Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 23952-23970 [06-3901]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 79 (Tuesday, April 25, 2006)]
[Notices]
[Pages 23952-23970]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-3901]


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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 23953]]

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 March 31, 2006 to April 13, 2006. The last 
biweekly notice was published on April 11, 2006 (71 FR 18371).

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 23954]]

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

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

    Date of amendment request: March 17, 2006.
    Description of amendment request: The proposed amendment would 
change the design criteria described in the Kewaunee Power Station 
(KPS) Updated Safety Analysis Report (USAR). The change would add new 
design criteria associated with internal flooding to the current 
licensing basis for KPS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change provides clarification to the existing 
functional requirements in the USAR by including specific design 
criteria for analyzing internal flooding in order to verify the 
capability of an SSC [structure, systems and components] to perform 
its design function. The proposed change does not affect any of the 
previously evaluated accidents in the KPS updated safety analysis 
report (USAR). No SSCs, operating procedures, or administrative 
controls that have the function of preventing or mitigating any of 
these accidents are affected.
    This proposed change to incorporate design criteria into the 
USAR provides added administrative assurance that internal flooding 
will be appropriately addressed, consistent with existing functional 
requirements, and that safety related SSCs will not be affected by a 
potential failure of a non-safety related SSC. The change does not 
affect any accident initiators or the facility accident analysis. 
Thus, the probability and the consequences of an accident remain 
unchanged.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to incorporate design criteria consistent 
with existing functional requirements into the USAR does not change 
the design function or operation of any safety related SSCs. The 
proposed change documents design criteria in use and therefore does 
not involve a physical change to the facility. The change, 
therefore, does not create the possibility of a new or different 
kind of accident due to credible new failure mechanisms, 
malfunctions, or accident initiators not considered in the design 
and licensing bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This proposed change does not affect any margin of safety as 
established in the Kewaunee USAR because it documents the design 
criteria presently used and is consistent with the functional 
requirements in the USAR. This proposed change provides added 
administrative assurance that safety related SSCs will not be 
affected by a potential failure of a non-safety related SSC due to a 
postulated internal flooding event. The proposed change adds 
criteria for the evaluation of internal flooding events that are 
more detailed than the existing functional requirements in the USAR. 
Therefore, the protection and subsequent availability of safety 
related SSCs is maintained consistent with previously assumed 
accident mitigation capabilities.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Branch Chief: L. Raghavan.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of amendment request: January 12, 2006.
    Description of amendment request: The proposed amendment would 
correctly modify the wording in Technical Specification Surveillance 
Requirement (SR) 3.6.6.3 Containment Cooling train cooling water flow 
rate to accurately reflect the plant configuration. The current SR is 
to verify flow to each train. The proposed revision to SR 3.6.6.3 would 
verify flow to each cooler (plant configuration is two coolers per 
train).
    Basis for proposed no significant hazards consideration 
determination:

[[Page 23955]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change will revise Technical Specifications (TS) 
Surveillance Requirement (SR) 3.6.6.3 containment cooling train 
cooling water flow rate to accurately reflect the existing plant 
configuration as described in the Updated Final Safety Analysis 
Report (UFSAR) Sections 6.2, ``Containment Systems,'' and 9.4, ``Air 
Conditioning, Heating, Cooling, and Ventilation Systems.'' The 
revision will specify the appropriate testing requirements for 
verification that each Containment Cooling System train Essential 
Service Water (SX) flow rate to each cooling unit is >= 2660 gpm 
[gallons per minute] and will therefore provide assurance that the 
design flow rate assumed in the safety analyses will be achieved and 
the Limited Conditions for Operation (LCO) will be met. This change 
is in the conservative direction, i.e., verification of flow rate to 
each cooling unit 3 2660 gpm is more conservative than 
verification of the same flow rate to each cooling train that 
consists of two cooling units. The performance of TS surveillance 
testing is not a precursor to any accident previously evaluated. 
Thus, the proposed change does not have any effect on the 
probability of an accident previously evaluated.
    The function of the Containment Cooling System in conjunction 
with the Containment Spray System is to provide containment 
atmosphere cooling to limit post accident pressure and temperature 
in containment to less than design values. There is no change to the 
design of the Containment Cooling System. Furthermore, the 
surveillance testing specified in SR 3.6.6.3 will provide assurance 
that the Containment Cooling System will perform as designed. Thus, 
the radiological consequences of any accident previously evaluated 
are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. The proposed change does not change or 
introduce any new equipment, modes of system operation or failure 
mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    Prior to conversion to ITS [Improved Technical Specifications], 
the SR equivalent to SR 3.6.6.3 required that each system of 
containment cooling fans be demonstrated OPERABLE by ``verifying an 
essential service water flow rate of greater than or equal to 2660 
gpm to each cooler.'' During the ITS conversion, standard verbiage 
for SR 3.6.6.3 was adopted; however, the specific plant design of 
two Reactor Containment Fan Coolers (RCFCs) per Containment Cooling 
train was inadvertently overlooked.
    This proposed amendment would correctly modify the wording in 
Technical Specifications (TS) Surveillance Requirement (SR) 3.6.6.3 
Containment Cooling System to accurately reflect the Braidwood and 
Byron existing plant design. The revision will provide the 
appropriate testing requirements for verification that each 
Containment Cooling System train SX cooling flow rate to each 
cooling unit is >= 2660 gpm. This verification provides assurance 
that the design flow rate assumed in the safety analyses will be 
achieved; and, therefore the LCO will be met. The change for 
verification of SX cooling flow rate from each cooling train to each 
cooling unit is in the conservative direction and will revise the 
existing non-conservative TS SR to be consistent with the plant 
design as described in the UFSAR.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

FAL Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: March 23, 2006.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station, Unit No. 1 (Seabrook) Operating License 
and Technical Specifications (TSs) to delete the license condition 
requiring reporting of violations of other requirements (e.g., 
conditions listed in Section 2.C of the operating license). The change 
is consistent with the notice published in the Federal Register on 
November 4, 2005, as part of the consolidated line item improvement 
process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and, therefore, does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the change does not create the 
possibility of a new of 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 deletes a reporting requirement. The change 
does not affect plant equipment or operation practices and, 
therefore, does not involve a significant reduction in a margin of 
safety.

    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, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Darrell J. Roberts.

FPL Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: March 23, 2006.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station Unit No. 1 (Seabrook) Technical 
Specifications (TSs) consistent with the NRC-approved Revision 4 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.''
    Additionally, the proposed amendment would revise Seabrook TS 
Surveillance Requirement 4.4.6.2.1 to be consistent with NUREG-1431, 
Revision 3, Improved Standard Technical Specifications Westinghouse 
Plants.
    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

[[Page 23956]]

(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 March 23, 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 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] to 
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.
    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, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Darrell J. Roberts.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: March 7, 2006.
    Description of amendment requests: The proposed amendments would 
modify the Technical Specifications (TS) of the units to change the 
reactor trip on turbine trip from the P-7 interlock to the P-8 
interlock. Specifically, the amendment would effect changes in TS Table 
3.3.1-1, ``Reactor Trip System Instrumentation,'' for Function 16, 
``Turbine Trip.'' The purpose of the proposed amendment is to decrease 
potentially unnecessary transients on the reactor and to increase plant 
availability when the cause of a turbine trip is readily correctable.
    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 as follows:

    (1) Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?

[[Page 23957]]

    Response: No.
    The proposed change revises the setpoint at which a reactor trip 
will occur by changing the interlock at which it is enabled from the 
P-7 interlock, at approximately 10 percent power, to the P-8 
interlock, at less than or equal to 31 percent power. The P-7 and P-
8 interlocks are not accident initiators and the change to the 
reactor trip setpoint does not create any new credible single 
failure. An analysis has shown that a turbine trip without a reactor 
trip at 31 percent power or below does not challenge the pressurizer 
power operated relief valves (PORVs), thereby not adversely 
affecting the probability of a small[-]break loss[-]of [-]coolant 
accident due to a stuck open PORV. The consequences of accidents 
previously evaluated are unaffected by this change because no change 
to any accident mitigation scenario has resulted and there are no 
additional challenges to fission product barrier integrity.
    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.
    No changes are being made to the plant that would introduce any 
new accident causal mechanisms. The proposed change to the power 
level at which a reactor trip on turbine trip is enabled does not 
adversely affect previously identified accident initiators and does 
not create any new accident initiators. The change does not affect 
how the associated trip function operates. No new single failures or 
accident scenarios are created by the proposed change and the 
proposed change does not result in any event previously deemed 
incredible being made credible.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    No safety analyses [will be] changed or modified as a result of 
the proposed change in reactor trip setpoint. All margins associated 
with the current safety analyses acceptance criteria are unaffected. 
The current safety analyses remain binding. The safety systems 
credited in the safety analyses will continue to be available to 
perform their mitigation functions. The proposed change does not 
affect the availability or operability of safety-related systems and 
components.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    Based on the licensee's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the requested amendments involve no 
significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: January 30, 2006.
    Description of amendment request: The proposed change would revise 
Cooper Nuclear Station (CNS) Technical Specification section 5.5.12, 
``Primary Containment Leakage Rate Testing Program,'' to allow a one-
time extension of no more than 5 years for the Type A, Integrated 
Leakage Rate Test (ILRT) interval. This revision is a one-time 
exception to the 10-year frequency of the performance-based leakage 
rate testing program for Type A tests as defined in Nuclear Energy 
Institute (NEI) document NEI 94-01, Revision 0, ``Industry Guideline 
for Implementing Performance-Based Option of 10 CFR part 50, appendix 
J,'' pursuant to 10 CFR 50, appendix J, option B. The requested 
exception is to allow the ILRT to be performed within 15 years from the 
last ILRT, last performed on December 7, 1998.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment proposes to revise the Technical 
Specifications to allow for a one-time extension of the ILRT 
interval from 10 years to 15 years. The containment function is 
solely to mitigate the consequences of an accident. No design basis 
accident is initiated by a failure of the containment leakage 
mitigation function. The extension of the ILRT will not create any 
adverse interactions with other systems that could result in 
initiation of a design basis accident. Continued containment 
integrity is also assured by the established programs for local 
leakage rate testing and inservice inspections which are unaffected 
by the proposed change. Therefore, the probability of occurrence of 
an accident previously evaluated is not significantly increased.
    The potential consequences of the proposed change have been 
quantified by analyzing the changes in risk that would result from 
extending the ILRT interval from 10 to 15 years. The increase in 
risk in terms of person-rem per year within 50 miles resulting from 
accidents was determined to be of a magnitude that NUREG-1493 
indicates is imperceptible. NPPD [Nebraska Public Power District] 
has also analyzed the increase in risk in terms of the frequency of 
large early releases from accidents. The increase in the large early 
release frequency resulting from the proposed extension was 
determined to be within the guidelines published in Nuclear 
Regulatory Commission (NRC) Regulatory Guide 1.174. Additionally, 
the proposed change maintains defense-in-depth by preserving a 
reasonable balance among prevention of core damage, prevention of 
containment failure, and consequence mitigation. NPPD has determined 
that the increase in conditional containment failure probability 
from reducing the ILRT frequency from one test in 10 years to one 
test in 15 years would be insignificant.
    Therefore, the probability of occurrence or the consequences of 
an accident previously analyzed are not significantly increased.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed extension of the current interval for the ILRT does 
not involve any change to the design or operation of any plant 
structure, system, or component (SSC). The plant will continue to be 
operated in the same manner. Since no changes to the design or 
operation of the plant are being made, the proposed one-time 
extension of the ILRT does not result in a new failure mode for an 
accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed extension to the ILRT test interval will not result 
in a change to the design or operation of any plant SSC used to shut 
down the plant, initiate Emergency Core Cooling Systems, or isolate 
the primary or secondary containment. Thus, the change will not 
impact the ability of CNS to mitigate any accident or transient. 
NUREG-1493, a generic study of the effects of extending containment 
leakage testing, documented that an extension in the ILRT interval 
from three per 10 years to one per 20 years resulted in an 
imperceptible increase in risk to the public. NUREG-1493 generically 
concluded that the design containment leakage rate contributes about 
0.1 percent to the individual risk, and that the decrease in the 
ILRT frequency would have a minimal effect on this risk since 95% of 
the potential leakage paths are detected by Type B and Type C 
testing. A risk assessment using the current CNS Probabilistic 
Safety Assessment internal events model concluded that the risk 
associated with this change is very small and not risk significant.
    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 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.

[[Page 23958]]

    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: March 15, 2006.
    Description of amendment request: The proposed amendment would 
revise Cooper Nuclear Station (CNS) Technical Specification 5.5.12, 
``Primary Containment Leakage Rate Testing Program,'' by adding two 
sub-paragraphs to note exemptions from Section III.A and Section III.B 
of Part 50 of Title 10 of the Code of Federal Regulations, Appendix J, 
Option B. These two sub-paragraphs allow the leakage contribution from 
the four main steam line penetrations, referred to as the Main Steam 
Isolation Valve (MSIV) leakage, to be excluded.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This proposed change to TS 5.5.12 does not modify existing 
structures, systems or components (SSC's) of the plant, and it does 
not introduce new SSC's. It does not change assumptions, methodology 
or results of previously evaluated accidents in the Updated Safety 
Analysis Report.
    It does not change operating procedures or administrative 
controls that affect the functions of SSC's. By excluding MSIV 
leakage from Type A and Type B and C test results, this change will 
make the CNS Primary Containment Leakage Rate Testing Program more 
closely aligned with the assumptions used in associated accident 
consequence analyses. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This proposed change to TS 5.5.12.a does not modify existing 
SSC's of the plant, and it does not introduce new SSC's. Thus, it 
does not affect the design function or operation of SSC's involved, 
and it does not introduce a new accident initiator. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Since MSIV leakage bypasses the containment and its filtration 
system (Standby Gas Treatment System) during a Loss-of-Coolant 
Accident (LOCA), the effects on release to the environment [are] 
analyzed and specifically accounted for in the CNS dose analysis 
methodology approved by Amendments 196 and 206. This proposed change 
to exclude MSIV leakage from Type A and Type B and C test results 
does not change dose analysis values, and thus, does not affect 
actual margin in the dose analysis. Therefore, the proposed change 
does not involve a significant reduction in an actual margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: December 29, 2005.
    Description of amendment request: The proposed change would delete 
Section 2.F of the Nine Mile Point, Unit 2 Facility Operating License 
(FOL), NPF-69, which requires the licensee report violations of the 
requirements contained in Section 2.C of this license. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
August 29, 2005 (70 FR 51098), on possible amendments to delete this 
reporting requirement, 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 November 4, 
2005 (70 FR 67202). The licensee affirmed the applicability of the 
following NSHC determination in its application dated December 29, 
2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the deletion of a reporting 
requirement. The change does not affect any plant equipment or 
operating practices and therefore does not significantly increase 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a margin of safety.

    Based on the above, 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: March 23, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.4, ``Loss of Power (LOP) Diesel 
Generator (DG) Start and Load Sequence Instrumentation''. The revision 
modifies the section title and corrects a nonconservatism in the 
degraded voltage time delay values in TS Surveillance Requirement (SR) 
3.3.4.3.b.
    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. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The diesel generators (DGs) provide emergency electrical power 
to the safeguard

[[Page 23959]]

buses in support of equipment required to mitigate the consequences 
of design basis accidents and anticipated operational occurrences, 
including an assumed loss of all offsite power. SR 3.3.4.3 verifies 
that the loss of power (LOP) DG start instrumentation channels 
respond to measured parameters within the necessary range and 
accuracy. The proposed amendment revises the section title and 
corrects nonconservative values in the allowed time delays for the 
degraded voltage protection function. The revised values are more 
restrictive than the previously allowed values.
    Reducing the time delays for the degraded voltage function as 
proposed does not significantly increase the probability of a loss 
of offsite power event. The degraded voltage analysis established 
both maximum time delay limits for a degraded voltage condition and 
minimum time delays to prevent premature disconnection from offsite 
power. The analyzed time delay limits considered prevention of 
premature disconnection from offsite power such that the probability 
of an unnecessary loss of offsite power is not significantly 
increased.
    The proposed change does not involve any hardware changes, nor 
does it affect the probability of any event initiators. There will 
be no change to normal plant operating parameters, accident 
mitigation capabilities, or accident analysis assumptions or inputs.
    Therefore, the probability or consequences of any accident 
previously evaluated will not be significantly increased as a result 
of the proposed change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. The revised surveillance requirements are 
more restrictive and will continue to assure equipment reliability 
such that plant safety is maintained or will be enhanced.
    Equipment important to safety will continue to operate as 
designed. The changes do not result in any event previously deemed 
incredible being made credible. The changes do not result in adverse 
conditions or result in any increase in the challenges to safety 
systems. Therefore, operation of the Point Beach Nuclear Plant in 
accordance with the proposed amendment will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    The diesel generators (DGs) provide emergency electrical power 
to the safeguard buses in support of equipment required to mitigate 
the consequences of design basis accidents and anticipated 
operational occurrences, including an assumed loss of all offsite 
power. SR 3.3.4.3 verifies that the loss of power (LOP) DG start 
instrumentation channels respond to measured parameters within the 
necessary range and accuracy. The proposed amendment corrects 
nonconservative values in the allowed time delays for the degraded 
voltage protection function. The revised values are more restrictive 
than the previously allowed values. The proposed change to this SR 
assures that design requirements of the emergency electrical power 
system continue to be met.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. The 
proposed amendment will not otherwise affect the plant protective 
boundaries, will not cause a release of fission products to the 
public, nor will it degrade the performance of any other structures, 
systems or components (SSCs) important to safety. Therefore, the 
requested change will not result in 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of amendment request: February 1, 2006.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements for inoperable 
snubbers by adding Limiting Condition for Operation (LCO) 3.0.8 for 
SSES 1 and 2. This change is based on the TS Task Force (TSTF) change 
traveler TSTF-372, Revision 4. A notice of availability for this TS 
improvement using the consolidated line item improvement process was 
published in the Federal Register on November 24, 2004, and May 4, 
2005.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
availability of a model no significant hazards consideration (NSHC) 
determination for referencing license amendment applications in the 
Federal Register on November 24, 2004 (69 FR 68412), and May 4, 2005 
(70 FR 23252). The licensee affirmed the applicability of the model 
NSHC determination in its application dated February 1, 2006.
    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. Criterion 1--The Proposed Change Does Not Involve a 
Significant Increase in the Probability or Consequences of an 
Accident Previously Evaluated.
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased, if at all. The consequences of an 
accident while relying on allowance provided by proposed LCO 3.0.8 
are no different than the consequences of an accident while relying 
on the TS required actions in effect without the allowance provided 
by proposed LCO 3.0.8. Therefore the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Criterion 2--The Proposed Change Does Not Create the 
Possibility of a New or Different Kind of Accident From Any 
Previously Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns.
    Thus, this change does not create the possibility of a new or 
different kind of accident from an accident previously evaluated.
    3. Criterion 3--The Proposed Change Does Not Involve a 
Significant Reduction in the Margin of Safety.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG 1.177. A bounding risk assessment was performed to 
justify the proposed TS changes. This application of LCO 3.0.8 is 
predicated upon the licensee's performance

[[Page 23960]]

of a risk assessment and the management of plant risk. The net 
change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Richard J. Laufer.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: March 28, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement 3.5.1.4 by 
changing the method and sample frequency for boron concentration 
verification for the emergency core cooling system (ECCS) accumulators.
    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 ECCS Accumulators are used only to respond to an accident 
and are not an accident initiator. Therefore, the probability of an 
accident has not increased.
    Boron concentration is controlled in the ECCS Accumulators to 
prevent either excessive boron concentrations or insufficient boron 
concentrations. Post-loss-of-coolant accident (LOCA) emergency 
procedures directing the operator to establish simultaneous hot and 
cold leg injection are based on the worst case minimum boron 
precipitation time. Maintaining the maximum ECCS Accumulator boron 
concentration within the upper limit ensures that the ECCS 
Accumulators do not invalidate these steps. The minimum boron 
requirements of 2100 (2550 after EPU [extended power uprate]) ppm 
[parts per million] ppm are based on beginning-of-life reactivity 
values and are selected to ensure that the reactor will remain 
subcritical during the reflood stage of a large break LOCA. During a 
large break LOCA, all control element assemblies are assumed not to 
insert into the core, and the initial reactor shutdown is 
accomplished by void formation during blowdown. Sufficient boron 
concentration must be maintained in the ECCS Accumulators to prevent 
a return to criticality during reflood. Level and pressure 
instrumentation is provided to monitor the availability of the ECCS 
Accumulators during plant operation.
    The Technical Specification Surveillance Requirement (SR 
3.5.1.4) verifies that the boron concentration remains within the 
required range by sampling. Currently, the boron concentration in 
each ECCS Accumulator is required to be verified by taking a sample 
of the water in the ECCS Accumulator every 31 days on a staggered 
test basis. A containment entry is required to take a sample from 
each of the two ECCS Accumulators. In addition, the makeup water 
source for the ECCS Accumulators is from the RWST [refueling water 
storage tank], which is maintained between 2300 ppm and 2600 ppm 
(2750 and 3050 after EPU) by SR 3.5.4.2, ensuring the ECCS 
Accumulators are not diluted during makeup/fill evolutions. However, 
the Reactor Coolant System boron concentration is lower during power 
operation than the boron concentration in the ECCS Accumulators. Two 
check valves in series prevent leakage from the Reactor Coolant 
System into the ECCS Accumulators.
    This proposed amendment would require inleakage monitoring to be 
done every twelve hours in addition to taking samples from each ECCS 
Accumulator every six months. Samples would continue to be taken to 
verify the inleakage observations remain conservative.
    The engineering analysis and risk insights combine to 
demonstrate that the method of ECCS Accumulator boron concentration 
verification can be changed from sampling every 31 days on a 
staggered test basis to monitoring inleakage every twelve hours and 
sampling each ECCS Accumulator every six months. The inleakage 
monitoring is based on a calculational method that has sufficient 
conservatism to predict the boron concentration of the ECCS 
Accumulator as shown by sample. Therefore, the ECCS Accumulator 
would remain capable of responding to an accident as described above 
and the consequences of an accident previously evaluated are not 
increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not alter the function of any 
equipment, nor cause it to operate differently than it was designed 
to operate. All equipment required to mitigate the consequences of 
an accident would continue to operate as before. The proposed change 
alters the method of verification of the ECCS Accumulator boron 
concentration, but not the boron concentration requirements 
themselves.
    Therefore, this 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 inleakage monitoring done to verify the concentration of 
boron in the ECCS Accumulators, is sufficiently conservative to 
ensure that a decrease in boron concentration would be detected, 
leading to attempts to increase the boron concentration or a need to 
sample the affected ECCS Accumulator. Sampling of the ECCS 
Accumulators every six months will continue to be done to ensure 
that the inleakage monitoring remains conservative and 
representative. If the boron concentration is maintained in the ECCS 
Accumulators, the system operates as assumed in the Updated Final 
Safety Analysis Report Chapter 15 analyses and the analyses 
continues to meet the dose consequences acceptance criteria given in 
the Updated Final Safety Analysis Report.
    Therefore, this 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 t