakage, and operational leakage.
    The structural integrity performance criterion is:
    Structural integrity performance criterion: All in-service steam 
generator tubes shall retain structural integrity over the full range 
of normal operating conditions (including startup, operation in the 
power range, hot standby, and cool down and all anticipated transients 
included in the design specification) and design basis accidents. This 
includes retaining a safety factor of 3.0 against burst under normal 
steady state full power operation primary to secondary pressure 
differential and a safety factor of 1.4 against burst applied to the 
design basis accident primary to secondary pressure differentials. 
Apart from the above requirements, additional loading conditions 
associated with the design basis accidents, or combination of accidents 
in accordance with the design and licensing basis, shall also be 
evaluated to determine if the associated loads contribute significantly 
to burst or collapse. In the assessment of tube integrity, those loads 
that do significantly affect burst or collapse shall be determined and 
assessed in combination with the loads due to pressure with a safety 
factor of 1.2 on the combined primary loads and 1.0 on axial secondary 
loads.
    The accident induced leakage performance criterion is: The primary 
to secondary accident induced leakage rate for any design basis 
accidents, other than a SG tube rupture, shall not exceed the leakage 
rate assumed in the accident analysis in terms of total leakage rate 
for all SGs and leakage rate for an individual SG. Leakage is not to 
exceed 540 gallons per day through any one SG, except for specific 
types of degradation at specific locations as described in paragraph c 
of the Steam Generator Program.
    The operational leakage performance criterion is: The RCS 
operational primary to secondary leakage through any one SG shall be 
limited to <= 75 gallons per day per SG.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that is analyzed as part of a plant's licensing basis. 
In the analysis of a SGTR event, a bounding primary to secondary 
leakage rate equal to the leakage rate associated with a double-ended 
rupture of a single tube is assumed.
    For other design basis accidents such as main steam line break 
(MSLB), control element assembly (CEA) ejection, and reactor coolant 
pump seized rotor/sheared shaft, the tubes are assumed to retain their 
structural integrity (i.e., they are assumed not to rupture). The 
accident induced leakage criterion introduced by the proposed changes 
account for tubes that may leak during design basis accidents. The 
accident induced leakage criterion limits this leakage to no more than 
the value assumed in the accident analysis.
    The SG performance criteria proposed change identify the standards 
against which tube integrity is to be measured. Meeting the performance 
criteria provides reasonable assurance that the SG tubing will remain 
capable of fulfilling its specific safety function of maintaining 
reactor coolant pressure boundary integrity throughout each operating 
cycle and in the unlikely event of a design basis accident. The 
performance criteria are only a part of

[[Page 61660]]

the Steam Generator Program required by the proposed change. The 
program, defined by NEI [Nuclear Energy Institute] 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates a 
balance of prevention, inspection, evaluation, repair, and leakage 
monitoring.
    The consequences of design basis accidents are, in part, functions 
of the Specific Activity 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 Specific Activity in primary coolant to ensure the 
plant is operated within its analyzed condition. For those analyzed 
events that do not result in faulted steam generators, greater than or 
equal to 75 gpd [gallons per day] primary to secondary leakage per 
steam generator is assumed in the analysis. For those analyzed events 
that result in a faulted steam generator (e.g., MSLB), 540 gpd primary 
to secondary leakage is assumed though the faulted steam generator 
while greater than or equal to 75 gpd primary to secondary leakage is 
assumed though the intact steam generator.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current technical specifications and 
enhances the requirements for SG inspections. The proposed change does 
not adversely impact any other previously evaluated design basis 
accident and is an improvement over the current Technical 
Specifications.
    Therefore, the proposed change does not affect the consequences of 
a SGTR accident and the probability of such an accident is reduced. In 
addition, the proposed changes do not affect the consequences of other 
design basis events.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No. The proposed performance based requirements are an 
improvement over the requirements imposed by the current technical 
specifications.
    Implementation of the proposed Steam Generator Program will not 
introduce any adverse changes to the plant design basis or postulated 
accidents resulting from potential tube degradation. The result of the 
implementation of the Steam Generator Program will be an enhancement of 
SG tube performance. Primary to secondary leakage that may be 
experienced during all plant conditions will be monitored to ensure it 
remains within current accident analysis assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component.
    The change enhances SG inspection requirements. Therefore, the 
proposed change does not create the possibility of a new or different 
type of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The SG tubes in pressurized water reactors are an 
integral part of the reactor coolant pressure boundary and, as such, 
are relied upon to maintain the primary system's pressure and 
inventory. As part of the reactor coolant pressure boundary, the SG 
tubes are unique in that they are also relied upon as a heat transfer 
surface between the primary and secondary systems such that residual 
heat can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of a SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the Steam Generator Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the Steam Generator Program are consistent with those in 
the applicable design codes and standards and are an improvement over 
the requirements in the current technical specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Section Chief: David Terao.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: July 21, 2005.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 6.9.1.2 related to Occupational Radiation 
Exposure Reports and TS 6.9.1.5, ``Monthly Operating Reports.''
    Basis for proposed no significant hazards consideration 
determination: The NRC staff issued a notice of availability of a model 
no significant hazards consideration (NSHC) determination for 
referencing in license amendment applications in the Federal Register 
on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability 
of the model NSHC determination in its application dated July 21, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No. The proposed change eliminates the Technical 
Specifications (TSs) reporting requirements to provide a monthly 
operating letter report of shutdown experience and operating statistics 
if the equivalent data is submitted using an industry electronic 
database. It also eliminates the TS reporting requirement for an annual 
occupational radiation exposure report, which provides information 
beyond that specified in NRC regulations. The proposed change involves 
no changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of analyzed 
events or assumed mitigation of accidents or transients. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change does not involve a physical 
alteration of the plant, add any new equipment, or require any existing 
equipment to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.

[[Page 61661]]

    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 upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Michael L. Marshall, Jr.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: September 1, 2005.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 6.9.1.2 related to Occupational Radiation 
Exposure Reports and TS 6.9.1.6, ``Monthly Operating Reports.''
    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?
    No. The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating report of 
shutdown experience and operating statistics if the equivalent data is 
submitted using an industry electronic database. It also eliminates the 
TS reporting requirement for an annual occupational radiation exposure 
report, which provides information beyond that specified in NRC 
regulations. The proposed change involves no changes to plant systems 
or accident analyses. As such, the change is administrative in nature 
and does not affect initiators of analyzed events or assumed mitigation 
of accidents or transients. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    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?
    No. This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure data, 
and has no effect on plant equipment, operating practices or safety 
analyses assumptions. For these reasons, the proposed change does not 
involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Michael L. Marshall, Jr.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: July 21, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) testing frequency for the 
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' 
Specifically, the proposed change would revise the frequency for SR 
3.1.4.2, control rod scram time testing, from ``120 days cumulative 
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in licensing amendment applications in the Federal Register on August 
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the 
model NSHC determination in its application dated July 21, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No. The proposed change extends the frequency for testing 
control rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency of 
surveillance testing is not an initiator of any accident previously 
evaluated. The frequency of surveillance testing does not affect the 
ability to mitigate any accident previously evaluated, as the tested 
component is still required to be operable. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No. The proposed change extends the frequency for testing 
control rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. 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 extends the frequency for testing 
control rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected. Therefore, the 
proposed change does not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.

[[Page 61662]]

    NRC Section Chief: David Terao.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: November 12, 2004, as supplemented by 
letters dated September 2 and September 16, 2005.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications 3.1.7, ``Standby Liquid Control (SLC) 
System,'' for Hatch, Units 1 and 2. The proposed amendments would 
update Figures 3.1.7-1 and 3.1.7-2 for Units 1 and 2 TS to reflect the 
increased concentration of Boron-10 in the solution. Conforming 
revisions to Bases B 3.1.7, ``Standby Liquid Control (SLC) System'' are 
also included.
    The proposed amendment was previously noticed on February 1, 2005 
(70 FR 5249).
    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?
    This is a proposed change to Figures 3.1.7-1 and 3.1.7-2 of the 
Units 1 and 2 TS [Technical Specifications]. Figure 3.1.7-1 is a plot 
of the weight percent of Sodium Pentaborate solution in the Standby 
Liquid Control (SLC) Tank, as a function of the gross volume of 
solution in the tank. Figure 3.1.7-2 is a plot of the Sodium 
Pentaborate temperature versus concentration requirements.
    Figure 3.1.7-1 is proposed to be changed in order to accommodate an 
injection of Sodium Pentaborate solution into the reactor, following an 
ATWS [anticipated transient without scram] event, such that the 
concentration of Boron-10 atoms in the reactor will be 800 ppm natural 
Boron equivalent. This is necessary to accommodate increased cycle 
energy requirements for the Hatch Units 1 and 2 cores. Both Figures 
3.1.7-1 and 3.1.7-2 are changed to reflect that the boundary between 
Region A and B is changing from 6.9% to 7.0%. The proposed change to 
the Figures will not increase the probability of an ATWS event because 
the curves have nothing to do with the prevention of an ATWS event. The 
new requirements will insure that, in the future, the core will have 
adequate shutdown margin to mitigate the consequences of an ATWS event.
    The minimum concentration of Sodium Pentaborate which also 
represents the boundary between Region A and Region B, is changing from 
6.9% to 7.0%. This increase in the concentration ensures a conservative 
margin to the ATWS equivalency determination required by 10 CFR 50.62.
    Also, no systems or components designed to ensure the safe shutdown 
of the reactor are being physically changed as a result of this 
proposed TS change. In fact, no safety related systems or components 
designed for the prevention of previously evaluated events are being 
altered by the amendment.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    This proposed TS revision results in a change to SLC TS Figures 
3.1.7-1 and 3.1.7-2 requirements. However, these changes do not result 
in physical changes to the SLC system. SLC pump operation, maintenance 
and testing remain the same. Accordingly, no changes to the operation, 
maintenance or surveillance procedures will result from this TS 
revision request. Therefore, no new modes of operation are introduced 
by this TS change.
    Since no new modes of operation are introduced, the proposed change 
does not create the possibility of a new or different type event from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction in 
the margin of safety.
    This proposed TS change is being made to increase the boron 
concentration requirements of the sodium pentaborate solution injected 
into the reactor vessel following an Anticipated Transient Without 
Scram (ATWS) event. The change is necessary due to new fuel designs and 
higher energy requirements for fuel cycles. Therefore, the change is 
being made to insure that shutdown requirements can be met for the ATWS 
event. This will insure the margin of safety with respect to ATWS will 
continue to be met.
    The increase in the minimum concentration from 6.9% to 7.0% ensures 
a conservative margin with respect to the ATWS equivalency 
determination. Consequently, this proposed TS change will not result in 
a decrease 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Evangelos C. Marinos.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: September 1, 2005 (TS-05-04).
    Description of amendment request: The proposed amendment would 
revise the reactor protection system turbine trip allowable value for 
low trip system pressure from greater than or equal to 43 pounds per 
square inch gauge (psig) to 39.5 psig. This change would allow the 
instrumentation that performs this trip function to be tested and 
verified to be operable within the capabilities of the pressure 
switches.
    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 revises the allowable value for 
reactor trip as a result of a turbine trip on low trip system pressure. 
This change will not alter any plant components, systems, or processes 
and will only provide a more appropriate value to assess operability of 
the associated pressure switches. Since the plant features and 
operating practices are not altered, the possibility of an accident is 
not affected. This reactor trip is not directly credited in SQN's 
accident analysis and is maintained as an anticipatory trip to enhance 
the overall reliability of the reactor trip system. As such, there is 
not a specific safety limit associated with this function and the 
generation of a reactor trip based on low trip system pressure is above 
the required actuations to ensure acceptable mitigation of accidents. 
As the proposed change will continue to provide an acceptable 
anticipatory trip signal, the offsite dose potential is not affected by 
this change. Therefore, the proposed change does not involve a 
significant increase in the probability or

[[Page 61663]]

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. As described above, this change will not alter any 
plant equipment or operating practices that have the ability to create 
a new potential for accident generation. The proposed change revises 
the operability limits for a function that generates a trip signal when 
appropriate conditions exist to require accident mitigation response. 
This type of function does not have the ability to create an accident 
as its purpose and function is to mitigate events. 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 revise an allowable value 
for a reactor trip initiator that results from a turbine trip 
condition. This change will not alter the setpoint, and the calibration 
of the associated pressure switches will continue to be set at the 
current values. The allowable value change is in response to accuracy 
aspects of the instrumentation and does not alter the ability of this 
trip function to operate when and as needed to mitigate accident 
conditions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: August 26, 2005.
    Description of amendment request: The amendment would authorize 
changes to the Updated Safety Analysis Report (USAR) for Wolf Creek 
Generating Station (WCGS) that would revise the methodology for the 
reactor coolant system (RCS) leak detection instrumentation. This 
revision would clarify the requirements of the containment atmosphere 
gaseous radioactivity monitor with regard to the RCS leak detection 
capability and would justify that the monitor can be considered 
operable in compliance with Limiting Condition for Operation 3.4.15, in 
Technical Specification (TS) 3.4.15, ``RCS Leakage Detection 
Instrumentation,'' during all applicable Modes. There are no proposed 
changes to the WCGS TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change has been evaluated and determined to not 
increase the probability or consequences of an accident previously 
evaluated. The proposed change does not make hardware changes and does 
not alter the configuration of any plant system, structure, or 
component (SSC). The proposed change only clarifies the design and 
OPERABILITY requirements for the containment atmosphere gaseous 
radioactivity monitors and identifies the capabilities of the monitors 
at low RCS [radio]activity levels. The containment atmosphere gaseous 
radioactivity monitors are not initiators of any accident; therefore, 
the probability of occurrence of an accident is not increased. The USAR 
and TSs will continue to require diverse means of [RCS] leakage 
detection equipment, thus ensuring that leakage due to cracks [in the 
RCS] would continue to be identified prior to propagating to the point 
of a[n] [RCS] pipe break. Therefore, the consequences of an accident 
[previously evaluated] are not increased.
    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 involve the use or installation of new 
equipment and the currently installed equipment will not be operated in 
a new or different manner. No new or different system interactions are 
created and no new processes are introduced. The proposed changes will 
not introduce any new failure mechanisms, malfunctions, or accident 
initiators not already considered in the design and licensing basis 
[for WCGS]. The proposed change does not affect any SSC associated with 
an accident initiator. Based on this evaluation, the proposed change 
does not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change does not alter any RCS leakage detection 
components. The proposed change only clarifies the design and 
operability requirements for the containment atmosphere gaseous 
radioactivity monitor and identifies the capabilities of the 
containment atmosphere gaseous radioactivity monitors at low RCS 
[radio]activity levels. This change is required since the level of 
radioactivity in the WCGS reactor coolant has become much lower than 
what was assumed in the USAR and the gaseous channel [(monitor)] can no 
longer promptly detect a small RCS leak under all operating conditions. 
The proposed amendment continues to require diverse means of [RCS] 
leakage detection equipment with [the] capability to promptly detect 
RCS leakage. Although not required by [the] TS[s], additional diverse 
means of leakage detection capability are available as described in the 
USAR Section 5.2.5. Early detection of [RCS] leakage, as the potential 
indicator of a crack(s) in the RCS pressure boundary, will thus 
continue to be in place so that such a condition is known and 
appropriate actions [are] taken well before any such crack would 
propagate to a more severe condition. Based on this evaluation, 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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Daniel S. Collins, Acting.

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 applicati