Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 40742-40759 [06-6246]

Agencies

• NUCLEAR REGULATORY COMMISSION

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


-----------------------------------------------------------------------

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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the 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, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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