Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 75489-75500 [05-24142]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 243 (Tuesday, December 20, 2005)]
[Notices]
[Pages 75489-75500]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-24142]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 23, 2005 to December 8, 2005. The 
last biweekly notice was published on December 6, 2005 (70 FR 72667).

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

[[Page 75490]]

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 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, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station (OCNGS), Ocean County, New Jersey

    Date of amendment request: October 18, 2005.
    Description of amendment request: The licensee proposes to revise 
the OCNGS Technical Specifications Surveillance Requirement 4.4.B.1 to 
provide an alternative means for testing the electromatic relief valves 
located on the main steam system. The proposed change would allow 
demonstration of the capability of the valves to perform their function 
without requiring that the valves be cycled with steam pressure while 
installed.
    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. The licensee's analysis 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 modifies Technical Specifications (TS) 
Surveillance Requirement (SR) 4.4.B.1 to provide an alternative 
means for testing the Electromatic Relief Valves (EMRVs). Accidents 
are initiated by the malfunction of plant equipment, or the failure 
of plant structures, systems, or components. The performance of EMRV 
testing is not a precursor to any accident previously evaluated and 
does not change the manner in which the valves are operated. The 
proposed testing requirements will not contribute to the failure of 
the relief valves nor any plant structure, system, or component. 
AmerGen Energy Company, LLC (AmerGen) has determined that the 
proposed change in testing methodology provides an equivalent level 
of assurance that the relief valves are capable of performing their 
intended safety functions. Thus, the proposed change does not affect 
the probability of an accident previously evaluated.
    The performance of EMRV testing provides confidence that the 
EMRVs are capable of depressurizing the reactor pressure vessel 
(RPV). This will protect the reactor vessel from overpressurization 
and allow the Core Spray system to inject into the RPV as designed. 
The proposed change involves the manner in which the EMRVs are 
tested, and has no effect on the types or amounts of

[[Page 75491]]

radiation released or the predicted offsite doses in the event of an 
accident. The proposed testing requirements are sufficient to 
provide confidence that the EMRVs are capable of performing their 
intended safety functions. In addition, a stuck open EMRV accident 
is analyzed in the Updated Final Safety Analysis Report (section 
15.6.1). Since the proposed testing requirements do not alter the 
assumptions for the stuck open EMRV accident, the consequences of 
any accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    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 affect the assumed accident 
performance of the EMRVs, nor any plant structure, system, or 
component previously evaluated. The proposed change does not involve 
the installation of new equipment, and installed equipment is not 
being operated in a new of different manner. The change in test 
methodology ensures that the EMRVs remain capable of performing 
their safety functions. No set points are being changed which would 
alter the dynamic response of plant equipment. Accordingly, no new 
failure modes are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will allow testing of the EMRV actuation 
electrical circuitry, including the solenoid, and mechanical 
actuation components, without causing the EMRV to open. Accordingly, 
in-situ EMRV cycling is avoided, reducing the potential for valve 
seat leakage. The valves will be tested in accordance with the 
Inservice Test (IST) Program that involves testing the valve at a 
test facility using steam. The combination of the IST and proposed 
actuator test provides confidence that the EMRVs will perform their 
design function.
    The proposed change does not affect the EMRV set points or the 
operational criteria that directs the EMRVs to be manually opened 
during plant transients. There are no changes proposed which alter 
the set points at which protective actions are initiated, and there 
is no change to the operability requirements for equipment assumed 
to operate for accident mitigation.
    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 
proposed amendment involves no significant hazards consideration.
    Attorney for licensee: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Darrell J. Roberts.

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

    Date of amendment request: October 27, 2005.
    Description of amendment request: This amendment proposes revisions 
to the Technical Specification (TS) Surveillance Requirements (SR) 
4.5.2e (Safety Injection), 4.6.2.1d (Containment Spray), and 4.7.3b 
(Component Cooling Water/Auxiliary Component Cooling Water), by 
removing the words ``during shutdown.'' Additionally, a revision to 
delete TS SR 4.7.12.1c (Essential Services Chilled Water) is requested.
    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.
    Deletion of TS SR 4.7.12.1c is an administrative change since 
there are no valves in the essential services chilled water system 
for which the TS SR 4.7.12.1c is applicable. The deletion of the 
``during shutdown'' restriction from TS SRs 4.5.2e (Safety 
Injection), 4.6.2.1d (Containment Spray), and 4.7.3b (Component 
Cooling Water/Auxiliary Component Cooling Water) does not impact 
system operation nor does it reduce TS SRs. Component actuations 
that will be allowed to be performed online for these TS SRs are 
either already actuated online for other TS SRs or the components to 
be actuated online are currently stroked online in accordance with 
the Inservice Testing Program. Therefore, the accident mitigation 
features of the plant for previously evaluated accidents are not 
affected by the proposed amendment.
    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.
    Deletion of TS SR 4.7.12.1c is an administrative change since 
there are no valves in the essential services chilled water system 
for which the TS SR 4.7.12.1c is applicable. The deletion of the 
``during shutdown'' restriction from TS SRs 4.5.2e (Safety 
Injection), 4.6.2.1d (Containment Spray), and 4.7.3b (Component 
Cooling Water/Auxiliary Component Cooling Water) does not impact 
system operation nor does it reduce TS SR. Component actuations that 
will be allowed to be performed online for these TS SRs are either 
already actuated online for other TS SRs or the components to be 
actuated online are currently stroked online in accordance with the 
Inservice Testing Program. Therefore, the proposed change introduces 
no new mode of plant operation and no new possibility for an 
accident is introduced.
    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.
    There are no automatic valves in the essential services chilled 
water system that actuate on an SIAS [safety injection actuation 
signal]. Deletion of the ``during shutdown'' limitation does not 
change the TS test requirements or surveillance frequency. 
Therefore, existing TS surveillance requirements are not reduced by 
the proposed change, thus no margins of safety are reduced.
    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: N.S. Reynolds, Esquire, Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817
    NRC Branch Chief: David Terao

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of amendment request: November 7, 2005
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs), to adopt NRC-approved Revision 4 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.'' The proposed amendment includes changes to the TS 
definition of Leakage, TS 3/4.4.6, ``Reactor Coolant System Leakage,'' 
TS 3/4.4.5, ``Steam Generators,'' and adds TS 6.19, ``Steam Generator 
(SG) Program,'' and TS 6.9.7, ``Steam Generator Tube Inspection 
Report.'' The proposed changes are necessary in order to implement the 
guidance for the industry initiative on Nuclear Energy Institute (NEI) 
97-06, ``Steam Generator Program Guidelines.''

[[Page 75492]]

    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 November 7, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change requires 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 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 SGTR event, a bounding primary to 
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits 
in the licensing basis plus the LEAKAGE rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as a 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 to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident.
    The performance criteria are only a part of the SG Program 
required by the proposed change to the TS. The program, defined by 
NEI 97-06, Steam Generator Program Guidelines, includes a framework 
that incorporates a balance of prevention, inspection, evaluation, 
repair, and leakage monitoring. The proposed changes do not, 
therefore, significantly increase the probability of an accident 
previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of 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 [kind] of accident from any accident 
previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in [a] 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 [a] SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    The NRC staff proposes to determine that the amendments request 
involves no significant hazards consideration.
    Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: October 21, 2005.
    Description of amendment request: The proposed amendment will 
revise the Technical Specifications to allow operation with a reduced 
reactor coolant system (RCS) flow rate of 300,000 gpm and a reduction 
in the maximum thermal power to 89 percent of the rated thermal power. 
The definition of rated thermal power remains unchanged at 2700 MWt. 
The flow rate of 300,000 gpm is expected to conservatively bound an 
analyzed steam generator tube plugging level of 42 percent per steam 
generator. The re-analysis performed to support this reduction in RCS 
flow used Westinghouse WCAP-9272-P-A methodology, the same methodology 
approved for St. Lucie Unit 2 in License Amendment 138.

[[Page 75493]]

    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 facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    None of the proposed changes to the Technical Specifications 
results in operation of the facility that adversely affects the 
initiation of any accident previously evaluated. There is no adverse 
impact on any plant system. Plant systems will continue to function 
as designed, and all performance requirements for these systems 
remain acceptable. The analysis, performed to support the proposed 
changes, has included evaluations and/or analyses of all the 
analyzed accident analyses, including the effects of changes on the 
SG tube sleeve design. The analyses and evaluations have verified 
that the accident analyses acceptance criteria continue to be met. 
Dose consequences acceptance criteria have been verified to be met 
for analyzed events. Therefore, the proposed changes do not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    No new accident scenarios, failure mechanisms or limiting single 
failures are introduced as a result of the proposed changes to the 
Technical Specifications. Although the allowable tube plugging level 
is increased, the criteria for tube plugging/sleeving and the tube 
integrity considerations remain unchanged. The proposed changes have 
no adverse effects on any safety-related systems and do not 
challenge the performance or integrity of any safety-related system. 
The DNBR [Departure from Nucleate Boiling Ratio] limits and trip 
setpoints associated with the respective reactor protection system 
functions have verified that the accident analyses criteria continue 
to be met. Therefore, this amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The safety analyses of all analyzed design basis accidents, 
supporting the proposed changes to the Technical Specifications, 
continue to meet the applicable acceptance criteria with respect to 
the radiological consequences, specified acceptable fuel design 
limits (SAFDLs), primary and secondary overpressurization, and 10 
CFR 50.46 requirements. The DNBR and the setpoint analyses are 
performed on a cycle-specific basis to verify that the reactor 
protection system functions continue to provide adequate protection 
against fuel design limits. Evaluation of the steam line break and 
LOCA [Loss of Coolant Accident] mass and energy releases determined 
that the overall containment response remains acceptable. The 
performance requirements for all systems have been verified to be 
acceptable from design basis accidents' consideration. The proposed 
amendment, therefore, will not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: July 25, 2005.
    Description of amendment request: The proposed amendment would add 
new Technical Specifications requirements to provide limiting 
conditions for operation (LCOs) and action statements and corresponding 
surveillance requirements for the Emergency Service Water (ESW) system. 
In the absence of such new requirement, the current requirement at 
Section 3.5.A.4 simply specifies that the unit be shutdown within 24 
hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Emergency Service Water (ESW) System is not an accident 
initiator. The proposed change provides operability requirements and 
surveillance requirements to ensure the ESW System is operable as 
required for accident mitigation. The proposed operability 
requirements and allowed outage time is consistent with the 
requirements for the systems supported by the ESW System. The 
[calculated radiological] dose to the public and the Control Room 
operators [due to a postulated accident] are unaffected by the 
proposed change. The proposed LCO provides direction with respect to 
actions to be taken when support systems are inoperable.
    The proposed Technical Specifications does not introduce new 
equipment operating modes, nor does the proposed change alter 
existing system relationships. The proposed amendment does not 
introduce new failure modes.
    Therefore, the proposed amendment will not significantly 
increase the probability or the 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 changes do not introduce new equipment operating 
modes, nor do they alter existing system relationships. The proposed 
changes do not introduce new failure modes. They do not alter the 
equipment required for accident mitigation and they appropriately 
consider the effects on supported systems when a support system is 
inoperable. When support systems are inoperable, actions are 
specified consistent with safe plant operation.
    Therefore, the proposed changes 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 
the margin of safety?
    Response: No.
    The proposed change provides specifications for the ESW System 
that are consistent with current Technical Specification 
requirements for other equipment. The proposed changes ensure that 
the ESW and other support systems will be available when required 
and provides adequate alternative actions when the support systems 
are not available. The allowed outage times for the ESW subsystem is 
consistent with that allowed for other equipment required for 
accident mitigation. Therefore, the proposed changes do not involve 
a significant reduction in the margin of safety.

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

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

    Date of amendment request: October 31, 2005.
    Description of amendment request: Omaha Public Power District (the 
licensee) has proposed to revise the Updated Safety Analysis Report 
(USAR) Safety Analysis, General, Section 14.1,

[[Page 75494]]

as well as the radiological consequences analyses for the events of 
Seized Rotor (SR), Section 14.6.2.8; Main Steam Line Break (MSLB), 
Section 14.12.6; Control Element Assembly Ejection (CEAE), Section 
14.13.4; and Steam Generator Tube Rupture (SGTR), Section 14.14.3. The 
USAR sections for radiological consequences of events need to be 
revised because of the planned replacement of the steam generators and 
pressurizer in the fall of 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed changes to the USAR discuss the changes to the 
Seized Rotor (SR), Control Element Assembly Ejection (CEAE), Steam 
Generator Tube Rupture (SGTR) and Main Steam Line Break (MSLB) 
events resulting from the installation of the replacement steam 
generators (RSGs) and the replacement pressurizer (RPZR). These 
changes do not affect an accident initiator previously evaluated in 
the USAR or the Technical Specifications and will not prevent any 
safety systems from performing their accident mitigating function as 
discussed in the USAR or the Technical Specifications.
    In all events evaluated, with the exception of the Control Room 
dose of the MSLB concurrent iodine spike case, there is no margin 
reduction. The Control Room dose of the MSLB concurrent iodine spike 
case is increased from 2.5 rem to 4.5 rem. The calculated doses 
resulting from the proposed changes to USAR Sections 14.1.6, 
14.6.2.8, 14.12.6, 14.13.4 and 14.14.3 remain below the regulatory 
limits set by 10 CFR 50.67.
    Therefore, these changes do not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are the result of changes in the analysis 
of the radiological consequences of the SR, CEAE, SGTR and MSLB 
events of the replacement of the steam generators (SGs) and the 
pressurizer. The proposed changes do not modify or install any 
safety related equipment. They do, however, change the licensing 
basis by using fuel gap fractions from Reference 7.6 in accordance 
with previously accepted license applications by other licensees and 
by assuming shorter concurrent iodine spike durations in accordance 
with Section 2.2 of Appendix E of RG 1.183, since the activity 
released during the eight-hour spike duration exceeds the available 
release.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The calculated doses resulting from the proposed changes to USAR 
Sections 14.1.6, 14.6.2.8, 14.12.6, 14.13.4 and 14.14.3 remain below 
the regulatory limits set by 10 CFR 50.67. In all events evaluated, 
with the exception of the Control Room dose of the MSLB concurrent 
iodine spike case, there is no margin reduction. The Control Room 
dose of the MSLB concurrent iodine spike case is increased from 2.5 
rem to 4.5 rem. This margin reduction is primarily due to the 
significant delay in the reactor coolant reaching 212 F with the 
RSGs and RPZR (i.e., at 159.2 hours versus the 10.94 hours 
applicable to the original steam generators). This analysis has 
conservatively used a spike duration of 4 hours. If the updated 
analysis took credit for the percentage of defective fuels 
associated with Technical Specification concentrations when 
developing the duration of the concurrent iodine spike (i.e., used 
0.28% defective fuel versus the conservatively assumed 1% defective 
fuel used in the analysis), the analysis would have resulted in an 
estimated spike duration of 2 hours instead of 4 hours and the 
control room dose would be significantly reduced.
    Therefore, the proposed changes do not involve a significant 
reduction in the safety margin.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: November 8, 2005.
    Description of amendment request: The proposed amendment would 
revise the Fort Calhoun Station (FCS) Technical Specifications (TS) to 
add a new Limiting Condition for Operation 2.8.3(6) and modify Table 3-
4, Table 3-5, and Design Features 4.3.1 to address criticality control 
during spent fuel cask loading operations in the spent fuel pool. This 
request applies only to spent fuel cask loading in the spent fuel pool 
and does not affect the licensing basis or invalidate our existing 
exemption from the criticality monitoring requirements of Title 10, 
Code of Federal Regulations (CFR) 70.24 for new and spent fuel storage.
    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.
    These proposed changes affect only operations in the spent fuel 
pool during spent fuel cask loading operations. Plant power 
operations and other spent fuel pool operations are not affected. 
There are no changes to the design or operation of the power plant 
that could affect system, component or accident functions resulting 
from these changes.
    Fuel loading into the spent fuel casks in the spent fuel pool 
will not require any significant changes to spent fuel pool 
structures, systems, or components, nor will their performance 
requirements be altered. The potential to handle a spent fuel cask 
was considered in the original design of the plant. Therefore, the 
response of the plant to previously analyzed Part 50 accidents and 
related radiological releases will not be adversely impacted, and 
will bound those postulated during cask loading activities in the 
cask loading area.
    Accordingly, the proposed changes do 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.
    These proposed changes affect only operations in the spent fuel 
pool during spent fuel cask loading operations. Plant power 
operations and other spent fuel pool operations are not affected. No 
new accident scenarios, failure mechanisms, or single failures are 
introduced as a result of the proposed changes. All systems, 
structures, and components previously required for mitigation of an 
event remain capable of fulfilling their intended design function 
with these changes to the TS.
    Fuel handling procedures and associated administrative controls 
for movement of spent fuel in the spent fuel pool remain applicable 
and are being appropriately augmented to accommodate spent fuel cask 
loading operations. Additionally, the soluble boron concentration 
required to maintain keff <=0.95 for postulated accidents 
associated with cask loading operations was also evaluated. The 
results of the analyses, using a methodology previously approved by 
the NRC, demonstrate that the amount of soluble boron assumed to be 
in the pool water during these postulated accidents (800 ppm [part 
per million]) is much less than the value at which the spent fuel 
pool is normally maintained (approximately 1900 ppm).
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.

[[Page 75495]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    An NRC-approved methodology was used to perform the criticality 
analyses that provide the basis to incorporate a boron concentration 
and a new burnup versus enrichment curve into the plant Technical 
Specifications to ensure criticality safety margins are maintained 
during spent fuel cask loading. Spent fuel casks at FCS are loaded 
in the spent fuel pool in an area adjacent to the spent fuel racks. 
No physical segregation such as a wall or gate exists between the 
spent fuel racks and spent fuel cask loading area. The cask loading 
area floor is approximately two feet lower than the floor on which 
the spent fuel racks are located. Therefore, the spent fuel pool 
water flows in and around the spent fuel racks and spent fuel casks 
being loaded in a common pool. Neutronic coupling between fuel in 
the spent fuel racks and fuel in the spent fuel cask has been 
appropriately considered in the criticality analysis, including 
accident events that postulate mis-loading of a fresh fuel assembly 
into the cask and dropping a fuel assembly between the spent fuel 
racks and spent fuel cask during loading.
    The normal condition criticality analysis was performed assuming 
no soluble boron in the spent fuel pool water and credit for fuel 
burnup. The proposed new Technical Specification requirement to 
permit only fuel assemblies with the minimum required burnup versus 
enrichment to be loaded into the spent fuel cask preserves this 
analysis basis. The accident condition criticality analysis was 
performed assuming a minimum of 800 ppm boron in the spent fuel pool 
during cask loading operations. All analyses account for 
uncertainties at a 95[-] percent probability/95-percent confidence 
level. The proposed new Technical Specification requirement to 
maintain a minimum boron concentration of 800 ppm in the spent fuel 
pool during spent fuel cask loading operations preserves this 
analysis basis. For defense-in-depth, the spent fuel pool boron 
concentration is typically maintained at approximately 1900 ppm 
during normal operations and would not be expected to be reduced 
during spent fuel cask loading operations.
    Therefore, there is no significant reduction in a margin of 
safety as a result of this change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Branch Chief: David Terao.

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

    Date of amendment requests: October 19, 2005.
    Description of amendment requests: 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 consistent with the program in place for 
complying with the requirements of 10 CFR 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.8 is added to the TS to provide this 
allowance and define the requirements and limitations for its use.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
November 24, 2004 (69 FR 68412), on possible amendments concerning 
TSTF-372, 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 4, 2005 (70 FR 23252). The 
licensee affirmed the applicability of the following NSHC determination 
in its application dated October 19, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change 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. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on 
allowance provided by proposed LCO 3.0.8 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.8. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change. The addition of a 
requirement to assess and manage the risk introduced by this change 
will further minimize possible concerns. Therefore, this change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Criterion 2--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.

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 Regulatory Guide 1.177. A bounding risk assessment 
was performed to justify the proposed TS changes. The proposed LCO 
3.0.8 defines limitations on the use of the provision and includes a 
requirement for the licensee to assess and manage the risk 
associated with operation with an inoperable snubber. The net change 
to the margin of safety is insignificant. Therefore, this change 
does not involve a significant reduction in a margin of safety.

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

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

    Date of amendment requests: October 19, 2005.
    Description of amendment requests: The proposed amendments would

[[Page 75496]]

update the Technical Specification (TS) 5.3, ``Unit Staff 
Qualifications,'' operator minimum qualification requirements contained 
in the March 28, 1980, NRC letter to all licensees with the more recent 
NRC-approved operator qualification requirements contained in American 
National Standards Institute/American Nuclear Society (ANSI/ANS) 3.1-
1993. In addition, the proposed changes remove the TS 5.3.1 plant staff 
retraining and replacement training program requirements which have 
been superseded by requirements contained in section 50.120 of Title 10 
of the Code of Federal Regulations (10 CFR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change is an administrative change to revise the 
Technical Specification (TS) 5.3.1 licensed operator minimum 
qualification requirements and remove the plant staff retraining and 
replacement training program requirements from the TS. The proposed 
change does not directly impact accidents previously evaluated. The 
Diablo Canyon Power Plant (DCPP) licensed operator training program 
is accredited by the National Academy for Nuclear Training (NANT) 
and is based on a systems approach to training consistent with the 
requirements of 10 CFR 55. Although licensed operator qualifications 
and training may have an indirect impact on accidents previously 
evaluated, the NRC considered this impact during the rulemaking 
process, and by promulgation of the revised 10 CFR 55 rule, 
concluded that this impact remains acceptable as long as the 
licensed operator training program is certified to be accredited and 
is based on a systems approach to training. The DCPP plant staff 
retraining and replacement training program meets the requirements 
of 10 CFR 50.120.
    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 accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in nature and does not 
affect the plant design, hardware, system operation, or operating 
procedures. The DCPP licensed operator training program is 
accredited by the NANT and is based on a systems approach to 
training consistent with the requirements of 10 CFR 55. Although 
licensed operator qualifications and training may have an indirect 
impact on accidents previously evaluated, the NRC considered this 
impact during the rulemaking process, and by promulgation of the 
revised 10 CFR 55 rule, concluded that this impact remains 
acceptable as long as the licensed operator training program is 
certified to be accredited and is based on a systems approach to 
training. The DCPP plant staff retraining and replacement training 
program meets the requirements of 10 CFR 50.120.
    Therefore, the proposed change does not create the possibility 
of a new or different 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 is administrative in nature and does not 
affect the plant design, hardware, system operation, or operating 
procedures. The change does not exceed or alter a design basis or 
safety limit and thus does not reduce the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: August 19, 2005.
    Description of amendment request: The amendment would relocate the 
Technical Specification response time testing tables to the Updated 
Final Safety Analysis Report.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment[s] relocate the instrument response time 
limits for the reactor trip system (RTS) and engineered safety 
feature actuation system (ESFAS) from the technical specifications 
to the Updated Final Safety Analysis Report (UFSAR). The proposed 
amendment[s] conform to the guidance given in Enclosures 1 and 2 of 
Generic Letter 93-08. Neither the response time limits nor the 
surveillance requirements for performing response time testing will 
be altered by this submittal. The overall RTS and ESFAS functional 
capabilities will not be changed and assurance that action 
requirements of the reactor trip and engineered safety features 
systems are completed within the time limits assumed in the accident 
analyses is unaffected by the proposed amendment[s].
    Therefore, operation of the facility in accordance with the 
proposed amendment[s] will not involve a significant increase in 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: No.
    The proposed amendment[s] will not change the physical plant or 
the modes of plant operation defined in the operating license[s]. 
The change does not involve the addition or modification of 
equipment nor does it alter the design or operation of plant 
systems.
    Therefore, operation of the facility in accordance with the 
proposed amendment[s] will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The measurement of instrumentation response times at the 
frequencies specified in the technical specification provides 
assurance that actions associated with the reactor trip and 
engineered safety features systems are accomplished within the time 
limits assumed in the accident analyses. The response time limits 
and the measurement frequencies remain unchanged by the proposed 
amendment[s].
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions.
    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.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Darrell J. Roberts.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of amendment request: November 15, 2005.
    Description of amendment request: The amendment would revise the 
Virgil

[[Page 75497]]

C. Summer Nuclear Station Technical Specifications (TS) 3/4.3.1, 
``Reactor Trip System Instrumentation,'' and TS 3/4.3.2, ``Engineered 
Safety Feature Actuation System Instrumentation,'' to implement the 
allowed outage time and bypass test time changes approved by the 
Nuclear Regulatory Commission in the Westinghouse topical report WCAP-
14333-P-A, Rev. 1, ``Probabilistic Risk Analysis of the Reactor Trip 
System and Engineered Safety Features Actuation System Test Times and 
Completion Times,'' dated October 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The same reactor trip system (RTS) 
and engineered safety feature actuation system (ESFAS) 
instrumentation will continue to be used. The protection systems 
will continue to function in a manner consistent with the plant 
design basis. These changes to the Technical Specifications do not 
result in a condition where the design, material, and construction 
standards that were applicable prior to the changes are altered.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event