[Federal Register Volume 71, Number 137 (Tuesday, July 18, 2006)]
[Notices]
[Pages 40742-40759]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-6246]


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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

[[Page 40743]]

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 June 23, 2006 to July 6, 2006. The last 
biweekly notice was published on July 5, 2006 (71 FR 38180).

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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide

[[Page 40744]]

when the hearing is held. If the final determination is that the 
amendment request involves no significant hazards consideration, the 
Commission may issue the amendment and make it immediately effective, 
notwithstanding the request for a hearing. Any hearing held would take 
place after issuance of the amendment. If the final determination is 
that the amendment request involves a significant hazards 
consideration, any hearing held would take place before the issuance of 
any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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

    Date of amendment request: May 15, 2006.
    Description of amendment request: The amendment would revise the 
Technical Specification (TS) requirements related to steam generator 
tube integrity. The proposed changes are generally consistent with 
Revision 4 to Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler, TSTF-449, ``Steam Generator 
Tube Integrity.'' The availability of this TS improvement was announced 
in the Federal Register, on May 6, 2005 (70 FR 24126) as part of the 
consolidated line item improvement process (CLIIP). The proposed 
amendment includes changes to licensing pages to delete License 
Condition 2.c.(8), ``Repaired Steam Generators;'' changes to TS 3.1.6, 
``LEAKAGE;'' changes to TS Section 3.1.1.2, ``Steam Generators and 
Steam Generator (SG) Tube Integrity;'' revising TS Section 4.19, 
``Steam Generator (SG) Tube Integrity;'' adding new TS 6.9.6, ``Steam 
Generator Tube Inspection Report;'' and adding new TS 6.19, ``Steam 
Generator (SG) Program.''
    Basis for proposed no significant hazards consideration 
determination (NSHC): The NRC staff published a notice of opportunity 
for comment in the Federal Register on March 2, 2005 (70 FR 10298), on 
possible amendments adopting TSTF-449, including a model safety 
evaluation and model NSHC determination, using the CLIIP. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on May 6, 2005 (70 FR 24126). The licensee affirmed the applicability 
of the following NSHC determination in its application dated May 15, 
2006. As required by 10 CFR 50.91(a), an analysis of the issue of no 
significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE.
    A Steam Generator Tube Rupture (SGTR) event is one of the 
design-basis accidents that are analyzed as part of a plant's 
licensing basis. In the analysis of a SGTR event, a bounding primary 
to secondary LEAKAGE rate equal to the operational LEAKAGE rate 
limits in the licensing basis plus the LEAKAGE rate associated with 
a double-ended rupture of a single tube is assumed.
    For other design-basis accidents such as Main Steam Line Break 
(MSLB), rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident-induced 
stresses. The accident-induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design-
basis accidents. The accident-induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TSs 
identifies the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining reactor coolant pressure 
boundary integrity throughout each operating cycle and in the 
unlikely event of a design-basis accident. The performance criteria 
are only a part of the SG Program required by the proposed change to 
the TSs. 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 proposed changes do not, 
therefore, significantly increase the probability of an accident 
previously evaluated.
    The consequences of design-basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design-basis 
accident assumes that the primary-to-secondary leak rate after the 
accident is 1 gallon per minute with no more than 500 gallons per 
day in any one SG, and that the reactor coolant activity levels of 
DOSE EQUIVALENT I-131 are at the TS values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG

[[Page 40745]]

inspections. The proposed change does not adversely impact any other 
previously-evaluated design-basis accident and is an improvement 
over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed change does not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously-evaluated accident.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From Any Previously 
Evaluated.
    The proposed performance-based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design-basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an 
enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The SG tubes in pressurized-water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TSs.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief: Darrell J. Roberts.

Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: June 1, 2006.
    Description of amendment request: The proposed amendments would 
revise the Updated Final Safety Analysis Report (UFSAR) to incorporate 
the use of a fiber-reinforced polymer (FRP) system to strengthen 
existing masonry walls against tornado effects.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Response: Physical protection from a tornado event is a design 
basis criterion rather than a requirement of a previously analyzed 
UFSAR accident analysis.
    The current licensing basis (CLB) for Oconee states that 
systems, structures, and components (SSC's) required to shut down 
and maintain the units in a shutdown condition will not fail as a 
result of damage caused by natural phenomena.
    The in-fill masonry walls to be strengthened using an FRP system 
are passive, non-structural elements. The use of an FRP system on 
existing Auxiliary Building masonry walls will allow them to resist 
uniform pressure loads resulting from a tornado and will not 
adversely affect the structure's ability to withstand other design 
basis events such as earthquakes or fires. Therefore, the proposed 
use of FRP on existing masonry walls will not significantly increase 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Response: The final state of the FRP system is passive in nature 
and will not initiate or cause an accident. More generally, this 
understanding supports the conclusion that the potential for new or 
different kinds of accidents is not created.
    3. Involve a significant reduction in a margin of safety.
    Response: The application of an FRP system to existing auxiliary 
building masonry walls will either act to restore the margin of 
safety described in the UFSAR, e.g., the Unit 3 Control Room north 
wall, or enhance the margin of safety, e.g., the West Penetration 
Room walls, by increasing the walls' ability to resist tornado-
induced differential pressure and/or tornado wind. Consequently, 
this 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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: May 22, 2006.
    Description of amendment request: The proposed license amendment 
request would revise: (1) Surveillance Requirement (SR) 3.8.1.11 to 
remove the MODE restriction from Note 2 for Diesel Generator (DG)-3 
only, (2) SR 3.8.1.12 to remove the MODE restriction from Note 2 for 
DG-3 only, (3) SR 3.8.1.16 to remove the MODE restriction from the Note 
for DG-3 only, and (4) Revise SR 3.8.1.19 to remove the MODE 
restriction from Note 2 for DG-3 only.
    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 operation of Columbia Generating Station in 
accordance with the proposed amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The DG and its associated emergency loads are accident 
mitigating features, not accident initiating equipment. Therefore, 
there will be no impact on any accident probabilities by the 
approval of the requested amendment. The design of plant equipment 
is not being modified by these proposed changes. The capability of 
DG-1 and DG-2 to supply power to their safety related buses as 
designed will not be compromised by permitting performance of DG-3 
testing during power operations. Columbia's Technical Specifications 
require the RCIC [reactor core isolation cooling] system to be 
operable whenever this testing is performed at power. This ensures 
that the high-pressure injection function is maintained during the 
time the HPCS injection valve is disabled

[[Page 40746]]

during testing. In the event of a design basis accident during 
testing, the HPCS [high-pressure core spray] system could be 
returned to service well within the 14-day outage time allowed by 
Technical Specifications. Additionally, the ability of the Standby 
Liquid Coolant (SLC) system to perform its design safety function 
would not be affected because SLC is connected downstream of the 
HPCS injection valve. Therefore, there would be no significant 
impact on any accident consequences.
    Based on the above, the proposed change to permit certain DG 
surveillance tests to be performed during plant operation will have 
no effect on accident probabilities or consequences. Therefore, the 
proposed change does not involve a significant Increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the operation of Columbia Generating Station in 
accordance with the proposed amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new accident causal mechanisms would be introduced as a 
result of NRC approval of this amendment request since no changes 
are being made to the plant that would introduce any new accident 
causal mechanisms. Equipment will be operated in the same 
configuration with the exception of the plant mode in which the 
testing is conducted. This amendment request does not impact any 
plant systems that are accident initiators; neither does it 
adversely impact any accident mitigating systems.
    Based on the above, implementation of the proposed changes would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the operation of Columbia Generating Station in 
accordance with the proposed amendment involve a significant 
reduction in the margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes to the testing requirements for the DG 
do not affect the operability requirements for the DG, as 
verification of such operability will continue to be performed as 
required. Continued verification of operability supports the 
capability of the DG to perform its required function of providing 
emergency power to plant equipment that supports or constitutes the 
fission product barriers. Consequently, the performance of these 
fission product barriers will not be impacted by implementation of 
this proposed amendment. In addition, the proposed changes involve 
no changes to setpoints or limits established or assumed by the 
accident analysis. On this, and the above basis, no safety margins 
will be impacted.
    Energy Northwest concludes that there is no 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: David Terao.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of amendment request: April 24, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) consistent with the NRC-
approved Revision 4 to TS Task Force (TSTF) Standard TS Change 
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated April 24, 2006. Basis for 
proposed no significant hazards consideration determination: As 
required by 10 CFR 50.91(a), an analysis of the issue of no significant 
hazards consideration is presented below:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change requires a SG [Steam Generator] Program that 
includes performance criteria that will provide reasonable assurance 
that the SG tubing will retain integrity over the full range of 
operating conditions (including startup, operation in the power 
range, hot standby, cooldown and all anticipated transients included 
in the design specification). The SG performance criteria are based 
on tube structural integrity, accident induced leakage, and 
operational LEAKAGE.
    A[n] SGTR [steam generator tube rupture] event is one of the 
design basis accidents that are analyzed as part of a plant's 
licensing basis. In the analysis of a[n] SGTR event, a bounding 
primary to secondary LEAKAGE rate equal to the operational LEAKAGE 
rate limits in the licensing basis plus the LEAKAGE rate associated 
with a double-ended rupture of a single tube is assumed.
    For other design basis accidents such as MSLB [main steamline 
break], rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change[s] to the TS[s] 
identify the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining reactor coolant pressure 
boundary integrity throughout each operating cycle and in the 
unlikely event of a design basis accident. The performance criteria 
are only a part of the SG Program required by the proposed change to 
the TS[s]. 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 proposed changes do not, 
therefore, significantly increase the probability of an accident 
previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.
    Criterion 2--The proposed change does not create the possibility 
of a new or different

[[Page 40747]]

kind of accident from any previously evaluated.
    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an 
enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in the margin of safety.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes. Steam generator 
tube integrity is a function of the design, environment, and the 
physical condition of the tube. The proposed change does not affect 
tube design or operating environment. The proposed change is 
expected to result in an improvement in the tube integrity by 
implementing the SG Program to manage SG tube inspection, 
assessment, repair, and plugging. The requirements established by 
the SG Program are consistent with those in the applicable design 
codes and standards and are an improvement over the requirements in 
the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    Based upon the reasoning presented above 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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Michael L. Marshall, Jr.

Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, 
Unit No. 2, St. Lucie County, Florida

    Date of amendment request: May 25, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) consistent with the NRC-
approved Revision 4 to TS Task Force (TSTF) Standard TS Change 
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated May 25, 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change requires a SG [Steam Generator] Program that 
includes performance criteria that will provide reasonable assurance 
that the SG tubing will retain integrity over the full range of 
operating conditions (including startup, operation in the power 
range, hot standby, cooldown and all anticipated transients included 
in the design specification). The SG performance criteria are based 
on tube structural integrity, accident induced leakage, and 
operational LEAKAGE.
    A[n] SGTR [steam generator tube rupture] event is one of the 
design basis accidents that are analyzed as part of a plant's 
licensing basis. In the analysis of a[n] SGTR event, a bounding 
primary to secondary LEAKAGE rate equal to the operational LEAKAGE 
rate limits in the licensing basis plus the LEAKAGE rate associated 
with a double-ended rupture of a single tube is assumed.
    For other design basis accidents such as MSLB [main steamline 
break], rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change[s] to the TS[s] 
identify the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining reactor coolant pressure 
boundary integrity throughout each operating cycle and in the 
unlikely event of a design basis accident. The performance criteria 
are only a part of the SG Program required by the proposed change to 
the TS[s]. 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 proposed changes do not, 
therefore, significantly increase the probability of an accident 
previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an

[[Page 40748]]

enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

    Criterion 3--The proposed change does not involve a significant 
reduction in the margin of safety.

    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes. Steam generator 
tube integrity is a function of the design, environment, and the 
physical condition of the tube. The proposed change does not affect 
tube design or operating environment. The proposed change is 
expected to result in an improvement in the tube integrity by 
implementing the SG Program to manage SG tube inspection, 
assessment, repair, and plugging. The requirements established by 
the SG Program are consistent with those in the applicable design 
codes and standards and are an improvement over the requirements in 
the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    Based upon the reasoning presented above 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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Michael L. Marshall, Jr.

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: April 27, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) consistent with the NRC-
approved Revision 4 to TS Task Force (TSTF) Standard TS Change 
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated April 27, 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change requires a SG [Steam Generator] Program that 
includes performance criteria that will provide reasonable assurance 
that the SG tubing will retain integrity over the full range of 
operating conditions (including startup, operation in the power 
range, hot standby, cooldown and all anticipated transients included 
in the design specification). The SG performance criteria are based 
on tube structural integrity, accident induced leakage, and 
operational LEAKAGE.
    A[n] SGTR [steam generator tube rupture] event is one of the 
design basis accidents that are analyzed as part of a plant's 
licensing basis. In the analysis of a[n] SGTR event, a bounding 
primary to secondary LEAKAGE rate equal to the operational LEAKAGE 
rate limits in the licensing basis plus the LEAKAGE rate associated 
with a double-ended rupture of a single tube is assumed.
    For other design basis accidents such as MSLB [main steamline 
break], rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change[s] to the TS[s] 
identify the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining reactor coolant pressure 
boundary integrity throughout each operating cycle and in the 
unlikely event of a design basis accident. The performance criteria 
are only a part of the SG Program required by the proposed change to 
the TS[s]. 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 proposed changes do not, 
therefore, significantly increase the probability of an accident 
previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an 
enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.

[[Page 40749]]

    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in the margin of safety.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes. Steam generator 
tube integrity is a function of the design, environment, and the 
physical condition of the tube. The proposed change does not affect 
tube design or operating environment. The proposed change is 
expected to result in an improvement in the tube integrity by 
implementing the SG Program to manage SG tube inspection, 
assessment, repair, and plugging. The requirements established by 
the SG Program are consistent with those in the applicable design 
codes and standards and are an improvement over the requirements in 
the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    Based upon the reasoning presented above 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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Michael L. Marshall, Jr.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: January 18, 2006.
    Description of amendment request: The proposed amendment would 
delete the reference to the hydrogen monitors in Technical 
Specification (TS) 3.6.11, ``Accident Monitoring Instrumentation'' 
consistent with the NRC-approved Industry/Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
447, ``Elimination of Hydrogen Recombiners and Change to Hydrogen and 
Oxygen Monitors.''
    The NRC staff issued a notice of availability of ``Model 
Application Concerning Technical Specification Improvement To Eliminate 
Hydrogen Recombiner Requirement, and Relax the Hydrogen and Oxygen 
Monitor Requirements for Light Water Reactors Using the Consolidated 
Line Item Improvement Process (CLIIP)'', in the Federal Register on 
September 25, 2003 (68 FR 55416). The notice included a model safety 
evaluation (SE), a model no significant hazards consideration (NSHC) 
determination, and a model application.
    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, by confirming the applicability of the model NSHC 
determination to NMP-1 and incorporating it by reference in its 
application. The model NSHC determination is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen [and oxygen] monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key 
variables that most directly indicate the accomplishment of a safety 
function for design-basis accident events. The hydrogen [and oxygen] 
monitors no longer meet the definition of Category 1 in RG 1.97. As 
part of the rulemaking to revise 10 CFR 50.44 the Commission found 
that Category 3, as defined in RG 1.97, is an appropriate 
categorization for the hydrogen monitors because the monitors are 
required to diagnose the course of beyond design-basis accidents. 
[Also, as part of the rulemaking to revise 10 CFR 50.44, the 
Commission found that Category 2, as defined in RG 1.97, is an 
appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert 
containment.]
    The regulatory requirements for the hydrogen [and oxygen] 
monitors can be relaxed without degrading the plant emergency 
response. The emergency response, in this sense, refers to the 
methodologies used in ascertaining the condition of the reactor 
core, mitigating the consequences of an accident, assessing and 
projecting offsite releases of radioactivity, and establishing 
protective action recommendations to be communicated to offsite 
authorities. Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2] and removal of 
the hydrogen [and oxygen] monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs [severe 
accident management guidelines], the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen [and oxygen] monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From Any Previously 
Evaluated.
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen [and oxygen] monitor equipment was intended to mitigate 
a design-basis hydrogen release. The hydrogen recombiner and 
hydrogen [and oxygen] monitor equipment are not considered accident 
precursors, nor does their existence or elimination have any adverse 
impact on the pre-accident state of the reactor core or post 
accident confinement of radionuclides within the containment 
building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in [a] Margin of Safety.
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that

[[Page 40750]]

this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI [Three Mile Island], 
Unit 2 accident can be adequately met without reliance on safety-
related hydrogen monitors.
    [Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.]
    Therefore, this change does not involve a significant reduction 
in [a] margin of safety. [The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors.] Removal of 
hydrogen [and oxygen] monitoring from TS will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    The NRC staff has reviewed the model NSHC determination and its 
applicability to NMP-1. Based on this review, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Richard J. Laufer.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: June 6, 2006.
    Description of amendment request: The proposed amendments would 
revise the design basis as described in the Point Beach Nuclear Plant 
Final Safety Analysis Report (FSAR) by incorporating an updated 
analysis for satisfying the reactor vessel Charpy upper-shelf energy 
requirements of 10 CFR part 50, Appendix G, Section IV.A.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. Would the proposed amendment involve a significant increase 
in the probability or consequences of any accident previously 
evaluated?
    The proposed change incorporates the updated analysis for 
satisfying the reactor vessel Charpy upper-shelf energy requirements 
of 10 CFR part 50, Appendix G, Section IV.A.1 into the FSAR. The 
proposed change does not adversely affect accident initiators or 
precursors nor alter the design assumptions, conditions, or the 
manner in which the plant is operated and maintained. The proposed 
change does not alter or prevent the ability of structures, systems, 
and components from performing their intended function to mitigate 
the consequences of an initiating event within the assumed 
acceptance limits. The proposed change does not affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the types or amounts of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposures. The proposed change is 
consistent with safety analysis assumptions and resultant 
consequences. Therefore, it is concluded that this change does not 
significantly increase the probability of occurrence of an accident 
previously evaluated.
    2. Would the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed change incorporates the updated analysis for 
satisfying the reactor vessel Charpy upper-shelf energy requirements 
of 10 CFR part 50, Appendix G, Section IV.A.1 into the FSAR. The 
change does not impose any new or different requirements or 
eliminate any existing requirements. The change does not alter 
assumptions made in the safety analysis. The proposed change is 
consistent with the safety analysis assumptions and current plant 
operating practice. Therefore, the proposed change would not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Would the proposed amendment result in a significant 
reduction in a margin of safety?
    The proposed change incorporates the updated analysis for 
satisfying the reactor vessel Charpy upper-shelf energy requirements 
of 10 CFR part 50, Appendix G, Section IV.A.1 into the FSAR. The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings or limiting conditions for operation 
are determined. The setpoints at which protective actions are 
initiated are not altered by the proposed change. Therefore, the 
proposed amendment does not result in a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: May 30, 2006.
    Description of amendment request: The proposed amendment would 
revise the Fort Calhoun Station, Unit 1 (FCS) Technical Specification 
(TS) requirements related to steam generator tube integrity. The change 
is consistent with NRC-approved Revision 4 to Technical Specification 
Task Force (TSTF) Standard Technical Specification Change Traveler 
TSTF-449, ``Steam Generator Tube Integrity.'' The availability of this 
TS improvement was announced in the Federal Register on May 6, 2005 (70 
FR 24126) as part of the consolidated line item improvement process 
(CLIIP).
    Omaha Public Power District (OPPD) also proposes to change the FCS 
TS by deleting the sleeving repair alternative to plugging for steam 
generator tubes. The FCS replacement steam generators (RSGs) to be 
installed during the fall of 2006 are manufactured by Mitsubishi Heavy 
Industries, Ltd. (MHI). The change is being requested because OPPD has 
determined that the sleeving repair alternative to plugging will not be 
used for the MHI RSGs.
    Basis for proposed no significant hazards consideration 
determination: OPPD stated that it had reviewed the proposed no 
significant hazards consideration determination published on March 2, 
2005 (70 FR 10298), as part of the CLIIP. OPPD has concluded that the 
proposed determination presented in the notice is applicable to FCS and 
the determination is incorporated by reference to satisfy the 
requirements of 10 CFR 50.91(a). 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 elimination from the TS surveillance requirements of leak 
tight sleeves as a repair method alternative to plugging defective 
steam generator tubes does not introduce an initiator to any 
previously evaluated accident. The frequency or periodicity of 
performance of the remaining surveillance requirements for steam 
generator tubes (including plugged tubes) is not affected by this 
change. Elimination of the tube repair method has no effect on the 
consequences of any previously evaluated accident. The

[[Page 40751]]

proposed changes will not prevent safety systems from performing 
their accident mitigation function as assumed in the safety 
analysis.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any 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 only affects the TS surveillance 
requirements. The proposed change is a result of installation of 
RSGs. The proposed change will eliminate a steam generator tube 
repair alternative which cannot be utilized or credited for the 
RSGs. This change will not alter assumptions made in the safety 
analysis and licensing bases and will not create new or different 
systems interactions.
    Therefore, this 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 deletes surveillance requirements for a 
steam generator tube repair alternative which will no longer be 
necessary or applicable. The remaining TS steam generator tube 
surveillance requirements, including inspection and plugging 
requirements, will continue to maintain the applicable margin of 
safety.
    Therefore, this TS change does not involve a significant 
reduction in the margin of safety.

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

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment requests: May 30, 2006.
    Description of amendment requests: The proposed amendment would 
revise the Technical Specifications (TSs) to adopt NRC-approved 
Revision 4 to Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler, TSTF-449, ``Steam Generator 
Tube Integrity.'' The proposed amendment includes changes to the TS 
definition of Leakage, TS 3.4.13, ``RCS [Reactor Coolant System] 
Operational Leakage,'' TS 5.5.9, ``Steam Generator (SG) Tube 
Surveillance Program,'' TS 5.6.10, ``Steam Generator (SG) Tube 
Inspection Report,'' and adds TS 3.4.17, ``Steam Generator (SG) Tube 
Integrity.'' The proposed changes are necessary in order to implement 
the guidance for the industry initiative on NEI 97-06, ``Steam 
Generator Program Guidelines.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated May 30, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change requires an SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). The SG performance criteria are based on tube 
structural integrity, accident-induced leakage, and operational 
LEAKAGE.
    A steam generator tube rupture (SGTR) event is one of the 
design-basis accidents that are analyzed as part of a plant's 
licensing basis. In the analysis of an SGTR event, a bounding 
primary to secondary LEAKAGE rate equal to the operational LEAKAGE 
rate limits in the licensing basis plus the LEAKAGE rate associated 
with a double-ended rupture of a single tube is assumed.
    For other design-basis accidents such as a main steamline break 
(MSLB), rod ejection, and reactor coolant pump locked rotor, the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs are 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident-induced 
stresses. The accident-induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design-
basis accidents. The accident-induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design-basis accident. The performance criteria are only a part of 
the SG Program required by the proposed change to the TS. The 
program, defined by NEI 97-06, ``Steam Generator Program 
Guidelines,'' includes a framework that incorporates a balance of 
prevention, inspection, evaluation, repair, and leakage monitoring. 
The proposed changes do not, therefore, significantly increase the 
probability of an accident previously evaluated.
    The consequences of design-basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design-basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design-basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of an SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From Any Previously 
Evaluated.
    The proposed performance-based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an 
enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation,

[[Page 40752]]

or primary or secondary coolant chemistry controls. In addition, the 
proposed change does not impact any other plant system or component. 
The change enhances SG inspection requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The SG tubes in pressurized-water reactors are an integral part 
of the the primary system's pressure and inventory. As part of the 
reactor coolant pressure boundary, the SG tubes are unique in that 
they are also relied upon as a heat transfer surface between the 
primary and secondary systems such that residual heat can be removed 
from the primary system. In addition, the SG tubes isolate the 
radioactive fission products in the primary coolant from the 
secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    The NRC staff proposes to determine that the amendment requests 
involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

 PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: May 1, 2006.
    Description of amendment request: The proposed amendment would 
eliminate the requirement for a power range, neutron flux, high 
negative rate trip and delete the references to this trip as functional 
Unit 4 in Salem Generating Station (Salem) Unit Nos. 1 and 2 Technical 
Specification (TS) Table 2.2-1, ``Reactor Trip System Instrumentation 
Trip Setpoints,'' TS Table 3.3-1, ``Reactor Trip System 
Instrumentation,'' TS Table 3.3-2, ``Reactor Trip System 
Instrumentation Response Times,'' and TS Table 4.3-1, ``Reactor Trip 
System Instrumentation Surveillance Requirements [SRs].'' The proposed 
changes are consistent with the methodology presented in the 
Westinghouse Topical Report WCAP-11394-P-A, ``Methodology for the 
Analysis of the Dropped Rod Event,'' which has been reviewed by the NRC 
and found acceptable for referencing in license applications. The 
amendment also would involve the correction of errata in the TS for 
Salem Unit Nos. 1 and 2.
    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 elimination of the Power Range, Neutron Flux, Negative Rate 
trip does not increase the probability or consequences of reactor 
core damage accidents resulting from Rod Cluster Control Assembly 
(RCCA) Misalignment events previously analyzed. The safety functions 
of other safety-related systems and components have not been 
altered. All other Reactor Trip System protection functions are not 
impacted by the elimination of the requirement for a Power Range, 
Neutron Flux, High Negative Rate trip. The Power Range, Neutron 
Flux, High Negative Rate trip circuitry detects and responds to 
negative reactivity insertion due to RCCA misoperation events, 
should they occur. Therefore, the Power Range, Neutron Flux, High 
Negative Rate trip is not assumed in the initiation of such events. 
The consequences of accidents previously evaluated in the Salem 
Generating Station (Salem) Updated Final Safety Analysis Report 
(UFSAR) are unaffected by the proposed changes because no change to 
any equipment response or accident mitigation scenario has resulted. 
The proposed changes do not modify the RCCAs or change the 
acceptance criteria for departure from nucleate boiling (DNB). The 
TS change reflects analysis described in the UFSAR and cycle-
specific analysis performed each fuel cycle.
    The proposed revisions to Salem Unit 1 Index page XII, Salem 
Unit 1 TS 4.2.2.2, Salem Unit 2 TS 4.2.2.2, Salem Unit 1 TS Table 
3.3-2, Salem Unit 2 SR number for boron concentration on page 3/4 9-
1, Salem Unit 1 TS 6.9.1.5.a, and Salem Unit 1 TS 6.9.1.5.b contain 
changes administrative in nature that correct errors and do not 
affect the intent of any TS requirements.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or radiological 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 elimination of the Power Range, Neutron Flux, High Negative 
Rate trip does not create the possibility of a new or different kind 
of accident from any accident previously evaluated in the UFSAR. No 
new accident scenarios, failure mechanisms, or limiting single 
failures are introduced as a result of the proposed changes. The 
proposed changes do not challenge the performance or integrity of 
the RCCAs or any other safety-related system. The proposed changes 
will have no adverse effect on the availability, operability, or 
performance of the safety-related systems and components assumed to 
actuate in the event of a design basis accident (DBA) or transient. 
It has been demonstrated that the Power Range, Neutron Flux, High 
Negative Rate trip can be eliminated by the NRC approved methodology 
described in WCAP-11394-P. The Salem fuel cycle specific analyses 
have confirmed that for a dropped RCCA event, no direct reactor trip 
or automatic power reduction is required to meet the DNB limits for 
this Condition II, ``Fault of Moderate Frequency,'' event. The Power 
Range, Neutron Flux, High Negative Rate trip is not credited either 
as a primary or backup mitigation feature for any other UFSAR event.
    The proposed revisions to Salem Unit 1 Index page XII, Salem 
Unit 1 TS 4.2.2.2, Salem Unit 2 TS 4.2.2.2, Salem Unit 1 TS Table 
3.3-2, Salem Unit 2 SR number for boron concentration on page 3/4 9-
1, Salem Unit 1 TS 6.9.1.5.a, and Salem Unit 1 TS 6.9.1.5.b contain 
changes administrative in nature that correct errors and do not 
affect the intent of any TS requirements.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is the difference between the DNB 
acceptance limit and the failure of the fuel rod cladding. The Salem 
fuel cycle specific analyses have confirmed that for a dropped RCCA 
event, DNB limits are not exceeded with the proposed changes. 
Conformance to the licensing basis acceptance criteria for DBAs and 
transients with the elimination of the Power Range, Neutron Flux, 
High Negative Rate trip is demonstrated and the DNB limits are not 
exceeded when the NRC approved methodology of WCAP-11394-P is 
applied. The margin of safety associated with the licensing basis 
acceptance criteria for any postulated accident is unchanged.
    The proposed revisions to Salem Unit 1 Index page XII, Salem 
Unit 1 TS 4.2.2.2, Salem Unit 2 TS 4.2.2.2, Salem Unit 1 TS Table 
3.3-2, Salem Unit 2 SR number for boron concentration on page 3/4 9-
1, Salem Unit 1 TS 6.9.1.5.a, and Salem Unit 1 TS 6.9.1.5.b contain 
changes administrative in nature that correct errors and do not 
affect the intent of any TS requirements.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 40753]]

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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Darrell J. Roberts.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: May 1, 2006.
    Description of amendment request: The amendment would move the main 
steamline discharge (safety valves and atmospheric dumps) radiation 
monitors (R46) from the radiation monitoring instrumentation Technical 
Specification (TS) 3.3.3.1, to the accident monitoring TS 3.3.3.7. The 
purpose of the R46 monitors is to provide continuous monitoring of 
high-level, post-accident releases of radioactive noble gases; 
therefore, relocation to TS 3.3.3.7 is appropriate. In addition, TS 
definition 1.31, ``Source Checks,'' would be modified to allow 
different methods to comply with the source check requirement. This 
change would affect the remaining instruments in TS 3.3.3.1, and would 
allow for appropriate testing consistent with the technology of the 
existing detectors, and replacement detectors in the future.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the R46 monitors presents no change in 
the probability or the consequence of an accident, since the 
monitors are used post-accident for the monitoring of high-level 
releases of radioactive noble gases.
    Relocation of the R46 monitors to the accident monitoring TS 
3.3.3.7 is appropriate for the function of the monitors. The R46 
monitors are designed to meet the requirements of NUREG-0737 Il.F.1 
and the intent of RG [Regulatory Guide] 1.97. The monitor's alarm 
function is used in the EOPs [Emergency Operating Procedures] to 
identify a Steam Generator Tube Rupture (SGTR) event EOP entry point 
and to identify which SG [steam generator] has ruptured. The 
relocation of the monitor to TS 3.3.3.7 has no affect on the 
function of the monitor.
    The proposed change to the definition of TS 1.31 also does not 
impact the accident analyses in any manner. The qualitative 
assessment of monitor response will continue to be performed 
verifying monitor operability.
    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 relocation of the R46 monitors is primarily 
administrative in nature; there will be no change in the function of 
the monitors. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes. Post accident monitoring instrumentation is not associated 
with the initiation of an accident.
    The proposed change to the definition of TS 1.31 also does not 
create a new or different kind of accident. The qualitative 
assessment of monitor response will continue to be performed 
verifying monitor operability.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed change to relocate the R46 monitors does not alter 
the manner in which safety limits, limiting safety systems settings 
or limiting conditions for operation are determined. The proposed 
change will not alter any assumptions, initial conditions or results 
specified in any accident analysis. There is no change in the R46 
monitor alarm setpoint.
    The proposed change to the TS definition of SOURCE CHECK does 
not alter the basic requirement that a qualitative assessment of the 
monitor response be performed; therefore the operability of the 
monitor will continue to be verified.

    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Darrell J. Roberts.

PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, 
Unit No. 2, Salem County, New Jersey

    Date of amendment request: April 6, 2006.
    Description of the amendment request: The proposed amendment 
changes the existing steam generator (SG) tube surveillance program to 
one that is consistent with the program proposed by the Technical 
Specification Task Force (TSTF) in TSTF-449. These changes revise 
Technical Specification (TS) 1.15, ``Identified Leakage,'' TS 1.21, 
``Pressure Boundary Leakage,'' TS 3/4.4.6, ``Steam Generator (SG) Tube 
Integrity,'' and TS 3/4.4.7.2, ``Operational Leakage,'' and add new 
administrative TS 6.8.4.i, ``Steam Generator (SG) Program,'' and TS 
6.9.1.10, ``Steam Generator Tube Inspection Report.'' Other editorial 
changes were also made.
    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, cool down 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:
    All in-service steam generator tubes shall retain structural 
integrity over the full range of normal operating conditions 
(including startup, operation in the power range, hot standby, and 
cool down and all anticipated transients included in the design 
specification) and design basis accidents. This includes retaining a 
safety factor of 3.0 against burst under normal steady state full 
power operation primary-to-secondary pressure differential and a 
safety factor of 1.4 against burst applied to the design basis 
accident primary-to-secondary pressure differentials. Apart from the 
above requirements, additional loading conditions associated with 
the design basis accidents, or combination of accidents in 
accordance with the design and licensing basis, shall also be 
evaluated to determine if the associated loads contribute 
significantly to burst or collapse. In the assessment of tube 
integrity, those loads that do significantly affect burst or 
collapse shall be determined and assessed in combination with the 
loads due to pressure with a safety factor of 1.2 on the combined 
primary loads and 1.0 on axial secondary loads.
    The accident induced leakage performance criterion is:
    The primary-to-secondary accident induced leakage rate for any 
design basis accidents, other than a SG tube rupture, shall not 
exceed the leakage rate assumed in the accident analysis in terms of 
total leakage rate for all SGs and leakage rate for an individual 
SG. Leakage is not to exceed 1 gpm [gallon per minute] per SG.
    The operational leakage performance criterion is:
    The reactor coolant system operational primary-to-secondary 
leakage through any

[[Page 40754]]

one SG shall be limited to 150 gallons per day.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that are analyzed as part of a plant's licensing 
basis. In the analysis of an SGTR event, a bounding primary-to-
secondary leakage rate equal to the operational leakage rate limits 
in the licensing basis plus the leakage rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as main steam line break 
(MSLB), rod ejection, and reactor coolant pump locked rotor, the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses assume that primary-to-
secondary leakage for all SGs is 1 gallon per minute or increases to 
1 gallon per minute as a result of accident-induced stresses. The 
accident induced leakage criterion retained by the proposed changes 
accounts for tubes that may leak during design basis accidents. The 
accident induced leakage criterion limits this leakage to no more 
than the value assumed in the accident analysis.
    The SG performance criteria proposed as part of these TS changes 
identify the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining reactor coolant pressure 
boundary integrity throughout each operating cycle and in the 
unlikely event of a design basis accident. The performance criteria 
are only a part of the Steam Generator Program required by the 
proposed addition of TS 6.8.4.i. The program defined by NEI [Nuclear 
Energy Institute] 97-06 includes a framework that incorporates a 
balance of prevention, inspection, evaluation, repair, and leakage 
monitoring.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary-to-secondary leakage rates resulting from an accident. 
Therefore, limits are included in the Salem TS for operational 
leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure 
the plant is operated within its analyzed condition. The typical 
analysis of the limiting design basis accident assumes that primary-
to-secondary leak rate after the accident is 1 gallon per minute 
with no more than 500 gallons per day through any one SG, and that 
the reactor coolant activity levels of DOSE EQUIVALENT I-131 are at 
the TS values before the accident.
    The proposed change that allows SR [Surveillance Requirement] 
4.4.7.2.1.d to not be performed until 12 hours after establishment 
of steady state operation is consistent with NUREG 1431, ``Standard 
Technical Specifications, Westinghouse Plants'', and ensures the 
surveillance requirement is appropriate for the LCO [Limiting 
Condition for Operation].
    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 TS 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 TS.
    Therefore, the proposed changes do not affect the consequences 
of an SGTR accident and the probability of such an accident is 
reduced.
    In addition, the proposed changes do not affect the 
probabilities or consequences of an MSLB, rod ejection, or a reactor 
coolant pump locked rotor event.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed performance based requirements are an improvement 
over the requirements imposed by the current TS.
    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 changes do 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.
    The proposed change that allows SR 4.4.7.2.1.d to not be 
performed until 12 hours after establishment of steady state 
operation is consistent with NUREG 1431, ``Standard Technical 
Specifications, Westinghouse Plants'', and ensures the surveillance 
requirement is appropriate for the LCO.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical 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 TS.
    The proposed change that allows SR 4.4.7.2.1.d to not be 
performed until 12 hours after establishment of steady state 
operation is consistent with NUREG 1431, ``Standard Technical 
Specifications, Westinghouse Plants'', and ensures the surveillance 
requirement is appropriate for the LCO.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed changes to the 
TS.

    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Darrell J. Roberts.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: June 2, 2006.
    Description of amendment requests: The amendment proposes to revise 
Technical Specification (TS) 3.8.1, ``AC [alternating current] 
Sources--Operating,'' and TS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and 
Starting Air,'' to increase the required amount of stored diesel fuel 
oil to support a change to Ultra Low Sulfur Diesel fuel from California 
diesel fuel presently in use. This change in the type of fuel oil is 
mandated by California air pollution control regulations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This proposed change increases the minimum amount of stored 
diesel fuel. The change supports the use of Ultra Low Sulfur Diesel 
(ULSD) fuel rather than the existing California Air Resources Board 
diesel fuel as mandated by California air pollution control 
regulations (Title 13 California Code of

[[Page 40755]]

Regulations Division 3, Chapter 5, Article 2, Sections 2280-2285).
    Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, 
and Starting Air,'' requires that each diesel generator have 
sufficient fuel to operate for a period of 7 days, while the diesel 
generator (DG) is supplying maximum post Loss of Coolant Accident 
(LOCA) load demand.
    Because the Lower Heating Value (LHV) per gallon of ULSD fuel is 
less than that of existing diesel fuel, it was necessary to re-
calculate the amount of fuel required to supply necessary loads for 
the required time periods. For Modes 1 through 4, the resulting 
minimum volumes of ULSD fuel are 48,400 gallons and 41,800 gallons 
for the 7-day and 6-day fuel supply, respectively. For Modes 5 and 
6, the required volumes of ULSD fuel are 43,600 gallons and 37,400 
gallons for a 7-day supply and a 6-day supply, respectively.
    The DGs and the associated support systems such as the fuel oil 
storage and transfer systems are designed to mitigate accidents and 
are not accident initiators. Increasing the minimum volumes of 
stored fuel in the storage and day tanks will not result in a 
significant increase in the probability of any accident previously 
evaluated.
    Following implementation of this proposed change, there will be 
no change in the ability of the diesel generators to supply maximum 
post-LOCA load demand for 7 days. The proposed minimum volumes of 
fuel, 48,400 gallons and 41,800 gallons, ensure that a 7-day and [a] 
6-day supply of fuel, respectively, are available in Modes 1 through 
4. The proposed minimum volumes of fuel, 43,600 gallons and 37,400 
gallons, ensure that a 7-day and a 6-day supply, respectively, of 
fuel is available in Modes 5 and 6. This is identical to the current 
requirements, except for the increased volume of fuel required due 
to the decreased heat content of the ULSD fuel. Therefore, this 
change will not result in a significant increase in the consequences 
of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Following this change, the diesel generators will still be able 
to supply maximum post-LOCA load demand. The current 7-day and 6-day 
fuel supply requirements will be maintained following this change. 
The new required fuel oil volumes are within the capacities of the 
fuel oil storage tanks.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any accident that has 
been previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The Bases to TS 3.8.3 state that ``[e]ach diesel generator (DG) 
is provided with a storage tank having a fuel oil capacity 
sufficient to operate that diesel for a period of 7 days, while the 
DG is supplying maximum post loss of coolant accident load demand.'' 
When the fuel oil tank level is less than required to support the 7-
day of operation, the required action depends on whether or not a 6-
day supply of fuel is available.
    The proposed tank level limits will maintain these 7-day and 6-
day fuel supply requirements in all operating Modes following 
changeout to ULSD fuel.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: David Terao.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: May 25, 2006.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to adopt NRC-approved Revision 4 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler TSTF-372, ``Addition of LCO [Limiting 
Condition for Operation] 3.0.8, Inoperability of Snubbers.'' The 
amendment would add (1) a new LCO 3.0.8 addressing when one or more 
required snubbers are unable to perform their associated support 
function(s) (i.e., the snubber is inoperable) and (2) a reference to 
LCO 3.0.8 in LCO 3.0.1 on when LCOs shall be met.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on November 24, 2004 (69 FR 68412), on possible 
license amendments adopting TSTF-372 using the NRC's consolidated line 
item improvement process (CLIIP) for amending licensee's TSs, which 
included a model safety evaluation (SE) and model no significant 
hazards consideration (NSHC) determination. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 4, 2005 
(70 FR 23252), which included the resolution of public comments on the 
model SE. The May 4, 2005, notice of availability referenced the 
November 24, 2004, notice. The licensee has affirmed the applicability 
of the following NSHC determination in its application.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased, if at all. The consequences of an 
accident while relying on allowance provided by proposed LCO 3.0.8 
are no different than the consequences of an accident while relying 
on the TS required actions in effect without the allowance provided 
by proposed LCO 3.0.8. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

    Criterion 2--Does the proposed change 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 (no new or different type of equipment will be installed). 
Allowing delay times for entering [a] supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    Criterion 3--Does the proposed change involve a significant 
reduction in the margin of safety?
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following

[[Page 40756]]

the three-tiered approach recommended in [NRC] RG [Regulatory Guide] 
1.177. A bounding risk assessment was performed to justify the 
proposed TS changes. This application of LCO 3.0.8 is predicated 
upon the licensee's performance of a risk assessment and the 
management of plant risk [, which is required by the proposed TS 
3.0.8]. The net change to the margin of safety is insignificant. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: May 25, 2006.
    Description of amendment request: The amendment would revise 
Technical Specifications 3.1.7, ``Rod Position Indication,'' 3.2.1, 
``Heat Flux Hot Channel Factor (FCQ(Z)) (FQ 
Methodology),'' 3.2.4, ``Quadrant Power Tilt Ratio (QPTR),'' and 3.3.1, 
``Reactor Trip System (RTS) Instrumentation.'' The proposed changes are 
to allow use of the Westinghouse proprietary computer code, the Best 
Estimate Analyzer for Core Operations--Nuclear (BEACON). The new BEACON 
power distribution monitoring system (PDMS) would augment the 
functional capability of the neutron flux mapping system for the 
purposes of power distribution surveillances at the Callaway Plant. 
Certain required actions, for when a limiting condition for operation 
is not met, and certain surveillance requirements are being changed to 
refer to power distribution measurements or measurement information of 
the core.
    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 PDMS performs continuous core power distribution monitoring 
with data input from existing plant instrumentation. This system 
utilizes an NRC-approved Westinghouse proprietary computer code, 
i.e., Best Estimate Analyzer for Core Operations [mu] Nuclear 
(BEACON), to provide data reduction for incore flux maps, core 
parameter analysis, load follow operation simulation, and core 
predication. The PDMS does not provide any protection or control 
system function. Fission product barriers are not impacted by these 
proposed changes. The proposed changes occurring with PDMS will not 
result in any additional challenges to plant equipment that could 
increase the probability of any previously evaluated accident. The 
changes associated with the PDMS do not affect plant systems such 
that their function in the control of radiological consequences is 
adversely affected. These proposed changes will therefore not affect 
the mitigation of the radiological consequences of any accident 
described in the Final Safety Analysis Report (FSAR) [for the 
Callaway Plant].
    Use of the PDMS supports maintaining the core power distribution 
within required limits. Further continuous on-line monitoring 
through the use of PDMS provides significantly more information 
about the power distributions present in the core than is currently 
available. This results in more time (i.e., earlier determination of 
an adverse condition developing) for operation action prior to 
having an adverse condition develop that could lead to an accident 
condition or to unfavorable initial conditions for an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do[es] the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Other than use of the PDMS to monitor core power distribution, 
implementation of the PDMS and associated Technical Specification 
changes has no impact on plant operations or safety, nor does it 
contribute in any way to the probability or consequences of an 
accident. No safety-related equipment, safety function, or plant 
operation [other than core power distribution monitoring] will be 
altered as a result of this proposed change. The possibility for a 
new or different type of accident from any accident previously 
evaluated is not created since the changes associated with [the] 
implementation of the PDMS do not result in a change to the design 
basis of any plant component or system [other than to the PDMS]. The 
evaluation of the effects of using the PDMS to monitor core power 
distribution parameters shows that all design standards and 
applicable safety criteria limits are met. [The PDMS is to monitor 
the core power distribution and is, therefore, not an accident 
initiator.]
    The proposed changes do not result in any event previously 
deemed incredible being made credible [by the implementation of the 
PDMS]. Implementation of the PDMS will not result in any additional 
adverse condition and will not result in any increase in the 
challenges to safety systems. The cycle-specific variables required 
by the PDMS are calculated using NRC-approved methods. The Technical 
Specifications will continue to require operation within the 
required core operating limits, and appropriate actions will 
continue to be [required to be] taken when or if limits are 
exceeded.
    The proposed change, therefore, does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do[es] the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    No margin of safety is adversely affected by the implementation 
of the PDMS. The margins of safety provided by [the] current 
Technical Specification requirements and limits remain unchanged, as 
the Technical Specifications will continue to require operation 
within the core limits that are based on NRC-approved reload design 
methodologies. [These NRC-approved reload design methodologies are 
not being changed.] Appropriate measures exist to control the values 
of these cycle-specific limits, and appropriate actions will 
continue to be specified and [required to be] taken for when limits 
are violated. Such actions remain unchanged.
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: June 2, 2006.
    Description of amendment request: The amendment would revise 
Surveillance Requirement 3.5.2.8 in the Technical Specifications by 
replacing the phrase ``trash racks and screens'' with the word 
``strainers.'' The amendment reflects the replacement of the 
containment sump suction inlet trash racks and screens with a complex 
strainer design with significantly larger effective area in the 
upcoming Refueling Outage 15. This is in response to Generic Letter 
2004-02, ``Potential Impact of Debris Blockage on Emergency 
Recirculation during Design Basis Accidents at Pressurized-Water 
Reactors,'' dated September 13, 2004.
    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

[[Page 40757]]

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 consequences of accidents evaluated in the Updated Safety 
Analysis Report (USAR) [for the Wolf Creek Generating Station] that 
could be affected by the proposed change are those involving the 
pressurization of containment and associated flooding of the 
containment and recirculation of this fluid within the Emergency 
Core Cooling System (ECCS) or the Containment Spray System (CSS) 
(e.g., Loss of Coolant Accidents). The proposed change does not 
impact the initiation or probability of occurrence of any accident. 
[The containment sump trash racks and screens, and the sump 
strainers that are replacing the trash racks and screens are not 
initiators of accidents.]
    Although the configurations of the existing containment 
recirculation sump trash racks and screen[s,] and the replacement 
sump strainer assemblies are different, they serve the same 
fundamental purpose of passively removing debris from the sump's 
suction supply of the supported system pumps. Removal of trash racks 
does not impact the adequacy of the pump NPSH [net positive suction 
head] assumed in the safety analysis. Likewise, the change does not 
reduce the reliability of any supported systems or introduce any new 
system interactions. The greatly increased surface area of the new 
strainer is designed to reduce head loss [at the containment sump] 
and reduce the approach velocity at the strainer face significantly, 
decreasing the risk of impact from large debris entrained in the 
sump flow stream.
    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 containment recirculation sump strainers are a passive 
system used for accident mitigation. As such, they cannot be 
accident initiators. Therefore, there is no possibility that this 
change could create any new or different kind of accident.
    No new accident scenarios, transient precursors, or limiting 
single failures are introduced as a result of the proposed change. 
There will be no adverse effect[s] or challenges imposed on any 
safety related system as a result of the change. Therefore, the 
possibility of a new or different type of accident is not created. 
[The containment recirculation sump suction inlet trash racks and 
screens are being replaced with a complex strainer design with 
significantly larger effective surface area to reduce head loss and 
reduce the approach velocity at the strainer face significantly, 
decreasing the risk of impact from large debris entrained in the 
sump flow stream.]
    There are no changes which would cause the malfunction of safety 
related equipment, assumed to be OPERABLE in the accident analyses, 
as a result of the proposed Technical Specification change. No new 
equipment performance burdens are imposed. The possibility of a 
malfunction of safety related equipment with a different result [or 
consequences] is not created.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind of] accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. The proposed change does not adversely affect the fuel, 
fuel cladding, Reactor Coolant System, or containment integrity. The 
radiological dose consequence acceptance criteria listed in the 
Standard Review Plan [for accidents] will continue to be met.
    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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of application for amendments: March 9, 2005, as supplemented 
by letter dated July 7, 2005.
    Brief description of amendments: The amendments revised the 
Millstone Power Station, Unit Nos. 2 and 3 Technical Specifications to 
incorporate wording related to the reactor coolant system, electrical 
power system and refueling operations to provide operational 
flexibility during mode changes or addition of coolant during shutdown 
operations.
    Date of issuance: June 28, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 293 and 230.
    Facility Operating License Nos. DPR-65 and NPF-49: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29788). The additional information provided in

[[Page 40758]]

the supplemental letter dated July 7, 2005, did not expand the scope of 
the application as noticed and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 28, 2006.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: January 5, 2005, as supplemented 
November 21, 2005.
    Brief description of amendments: The amendments revised Technical 
Specifications (TSs) 5.5.19.b, 5.1.19.c, and TS Surveillance 
Requirement (SR) 3.8.1.9 associated with the Lee Combustion Turbine 
(LCT) testing program. TS 5.5.19 required verification that an LCT can 
supply the equivalent of one unit's maximum safeguards loads, plus two 
units' Mode 3 loads when connected to the system grid every 12 months. 
The amendments clarified this requirement as ``Verify an LCT can supply 
equivalent of one unit's Loss of Coolant Accident (LOCA) loads plus two 
units' Loss of Offsite Power (LOOP) loads when connected to system grid 
every 12 months.'' TS 5.5.19.c and SR 3.8.1.9 were revised for 
consistency.
    Date of Issuance: July 5, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 352/354/353.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7764). The additional information provided in the supplemental 
letter dated November 21, 2005, did not expand the scope of the 
application as noticed and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 5, 2006.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: December 29, 2005.
    Brief description of amendment: The amendment deleted License 
Condition, Section 2.F, that requires the reporting of violations in 
Section 2.C of the Facility Operating License.
    Date of issuance: June 28, 2006.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 116.
    Facility Operating License No. NPF-69: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: April 25, 2006 (71 FR 
23958).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 28, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 17, 2006.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) adding Limiting Condition for Operation 
(LCO) 3.0.8 to allow a delay time for entering a supported system TS 
when the inoperability is due solely to an inoperable snubber, if risk 
is assessed and managed consistent with the program in place for 
complying with the requirements of 10 CFR 50.65(a)(4).
    Date of issuance: June 29, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 250/194.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: April 25, 2006 (71 FR 
23960).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 29, 2006.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Unit Nos. 1 and 2, Burke County, 
Georgia

    Date of application for amendments: February 17, 2006.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) adding Limiting Condition for Operation 
(LCO) 3.0.8 and renumbering existing LCO 3.0.8 to LCO 3.0.9 to allow a 
delay time for entering a supported system TS when the inoperability is 
due solely to an inoperable snubber, if risk is assessed and managed 
consistent with the program in place for complying with the 
requirements of 10 CFR 50.65(a)(4).
    Date of issuance: June 29, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 141/121.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Licenses and the Technical Specifications.
    Date of initial notice in Federal Register: April 25, 2006 (71 FR 
23960).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 29, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 19, 2005, as 
supplemented by letter dated March 30, 2006.
    Brief description of amendments: The amendments modified several 
parts of Technical Specification Surveillance Requirement (SR) 4.0.5, 
both to change the surveillance intervals for which the 25 percent 
extension provided in SR 3.0.2 would apply, and to replace the 
references in SR 4.0.5 to the American Society of Mechanical Engineers 
(ASME) Boiler and Pressure Vessel Code, Section XI, with the ASME 
Operation and Maintenance Code.
    Date of issuance: June 16, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 308 and 297.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the technical specifications.
    Date of initial notice in Federal Register: February 14, 2006 (71 
FR 7183).
    The supplemental letter dated March 30, 2006, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 16, 2006.
    No significant hazards consideration comments received: No.

[[Page 40759]]

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of application for amendments: April 20, 2006, as supplemented 
on May 15, 2006.
    Brief description of amendments: These amendments revised the 
reactor coolant pressure and temperature limits, low-temperature 
overpressure protection system (LTOPS) setpoint values, and LTOPS 
enable temperatures for up to 28.8 effective full-power years (EFPYs) 
and 29.4 EFPYs of operation at Surry Power Station, Unit Nos. 1 and 2, 
respectively.
    Date of issuance: June 29, 2006.
    Effective date: As of the date of issuance.
    Amendment Nos.: 248/247.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 2006 (71 FR 
25249).
    The May 15, 2006, supplement contained clarifying information only 
and did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 29, 2006.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 11th day of July.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 06-6246 Filed 7-17-06; 8:45 am]
BILLING CODE 7590-01-P