[Federal Register Volume 69, Number 187 (Tuesday, September 28, 2004)]
[Notices]
[Pages 57978-57999]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-21345]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from September 3, 2004, through September 16, 
2004. The last biweekly notice was published on September 14, 2004 (69 
FR 55466).

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

[[Page 57979]]

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 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the basis 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

[[Page 57980]]

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, 
<a href="/cdn-cgi/l/email-protection#29616c687b60676e6d666a626c7d69677b6a076e667f"><span class="__cf_email__" data-cfemail="6028252132292e27242f232b2534202e32234e272f36">[email protected]</span></a>; 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 <a href="/cdn-cgi/l/email-protection#88c7cfcbc5e9e1e4cbede6fcedfac8e6faeba6efe7fe"><span class="__cf_email__" data-cfemail="5e11191d133f37321d3b302a3b2c1e302c3d70393128">[email protected]</span></a>. 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 O1F21, 11555 Rockville Pike (First Floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, (301) 415-4737 or by e-mail 
to <a href="/cdn-cgi/l/email-protection#b4c4d0c6f4dac6d79ad3dbc2"><span class="__cf_email__" data-cfemail="ee9e8a9cae809c8dc0898198">[email protected]</span></a>.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendments request: July 9, 2004.
    Description of amendments request: The amendments would revise the 
Technical Specifications (TSs) to allow operation of Palo Verde Nuclear 
Generating Station (PVNGS), Units 1 and 3 up to a maximum reactor core 
power level of 3990 Megawatts thermal (MWt), an increase of 2.94 
percent above the current licensed power level of 3876 MWt. The 
proposed amendments would also make administrative changes to the PVNGS 
Unit 2 TSs so that the changed pages would apply to the three PVNGS 
units. Operation at the uprated power level with replacement steam 
generators has been approved for PVNGS Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Do the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Response: No.
    (a) Evaluation of the Probability of Previously Evaluated 
Accidents
    Plant Structures, Systems and Components (SSCs) have been 
verified to be capable of performing their intended design functions 
at uprated power conditions. Where necessary, a small number of 
minor modifications will be made prior to implementation of uprated 
power operations so that surveillance test acceptance criteria 
continues to be met. The analysis has concluded that operation at 
uprated power conditions will not adversely affect the capability or 
reliability of plant equipment. Current technical specification (TS) 
surveillance requirements ensure frequent and adequate monitoring of 
system and component operability. All systems will continue to be 
operated within current operating requirements at uprated 
conditions. Therefore, no new structure, system or component 
interactions have been identified that could lead to an increase in 
the probability of any accident previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR).
    (b) Evaluation of the Consequences of Previously Evaluated 
Accidents
    The radiological consequences were reviewed for all design basis 
accidents (DBAs) (i.e., both LOCA [loss-of-coolant accident] and 
non-LOCA accidents) previously analyzed in the UFSAR. The analysis 
showed that the resultant radiological consequences for both LOCA 
and non-LOCA accidents remain either unchanged or have not 
significantly increased due to operation at uprated power 
conditions. The radiological consequences of all DBAs continue to 
meet established regulatory limits.
    (2) Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Response: No.
    The configuration, operation and accident response of the PVNGS 
[Palo Verde Nuclear Generating Station] Units I and 3 structures, 
systems, and components are unchanged by operation at uprated power 
conditions or by the associated proposed TS changes. Analyses of 
transient events have confirmed that no transient event results in a 
new sequence of events that could lead to a new accident or 
different scenario.
    The effect of operation at uprated power conditions on plant 
equipment has been evaluated. No new operating mode, safety-related 
equipment lineup, accident scenario, or equipment failure mode was 
identified as a result of operating at uprated conditions. In 
addition, operation at uprated power conditions does not create any 
new failure modes that could lead to a different kind of accident. 
Minor plant modifications, to support Implementation of uprated 
power conditions, will be made as required to existing SSCs. The 
basic design function of all SSCs remains unchanged and no new 
equipment or systems have been installed that could potentially 
introduce new failure modes or accident sequences.
    Based on this analysis, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed changes. The proposed changes 
do not have an adverse effect on any safety-related system or design 
basis function. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed changes involve a significant reduction in 
a margin of safety?
    Response: No.
    A comprehensive analysis was performed to evaluate the effects 
of power uprate on PVNGS Units 1 and 3. This analysis identified and 
defined the major input parameters to the NSSS [nuclear steam supply 
system], reviewed NSSS design transients, and reviewed the 
capabilities of the NSSS and BOP [balance of plant] fluid systems, 
NSSS/BOP interfaces, NSSS and BOP control systems, and NSSS and BOP 
SSCs. All appropriate NSSS accident analyses were re-performed to 
confirm that acceptable results were maintained and that the 
radiological consequences remained within regulatory and Standard 
Review Plan (SRP) limits. The nuclear and thermal hydraulic 
performance of nuclear fuel was also reviewed to confirm acceptable 
results. The analyses confirmed that all NSSS and BOP SSCs are 
capable, some with minor modifications, to safely support operations 
at uprated power conditions.
    The margin of safety of the reactor coolant pressure boundary is 
maintained under uprated power conditions. The design pressure of 
the reactor pressure vessel and reactor coolant system will not be 
challenged as the pressure mitigating systems were confirmed to be 
sufficiently sized to adequately control pressure under uprated 
power conditions.
    Reanalysis of containment structural integrity under Design 
Basis Accident (DBA) conditions indicates that the calculated peak

[[Page 57981]]

containment pressure (Pa) increases from 52.0 psig [pounds per 
square inch gauge] to 58.0 psig, but remains less than the 
containment internal design pressure of 60 psig. The proposed value 
for Pa has been rounded up from the actual calculated value of 57.85 
psig.
    Radiological consequences of the following accidents were 
reviewed: Main Steam Line Break, Locked Reactor Coolant Pump (RCP) 
Rotor, CEA Ejection, Small Steam Line Break Outside Containment, 
Steam Generator Tube Rupture, LBLOCA, SBLOCA, Waste Gas Decay Tank 
Rupture, Liquid Waste Tank Failure, and Fuel Handling Accident. The 
resultant radiological consequences for each of these accidents did 
not show a significant change due to uprated power conditions and 10 
CFR 100 and SRP limits continue to be met.
    The analyses supporting operation at power uprate conditions 
have demonstrated that all systems and components are capable of 
safely operating at uprated power conditions. All design basis 
accident acceptance criteria will continue to be met. Therefore, it 
is concluded that 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 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Section Chief: Robert Gramm.

Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: August 19, 2004.
    Description of amendment request: The proposed amendment would 
revise the reactor coolant system (RCS) pressure and temperature limits 
by replacing Technical Specification Section 3.4.3, ``RCS Pressure and 
Temperature (P/T) Limits,'' Figures 3.4.3-1 and 3.4.3-2, with figures 
that are applicable up to 35 effective full-power years (EFPY).
    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?
    The proposed RCS P/T limits are based on NRC-approved 
methodology and will continue to maintain appropriate limits for the 
HBRSEP [H.B. Robinson Steam Electric Plant], Unit No. 2, RCS up to 
35 EFPY. These changes provide appropriate limits for pressure and 
temperature during heatup and cooldown of the RCS, thus ensuring 
that the probability of RCS failure is maintained acceptably low. 
These limits are not directly related to the consequences of 
accidents.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    The proposed changes will continue to ensure that the RCS will 
be maintained within appropriate pressure and temperature limits 
during heatup and cooldown. No physical changes to the HBRSEP, Unit 
No. 2, systems, structures, or components are being implemented. 
There are no new or different accident initiators or sequences being 
created by the proposed Technical Specifications changes. Therefore, 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed changes ensure that the margin of safety for the 
fission product barriers protected by these functions will continue 
to be maintained. This conclusion is based on use of the applicable 
NRC-approved methodology for developing and establishing the 
proposed RCS P/T limits. Therefore, these changes do not involve a 
significant reduction in the margin of safety.
    Based on the preceding discussion, the requested change does not 
involve a significant hazards consideration.

    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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael Marshall (Acting).

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of amendment request: August 11, 2004.
    Description of amendment requests: The Haddam Neck Plant (HNP) is 
currently undergoing active decommissioning. The proposed amendment 
would revise Technical Specifications (TS) to reflect removal of all 
Spent Nuclear Fuel (SNF) from the HNP spent fuel pool, and delete the 
requirement for submittal of an annual Occupational Radiation Exposure 
Report consistent with Industry's Technical Specifications Task Force 
(TSTF)-369, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with 10 CFR 50.92, CYAPCO has reviewed the 
proposed changes and concluded that the proposed changes do not 
involve a Significant Hazard Consideration (SHC). The following is 
provided in support of this conclusion:
    Incorporation of TSTF-369, Revision 1: CYAPCO has reviewed the 
no significant hazards consideration determination published in the 
Federal Register (69 FR 35067) as part of the CLIIP. CYAPCO has 
concluded that the determination presented in the Federal Register 
is applicable to the HNP and is hereby incorporated by reference to 
satisfy the requirements of 10 CFR 50.91.
    Deletion and Relocation of Technical Specifications: The 
proposed changes do not involve an SHC because the changes would 
not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes (deletion of operational requirements and 
certain design requirements) reflect the complete transfer of the 
spent fuel from the spent fuel pool to the Independent Spent Fuel 
Storage Installation (ISFSI). Design basis accidents related to the 
spent fuel pool are discussed in the Haddam Neck Plant (HNP) Updated 
Final Safety Analysis (UFSAR) Chapter 15. These postulated accidents 
are predicated on spent fuel being stored in the spent fuel pool. 
With the removal of the spent fuel from the spent fuel pool, there 
are no remaining safety related Structures, Systems, and Components 
(SSCs) to be monitored and there are no credible accidents that 
require the actions of a Certified Fuel Handler or an Equipment 
Operator to prevent occurrence or mitigate the consequences of an 
accident.
    In addition, the HNP UFSAR Chapter 15 also provides a discussion 
of other radiological events postulated to occur as a result of 
decommissioning with the bounding consequences resulting from a fire 
in a resin container. The proposed changes do not have an adverse 
impact on decommissioning activities or any of their postulated 
consequences.
    The proposed changes related to the relocation of certain 
administrative requirements do not affect operating procedures or 
administrative controls that have the function of preventing or 
mitigating any design basis accidents. In addition, these proposed 
changes are consistent with the guidance of NRC Administrative 
Letter 95-06.
    Therefore, the proposed changes do not involve a significant 
increase in the

[[Page 57982]]

probability or consequences of any accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes eliminate the operational requirements and 
certain design requirements associated with the storage of the spent 
fuel in the spent fuel pool, and relocate certain administrative 
controls to the Connecticut Yankee Quality Assurance Program 
(CYQAP). With the complete removal of the spent fuel from the spent 
fuel pool, there are no safety related SSCs that remain at the 
plant. Thus the proposed changes will not have any effect on the 
operation or design function of safety related SSCs. The proposed 
changes do not introduce any new failure modes. Therefore, the 
proposed changes will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The design basis and accident assumptions within the HNP UFSAR 
and the Technical Specifications relating to spent fuel are no 
longer applicable. The proposed changes do not affect remaining 
plant operations, systems, or components supporting decommissioning 
activities. In addition, the proposed changes do not result in a 
change in initial conditions, system response time, or in any other 
parameter affecting the course of a decommissioning activity 
accident analysis. 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.

    NRC Section Chief: Claudia Craig.

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

    Date of amendment request: May 27, 2004.
    Description of amendment request: The proposed amendment would 
delete the requirements from the Technical Specifications (TS) to 
maintain hydrogen recombiners and hydrogen monitors. A notice of 
availability for the TS improvement using the consolidated line item 
improvement process was published in the Federal Register on September 
25, 2003 (68 FR 554416). Licensees were generally required to implement 
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile 
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
combustible gas control were imposed by Order for many facilities and 
were added to or included in the TSs for nuclear power reactors 
currently licensed to operate.
    The revised 10 CFR 50.44, ``Standards for Combustible Gas Control 
System in Light-Water-Cooled Power Reactors,'' eliminated the 
requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated May 27, 2004. 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44, the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TSs, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TSs, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to

[[Page 57983]]

approximately 24 hours after the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TSs 
will not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Mary Jane Ross-Lee (Acting).

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

    Date of amendment request: August 20, 2004.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.3.8, ``Post Accident Monitoring 
[PAM] Instrumentation,'' to eliminate TS requirements associated with 
the reactor building spray (RBS) flow instruments commensurate with the 
importance of their revised post-accident function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--The proposed amendment would not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated

    Duke proposes to remove the RBS flow instrument from Technical 
Specification Table 3.3.8-1 based on a change in its purpose due to 
recent modifications completed at Oconee. The TS 3.3.8 requirement 
to declare the affect [affected] RBS System train inoperable is 
conservative (and inappropriate) when the associated RBS flow 
instrument is inoperable. Due to recent plant modifications, the RBS 
flow instruments are no longer needed to allow the operator to 
throttle flow to preclude RBS pump runout post accident. The revised 
post accident function of this PAM instrument is to provide 
information to indicate the operation of the RBS System. There are 
alternate means to verify that the RBS is in operation, such as, 
verifying the RBS pump and valve status. The failure of an RBS flow 
instrument has no impact on the probability of an accident analyzed 
in the UFSAR [Updated Final Safety Analysis Report]. The RBS flow 
instrument is no longer needed to mitigate the consequences of an 
accident analyzed in the UFSAR. As such, the proposed LAR [license 
amendment request] does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

Criterion 2--The proposed amendment would not create the possibility of 
a new or different kind of accident from any kind of accident 
previously evaluated

    Duke proposes to remove the RBS flow instrument from Technical 
Specification Table 3.3.8-1 based on a change in its purpose due to 
recent modifications completed at Oconee. The TS 3.3.8 requirement 
to declare the affect [affected] RBS System train inoperable is 
conservative (and inappropriate) when the associated RBS flow 
instrument is inoperable. Due to recent plant modifications, the RBS 
flow instruments are no longer needed to allow the operator to 
throttle flow to preclude RBS pump runout post accident. These 
changes do not alter the nature of events postulated in the Safety 
Analysis Report nor do they introduce any unique precursor 
mechanisms. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

Criterion 3--The proposed amendment would not involve a significant 
reduction in a margin of safety

    The proposed TS changes do not unfavorably affect any plant 
safety limits, set points, or design parameters. The changes also do 
not unfavorably affect the fuel, fuel cladding, RCS [reactor coolant 
system], or containment integrity. Therefore, the proposed TS 
change, which changes TS requirements associated with revised PAM 
function of the RBS flow instrument channels, does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

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

    Date of amendment request: August 26, 2004.
    Description of amendment request: The proposed amendments would add 
new Technical Specification (TS) 3.3.29 and TS Bases 3.3.29, ``Reactor 
Building Auxiliary Cooler (RBAC) Isolation Circuitry,'' to accommodate 
new circuitry that isolates non-safety portions of the low pressure 
service water (LPSW) system piping inside containment that supply the 
RBACs. This isolation eliminates potentially damaging water hammers 
that could occur in the event of certain design-bases events or 
transients.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
Criterion 1--The Proposed Amendment Would Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The requested license amendment would add a new Technical 
Specification to provide appropriate controls for the Reactor Building 
(RB) Auxiliary Cooler (RBAC) isolation circuitry that is being added to 
the design of the three Oconee units. The RBAC isolation circuitry 
provides an automatic means to isolate the LPSW flow stream to the 
RBACs on a loss of LPSW flow that can lead to a column closure water 
hammer inside the RB when LPSW flow is restarted. The new circuitry 
ensures that significant waterhammers do not occur in the LPSW piping 
to the RBACs and other RB components. The new circuitry will eliminate 
an Operable but degraded/non-conforming condition associated with 
potentially damaging waterhammers.
    The proposed RBAC isolation circuitry Technical Specification will 
provide means to assure that the RBAC isolation circuitry operates at a 
performance level necessary to provide for safe operation of the LPSW 
system following installation of the LPSW modification and RBAC 
isolation circuitry at each of the three units. The addition of the 
RBAC isolation circuitry Technical Specification does not increase the 
probability or consequences of any accident previously evaluated.
Criterion 2--The Proposed Amendment Would Not Create the Possibility of 
a New or Different Kind of Accident From Any Accident Previously 
Evaluated
    The proposed RBAC isolation circuitry Technical Specification 
provides a means to assure the isolation circuitry operates at a 
performance level necessary to provide for safe operation

[[Page 57984]]

of the modified LPSW system flow to the RBACs. The change enhances the 
plant design by eliminating the possibility of significant waterhammers 
that could occur inside the RB on a loss of LPSW flow to the RBACs.
    The proposed Technical Specification will not create the 
possibility of a new or different kind of accident from any kind of 
accident previously evaluated.
Criterion 3--The Proposed Amendment Would Not Involve a Significant 
Reduction in a Margin of Safety.

    The proposed change does not adversely affect any plant safety 
limits, set points, or design parameters. The change also does not 
adversely affect the fuel, fuel cladding, Reactor Coolant System, or 
containment integrity. The RBACs will continue to be isolated during 
ES events. The modification eliminates significant waterhammers in 
the LPSW piping to the RBACs.
    The change will enhance the ability to provide LPSW flow to 
safety related loads following LOOP events. Therefore, the proposed 
change does not involve a 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: Anne W. Cottingham, Winston and Strawn LPP, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: June 24, 2004.
    Description of amendment request: The proposed amendment would 
allow entry into a mode or other specified condition in the 
applicability of a Technical Specification (TS), while in a condition 
statement and the associated required actions of the TSs, provided the 
licensee performs a risk assessment and manages risk consistent with 
the program in place for complying with the requirements of Title 10 of 
the Code of Federal Regulations (10 CFR), Part 50, Section 50.65(a)(4). 
Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual 
TSs would be eliminated, several notes or specific exceptions would be 
revised to reflect the related changes to LCO 3.0.4, and Surveillance 
Requirement (SR) 3.0.4 would be revised to reflect the LCO 3.0.4 
allowance.
    This change was proposed by the industry's TS Task Force (TSTF) and 
is designated TSTF-359. The NRC staff issued a notice of opportunity 
for comment in the Federal Register on August 2, 2002 (67 FR 50475), on 
possible amendments concerning TSTF-359, 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 April 4, 2003 (68 FR 16579). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated June 24, 2004.
    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 entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. 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). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, 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 a Margin of Safety

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing 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 proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc. 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: September 1, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation 
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 1, 2004. 
Basis for proposed no significant hazards consideration determination: 
As required by 10 CFR 50.91(a), an analysis of the issue of no 
significant

[[Page 57985]]

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 eliminates the TS reporting requirements to 
provide a monthly operating report of shutdown experience and 
operating statistics if the equivalent data is submitted using an 
industry electronic database. It also eliminates the Technical 
Specification reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC regulations. The proposed change involves no 
changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 2--The proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?

    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

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

    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: June 24, 2004.
    Description of amendment request: The proposed amendment would 
incorporate several Technical Specification Task Force (TSTF) changes 
to the licensees Technical Specifications (TSs). The specific TSTF 
changes that would be incorporated are:
    (1) TSTF-5, Rev. 1, Delete Safety Limit Violation Notification 
Requirement--This change modifies TS Section 2.2 to remove the 
requirements to report safety limit violations. Associated references 
to Title 10 of the Code of Federal Regulations (10 CFR), Sections 50.72 
and 50.73, are also removed.
    TSTF-208, Rev. 0, Extension of Time to Reach Mode 2 in LCO 
(Limiting Condition for Operation) 3.0.3--This TSTF modifies TS Section 
LCO 3.0.3 to revise the time to be in Mode 2 once LCO 3.0.3 is entered 
from 7 hours to a bracketed site-specific time depending on the 
individual plant's ability to reach Mode 2 in a controlled shutdown.
    TSTF-222, Rev. 1, Control Rod Scram Time Testing and TSTF-229, Rev. 
0, Revise Surveillance Requirement 3.2.2.2 for Consistency with 
3.1.4.4--This TSTF modifies the TSs to clarify the frequency of 
performing control rod scram time testing subsequent to performance of 
an outage that involved the movement of fuel. The current wording of 
Surveillance Requirement (SR) 3.1.4.1 could be interpreted that all 
control rods need to be scram time tested even if the shutdown was for 
a brief amount of time and only a limited amount of fuel was moved in 
the reactor (e.g., if only one bundle is moved in a mid-cycle fuel 
replacement). This change clarifies the intent of the TSs.
    TSTF-297, Rev. 1, and TSTF-227, Rev. 0--These two TSTFs affect the 
following three TS Sections:

3.3.2.2--Feedwater and Main Turbine High Water Level Trip 
Instrumentation
3.3.4.1--Anticipated Transient Without Scram Recirculation Pump Trip 
(ATWS-RPT) Instrumentation
3.3.4.2--End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation

    TSTF-297, Rev. 1--This TSTF modifies the TSs to add a new Required 
Action and corresponding note to allow affected feedwater pump(s) and 
main turbine valve(s) to be removed from service. This change is 
necessary to allow components to be removed from service to fulfill the 
safety function without a reduction in power to less than 25% rated 
thermal power. A similar note is added to TS Sections 3.3.4.1 and 
3.3.4.2 to provide the same clarification for when the associated 
Required Action is the appropriate action.
    TSTF-227, Rev. 0--This TSTF modifies the TSs to eliminate ambiguity 
in the EOC-RPT Instrumentation Condition A. Since the LCO allows for 
having EOC-RPT instrumentation OPERABLE or certain fuel thermal limits 
are met, Condition A was inappropriately worded. The wording of 
Condition A is revised to add the word `required' if one or more 
channels are inoperable. Without the word `required', one could 
interpret Condition A as needing entry even if the fuel thermal limits 
were being applied instead of applying the operability requirements to 
the EOC-RPT instrumentation.
    TSTF-295, Rev. 0, Post-Accident Monitoring Clarifications--This 
TSTF modifies the TSs to clarify that a separate Condition entry is 
allowed for each penetration flow path for the Post Accident Monitoring 
(PAM) instrumentation Primary Containment Isolation Valve (PCIV) 
indication function.
    TSTF-275, Rev. 0, ECCS Instrumentation Clarifications--This TSTF 
modifies the TSs to clarify which Emergency Core Cooling System (ECCS) 
instrumentation is required to be OPERABLE to support Emergency Diesel 
Generator (EDG) operability. Footnote (a) to Table 3.3.5.1-1 has been 
changed to only require the affected functions to be OPERABLE in Modes 
4 and 5 when the associated ECCS is required to be OPERABLE per LCO 
3.5.2.
    TSTF-306, Rev. 2, Traversing In-Core Probe Instrumentation 
Specification Requirements--This TSTF modifies the TSs by adding a note 
that penetration flow path may not be isolated intermittently under 
administrative control to conform to what is already allowed for 
similar specifications for Primary Containment Isolation Valves 
(PCIVs). Also, the Traversing In-core Probe (TIP) system isolation is 
set apart as a separate function including the allowance of isolating 
the penetration instead of requiring a plant shutdown.
    TSTF-416, Rev. 0, Clarification of LPCI Operability during Decay 
Heat Removal Operations--This TSTF modifies the TSs by moving the note 
that modifies Low Pressure Coolant Injection (LPCI) surveillances to 
the LCO in LCO 3.5.1 and LCO 3.5.2. These notes provide clarity that 
the LPCI may be considered OPERABLE during alignment and operation in 
the decay heat removal Mode.
    TSTF-17, Rev. 2, Containment Airlock Testing Frequency--This TSTF 
modifies the TSs to extend the testing frequency of the containment 
interlock mechanism from 184 days to 24 months. Also, the corresponding 
note for this surveillance is no longer required due to the longer 
surveillance frequency.

[[Page 57986]]

    TSTF-30, Rev. 3, TSTF-323, Rev. 0, TSTF-45, Rev. 2, TSTF-46, Rev. 
1, and TSTF-269, Rev. 2, Containment Isolation Valve Specification 
Changes--These TSTFs modify TS Sections 3.6.1.3 concerning Primary 
Containment Isolation Valves (PCIVs) and 3.6.4.2 concerning Secondary 
Containment Isolation Valves (SCIVs).
    TSTF-30, Rev. 3 & TSTF-323, Rev. 0--These TSTFs revise TS 3.6.1.3 
to allow for a 72-hour completion time for a closed system flow path 
with an inoperable isolation valve and allow for a 72-hour completion 
time for a penetration flow path with an inoperable Excess Flow Check 
Valve (EFCV).
    TSTF-45, Rev. 2--This TSTF revises TSs 3.6.1.3 and 3.6.4.2 to 
revise surveillance requirements for valve line-ups. Specifically, if a 
containment isolation valve is locked, sealed, or otherwise secured, 
they are not required to be verified to be closed during the 
performance of the surveillance test.
    TSTF-46, Rev. 1--This TSTF revises containment isolation valve 
surveillances to delete the reference to verifying the isolation time 
of `each power operated' containment isolation valve and only require 
verification of each `automatic isolation valve'.
    TSTF-269, Rev. 2--This TSTF allows for verification of valve status 
by administrative means for repetitive verification of locked, sealed, 
or secured valves.
    TSTF-322, Rev. 2, Secondary Containment Operability Clarification--
This TSTF modifies the TSs to clarify the intent of the secondary 
containment boundary integrity. Associated surveillances currently 
imply that secondary containment would be inoperable if a Standby Gas 
Treatment (SGT) subsystem was inoperable.
    TSTF-276, Rev. 2, Power Factor for Emergency Diesel Generator (EDG) 
Surveillances--This TSTF modifies the TSs to allow for certain EDG 
testing to be performed even if the specified power factor cannot be 
achieved.
    TSTF-65, Rev. 1, Generic Organization Titles--This TSTF modifies 
the TSs to allow the use of generic organizational titles in place of 
plant-specific titles. Therefore, for the TSs, a change is requested to 
replace plant-specific titles with generic titles.
    TSTF-299, Rev. 0, Primary Coolant Sources Inspection Requirements--
This TSTF modifies the TSs Section 5.2.2, `Primary Coolant Sources 
Outside Containment' to clarify the intent of refueling cycle intervals 
with respect to the system leak test requirements and adds a sentence 
that the leak test is subject to the provisions of Surveillance 
Requirements (SR) 3.0.2.
    TSTF-279, Rev. 0, Inservice Testing Program Clarifications--This 
TSTF modifies TSs Section 5.5.8, ``Inservice Testing Program,'' to 
delete the reference to `applicable supports' as part of the 
description for the Inservice Testing Program. The applicable TS 
Section is 5.5.6.
    TSTF-118, Rev. 0, Diesel Generator Fuel Oil Testing Program 
Clarifications--This TSTF modifies TSs Section 5.5.13, ``Diesel Fuel 
Oil Testing Program,'' to allow for the provisions of SR 3.0.2 (25% 
extension) and SR 3.0.3 (missed surveillance actions) to apply to 
surveillances. The applicable TS Section is 5.5.9.
    TSTF-106, Rev. 1, Diesel Generator Fuel Oil Testing Program 
Clarifications--This TSTF modifies the TSs to clarify that Section 
5.5.10.b, concerning verification of the diesel fuel oil that was 
sampled meets the required ASTM properties, only applies to new fuel. 
As written, it could be interpreted that this testing is required for 
existing fuel that is routinely sampled. The applicable TS Section is 
5.5.9.b.
    TSTF-152, Rev. 0, Routine Reporting Requirements Upgrade--This TSTF 
modifies the TSs to revise the Occupational Radiation Exposure Report 
and the Radioactive Effluent Release Report requirements to be 
consistent with other regulatory changes that have occurred.
    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:

    A. TSTF-5, Rev. 1, Delete Safety Limit Violation Notification 
Requirements.
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This action does not affect the plant or operation of the plant. 
The change simply removes duplicative information from the Technical 
Specifications that is covered in the NRC regulations. 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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system. 
This change is considered an administrative action to remove 
duplicative reporting requirements.
    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.
    This administrative action does not involve any reduction in a 
margin of safety. Removal of duplicative information does not affect 
compliance with the regulations. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.
    B. TSTF-208, Rev. 0, Extension of Time to Reach Mode 2 in LCO 
3.0.3.
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The time frame to take response action in accordance with LCO 
3.0.3 is not an initiating condition for any accident previously 
evaluated and the accident analyses do not assume that any equipment 
is out of service such that LCO 3.0.3 is entered. The small increase 
in the time allowed to reach Mode 2 would not place the plant in any 
significantly increased probability of an accident occurring. The 
plant would already be proceeding to a plant shutdown condition 
because of the 1 hour requirement to initiate shutdown actions. 
There is no change in the time period to reach Mode 3. The Mode 3 
Condition is the point where the plant is shutdown. Therefore, since 
there is no change to the 1 hour requirement to initiate the 
shutdown nor any change to the time period to reach the shutdown 
Condition, the small change in the time to reach the Mode 2 status 
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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
There are no plant physical alterations proposed. The proposed 
changes have no adverse effects on any safety-related system or 
component and do not challenge the performance or integrity of any 
safety related system. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The time period to reach Mode 3 and Mode 4 are unaffected by 
this activity. This change simply provides a plant specific value 
for reaching Mode 2 if LCO 3.0.3 is entered

[[Page 57987]]

which is within the intent of LCO 3.0.3 for performing a controlled 
plant shutdown. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    C. TSTF-222. Rev. 1, Control, Red Scram Time Testing, and TSTF-
229, Rev. 0, Revise Surveillance Requirement 3.2.2.2 for Consistency 
with 3.1.4.4
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    These changes are considered clarifications to the original 
intent of the Technical Specifications. Adequate testing of control 
rods is ensured by this change. Control rod operability is not 
affected by these changes. 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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety-related system. 
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.
    This change is administrative in nature and does not affect any 
safety analyses assumptions. Adequate control rod testing continues 
to be maintained with implementation of this activity. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.
    (D) TSTF 297, Rev. 1, and TSTF 227, Rev. 0, Enhancements to 
Feedwater/Main Turbine High Water Level Trip, EOC-RPT, and ATWS RPT 
Specifications
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    There are no changes to the plant configuration assumed for any 
accident. The removal from service of equipment that results in its 
safety function being met can not adversely affect the consequences 
of accidents previously evaluated. Other changes are administrative 
clarifications that have no affect on accidents. 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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety-related system. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The actions involved with this activity ensure that safety 
functions are met. There are no changes in the overall requirements 
of having trip instrumentation available for event mitigation. There 
are no affects on the plant safety analyses. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.
    (E) STF-295, Rev. 0, Post-Accident Monitoring Clarications
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The equipment involved with the revised Technical Specifications 
are for post-accident monitoring. This equipment has no possibility 
of increasing the probability of occurrence of the accident since it 
is monitoring equipment only. The consequences of an accident are 
not affected since this change maintains the original intent of the 
Technical Specifications in having available monitoring information 
for each PCIV penetration path. 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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The Technical Specifications continue to require appropriate 
post accident monitoring equipment to be OPERABLE. Adequate 
instrumentation for post-accident monitoring will be ensured by the 
Technical Specification requirements. There are no changes to the 
plant safety analyses involved with this change. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    (F) TSTF-275, Rev. 0, ECCS Instrumentation Clarifications
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The equipment involved is for mitigative purposes and will not 
affect the probability of occurrence of an accident. Technical 
Specifications ensures that adequate mitigative equipment continues 
to be OPERABLE for any event that may occur in Modes 4 and 5. This 
change is considered an upgrade to the specifications that will 
provide more consistency within the Technical Specifications. There 
are no changes to requirements that ensure appropriate Emergency 
Core Cooling Systems are OPERABLE. 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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety-related system.
    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 is no impact on mitigative equipment that is required to 
respond to events while in Modes 4 and 5. There is no impact on the 
plant safety analyses. This change is considered as an upgrade to 
Technical Specifications that will improve consistency within the 
Technical Specifications. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    (G) TSTF-306, Rev. 2, Traversing In-Core Probe Instrumentation 
Specifications Requirements
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The addition of a note that the penetration flow path may be un-
isolated under administrative control simply provides

[[Page 57988]]

consistency with what is already allowed elsewhere in [the] 
Technical Specifications. The isolation function of the TIP valves 
are mitigative equipment. They do not create any increased 
possibility of an accident since they are mitigative. Also, the 
operation of the manual shear valves is unaffected by this activity. 
The ability to manually isolate the TIP system by either the normal 
isolation valve or the shear valve would be unaffected by the 
inoperable instrumentation. Therefore, the same action as for manual 
isolation Functions provides an appropriate level of safety. 
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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The addition of a note that the penetration flow path may be un-
isolated under administrative control simply provides consistency 
with what is already allowed elsewhere in Technical Specifications. 
The ability to manually isolate the TIP system by either the normal 
isolation valve or the shear valve would be unaffected by the 
inoperable instrumentation. Therefore, the same action as for manual 
isolation Functions provides an appropriate level of safety. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    (H) TSTF-416, Rev. 0 Clarification of LPCI Operability during 
Decay Heat Removal Operations
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change makes the Technical Specifications and their 
Bases consistent in their consideration of an LPCI subsystem aligned 
for decay heat removal being considered OPERABLE for ECCS. The LCO 
3.5.1 and LCO 3.5.2 Bases state that a LPCI subsystem may be 
considered OPERABLE during alignment and operation for decay heat 
removal. As a result, no initiators to accidents previously 
evaluated are affected and no mitigating equipment assumed in the 
accidents previously evaluated are affected since the allowance for 
LPCI being considered operable during these type of shutdown cooling 
alignments or operations was the intent of the current technical 
Specifications. Consequently, the probability or consequences of an 
accident previous evaluated is not significantly increased.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed change makes the Technical Specifications and their 
Bases consistent in their consideration of an LPCI subsystem aligned 
for decay heat removal being considered OPERABLE for ECCS. The LCO 
3.5.1 and LCO 3.5.2 Bases state that an LPCI subsystem may be 
considered OPERABLE during alignment and operation for decay heat 
removal. As the operability requirements of the LPCI subsystem are 
unaffected, the margin of safety is unaffected. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    (I) STF-17, Rev. 2, Containment Airlock Testing Frequency
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The containment airlock is considered as mitigative equipment. 
Therefore, there are no impacts on the probability of accidents. The 
proposed surveillance frequency assures that the interlock is 
working such that there is no unintentional opening of both airlock 
doors when containment is required. Because the interlock is assured 
to be working, there will be no significant increase in the 
consequences of an accident. There is no degradation in the ability 
of the interlock to assure the containment integrity function is 
maintained. 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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The frequency of 24 months for the interlock testing has been 
demonstrated to be adequate with regards to the reliability of the 
airlock. There is no impact on the leak testing requirements. There 
is no affect on the plant safety analyses. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.
    (J) TSTF-30, Rev. 3, TSTF-323, Rev. 0, TSTF-45, Rev. 2, TSTF-46, 
Rev. 1, and TSTF-269, Rev. 2, Containment Isolation on Valve 
Specification Changes
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The equipment affected by these changes is for mitigative 
purposes. Therefore, there cannot be an increase in the probability 
of occurrence of an accident. The controls required in the Technical 
Specifications are adequate to ensure that the containment barriers 
are ensured. Isolation valves will be assured to be in their correct 
positions. Also, inoperable isolation valves in closed systems and 
inoperable EFCVs have been evaluated to not have any significant 
impact to the consequences of an accident due to the closed system 
providing a barrier for the inoperable closed system isolation valve 
and bounding analyses have been performed for EFCV instrument line 
failures. 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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The equipment affected by these changes is for mitigative 
purposes. The controls

[[Page 57989]]

required in the Technical Specifications are adequate to ensure that 
the containment barriers are ensured. There is no affect on the 
plant safety analyses. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    (K) STF-322, Rev. 2, Secondary Containment Operability 
Clarification
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change involves an administrative clarification to reflect 
the original intent of the Technical Specifications. There is no 
impact on the availability of the secondary containment. 
Additionally, secondary containment is mitigative equipment. 
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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    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.
    This change involves an administrative clarification to reflect 
the original intent of the Technical Specifications. There is no 
impact on the availability of the secondary containment. There is no 
impact on the plant safety analyses. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.
    (L) TSTF-276, Rev. 2, Power Factor for Emergency Diesel 
Generator (EDG) Surveillences
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    These changes only affect mitigative equipment and therefore, 
would not have an impact on the probability of an accident. Also, 
the performance of the surveillances ensures that mitigative 
equipment is capable of performing its intended function. 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 new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The performance of the surveillances ensures that mitigative 
equipment is capable of performing its intended function. There are 
no degradations in equipment readiness to mitigate design events. 
There is no adverse affect on the plant safety analysis. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    (M) TSTF-65, Rev. 1, Generic Organizational Titles;
    TSTF-299, Rev. 0, Primary Coolant Sources Inspection 
Requirements;
    TSTF-279, Rev. 0, Inservice Testing Program Clarifications;
    TSTF-118, Rev. 0, and TSTF-106, Rev. 1, Diesel Generator Fuel 
Oil Testing Program Clarifications;
    TSTF-152, Rev. 0, Routine Reporting Requirement Upgrade
    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The changes to Technical Specification 5.0, Administrative 
Controls, are considered administrative changes. There are no 
changes to plant structures, systems or components involved with 
this change. There are no degradations in the availability of 
mitigative plant equipment. The proposed changes provide 
enhancements to the administrative controls in Technical 
Specifications, therefore, there is no affect on any plant safety 
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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The changes to Technical Specification 5.0, Administrative 
Controls, are considered administrative changes. There are no 
changes to plant structures, systems or components involved with 
this change. There are no degradations in the availability of 
mitigative plant equipment. The proposed changes provide 
enhancements to the administrative controls in Technical 
Specifications; therefore, there is no affect on any plant safety 
analyses. 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: Thomas S. O'Neill, Associate and General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Daniel S. Collins, Acting.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: August 20, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) regarding the requirement to 
demonstrate transfer of the unit A.C. electrical power supply to each 
offsite circuit and would increase the surveillance exceptions for the 
A.C. electrical sources in shutdown Modes 5 and 6. Also, the proposed 
amendment would delete the TS requirement that the auto-connected loads 
to each emergency diesel generator (EDG) do not exceed the 2000-hour 
rating of the EDG.
    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?

[[Page 57990]]

    No. The proposed surveillance requirement changes do not alter 
the design or operation of any structure, system, or component. No 
previously analyzed accident scenario is changed. Initiating 
conditions and assumptions remain as previously analyzed. The 
revised surveillance requirements will continue to assure adequate 
performance of structures, systems, and components. Therefore, 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?
    No. The proposed surveillance requirement changes do not alter 
the design or operation of any structure, system, or component. No 
new or different accident initiators are created as a result of the 
proposed changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    No. The proposed surveillance requirement changes do not reduce 
or adversely affect the capabilities of the offsite and onsite 
electrical power sources. The revised surveillance requirements will 
continue to assure adequate performance of structures, systems, and 
components. The proposed changes do not affect conformance of the 
electrical power systems to the applicable design criteria. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: September 1, 2004.
    Description of amendment request: The proposed amendments would 
revise the Operating Licenses' licensing basis to allow use of the code 
for Generation of Thermal-Hydraulic Information for Containment, 
Version 7.1patch1 (GOTHIC 7) to model Prairie Island Nuclear Generating 
Plant (PINGP) containment response for loss of coolant accidents (LOCA) 
and main steam line break (MSLB) accidents. The current PINGP 
containment response analyses are performed utilizing CONTEMPT. The 
Nuclear Management Company is making this request to support a 
transition option from internal analyses using CONTEMPT to an external 
analyses vendor (Westinghouse), which supports GOTHIC 7.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR), Part 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.
    The proposed amendment will change the Prairie Island Nuclear 
Generating Plant licensing basis by allowing use of the Generation 
of Thermal-Hydraulic Information for Containment, Version 7.1patch1, 
to model containment response for loss of coolant accident (LOCA) 
and main steam line break (MSLB) accidents.
    The containment is not an accident initiator, thus changing the 
containment modeling methodology does not increase the probability 
of an accident. This license amendment proposes to use a new 
methodology for modeling containment response analyses following an 
accident inside containment involving release of steam and water. 
This amendment does not alter the nuclear reactor core or reactor 
coolant system equipment, nor does it alter the methods or equipment 
used directly in mitigation of an accident. Thus radioactive 
releases inside containment due to an accident and radioactive 
releases from containment are not affected by the proposed change in 
analysis methodology. As discussed in Exhibits C and D, the Gothic 7 
sample results for the LOCA and MSLB transients predicted that the 
containment would remain below design pressure for both cases. 
Therefore, this change does not increase the consequences of an 
accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (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 amendment will change the Prairie Island Nuclear 
Generating Plant licensing basis by allowing use of the Generation 
of Thermal-Hydraulic Information for Containment, Version 7.1patch1, 
to model containment response for LOCA and MSLB accidents.
    The proposed amendment does not involve changes to plant design, 
hardware, system operation, or procedures involved with containment 
function. The proposed changes include application of new 
methodology for analysis of containment response following a loss of 
coolant accident or steam line break accident. The results of the 
analyses are used to demonstrate that the acceptance criteria for 
the containment structure continue to be met. These changes do not 
create the possibility for a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    (3) Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment will change the Prairie Island Nuclear 
Generating Plant (PINGP) licensing basis by allowing use of the 
Generation of Thermal-Hydraulic Information for Containment, Version 
7.1patch1 (GOTHIC 7), to model containment response for LOCA and 
MSLB accidents.
    The proposed licensing basis change to use GOTHIC 7 affects the 
design basis LOCA and MSLB containment accident analyses. As 
discussed in Exhibits C and D, the GOTHIC 7 sample results for the 
LOCA and MSLB transients predicted that the containment would remain 
below design pressure for both cases. The GOTHIC 7 accuracy in this 
application has been verified through benchmark analyses against the 
current analyses of record, validated against recognized standard 
data, and found to be appropriate for application to the PINGP 
design basis accidents. Safety analysis acceptance criteria are 
satisfied and adherence to safety analysis acceptance criteria using 
GOTHIC 7 assures that Technical Specification limits will not be 
exceeded during normal operation. 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 requests involve 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 Section Chief: L. Raghavan.

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: May 21, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the technical specifications (TS) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as

[[Page 57991]]

described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI Unit 2. Requirements related to 
combustible gas control were imposed by Order for many facilities and 
were added to or included in the TS for nuclear power reactors 
currently licensed to operate. The revised 10 CFR 50.44, ``Standards 
for Combustible Gas Control System in Light-Water-Cooled Power 
Reactors,'' eliminated the requirements for hydrogen recombiners and 
relaxed safety classifications and licensee commitments to certain 
design and qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated May 21, 2004.
    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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief (Acting): Mary Jane Ross-Lee.

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

    Date of amendment requests: August 26, 2004.
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TS) to implement 
ZIRLO<SUP>TM</SUP> fuel rod cladding material into the fuel design for 
San Onofre Nuclear Generating Station (SONGS), Units 2 and 3. 
Specifically, the licensee requests to add reference to 
ZIRLO<SUP>TM</SUP> clad fuel and filler rods in TS 4.2.1, ``Fuel 
Assemblies,'' and in TS 5.7.1.5, ``Core Operating Limits Report 
(COLR),'' add the following references to the list of analytical 
methods used to determine the core operating limits: ``Calculative 
Methods for the C-E Nuclear Power Large Break LOCA [loss-of-coolant 
accident] Evaluation Model,'' CENPD-1 32, Supplement 4-P-A, August 
2000, and ``Implementation of ZIRLO<SUP>TM</SUP> Cladding Material in 
CE [Combustion Engineering, Inc.] Nuclear Power Fuel Assembly 
Designs,'' CENPD-404-P-A, November 2001.

[[Page 57992]]

    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 allows the use of methods required for the 
implementation of ZIRLO<SUP>TM</SUP> clad fuel rods in San Onofre 
Nuclear Generating Station (SONGS) Units 2 and 3. The use of this 
methodology will not increase the probability of an accident because 
the plant systems will not be operated outside of design limits, no 
different equipment will be operated, and system interfaces will not 
change.
    As ZIRLO<SUP>TM</SUP> material is introduced to the reactor, 
transition cores will exist in which fuel assemblies containing 
ZIRLO<SUP>TM</SUP> and Zircaloy clad fuel rods are co-resident. Each 
type of fuel assembly (ZIRLO<SUP>TM</SUP> or Zircaloy clad fuel 
rods) will be evaluated based on the approved topical reports listed 
in TS 5.7.1.5.
    The use of this additional methodology will not increase the 
consequences of an accident because Limiting Conditions of Operation 
(LCOs) will continue to restrict operation to within the regions 
that provide acceptable results, and Reactor Protection System (RPS) 
trip setpoints will restrict plant transients so that the 
consequences of accidents will be acceptable. In addition, the 
consequences of the accidents will be calculated using NRC accepted 
methodologies.
    The transition cores that will exist as ZIRLO<SUP>TM</SUP> clad 
fuel is introduced to the reactor will not increase the consequences 
of an accident. Operation within the LCOs and RPS setpoints will 
continue to restrict plant transients so that the consequences of 
accidents will be acceptable.
    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 add any new equipment, modify any 
interfaces with any existing equipment, alter the equipment's 
function, or change the method of operating the equipment. The 
proposed change does not alter plant conditions in a manner that 
could affect other plant components. The proposed change does not 
cause any existing equipment to become an accident initiator. The 
ZIRLO<SUP>TM</SUP> clad fuel rod design does not introduce features 
that could initiate an accident.
    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.
    Safety Limits ensure that Specified Acceptable Fuel Design 
Limits (SAFDLs) are not exceeded during steady state operation, 
normal operational transients and anticipated operational 
occurrences. All fuel limits and design criteria shall be met based 
on the approved methodologies defined in the topical reports. The 
RPS in combination with the LCOs will continue to prevent any 
anticipated combination of transient conditions for reactor coolant 
system temperature, pressure, and thermal power level that would 
result in a violation of the Safety Limits. Therefore, the proposed 
changes will have no impact on the margins as defined in the 
Technical Specification bases.
    The safety analyses determine the LCO settings and RPS setpoints 
that establish the initial conditions and trip setpoints, which 
ensure that the Design Basis Events (Postulated Accidents and 
Anticipated Operational Occurrences) analyzed in the Updated Final 
Safety Analysis Report (UFSAR) produce acceptable results. In 
addition, all fuel limits and design criteria shall be satisfied. 
The Design Basis Events that are impacted by the implementation of 
ZIRLO<SUP>TM</SUP> cladding will be analyzed using the NRC accepted 
methodology described in CENPD-404-P-A.
    The change in the fuel rod cladding material and the use of the 
Emergency Core Cooling System (ECCS) performance evaluation models, 
CENPD-132, Supplement 4-P-A, ``Calculative Methods for the CE 
Nuclear Power Large Break LOCA Evaluation Model'' and CENPD-137, 
Supplement 2-P-A, ``Calculative Methods for the ABB [Asea Brown 
Boveri] CE Small Break LOCA Evaluation Model'' will not involve a 
reduction in the margin of safety because LCOs and Limiting Safety 
System Settings (LSSS) will be adjusted, if necessary, to maintain 
acceptable results for the impacted Design Basis Events.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Robert Gramm.

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

    Date of amendment request: May 21, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) to maintain 
hydrogen recombiners (Unit 2 only) and hydrogen and oxygen monitors. A 
notice of availability for this TS improvement using the consolidated 
line item improvement process was published in the Federal Register on 
September 25, 2003 (68 FR 55416). Licensees were generally required to 
implement upgrades as described in NUREG-0737, ``Clarification of TMI 
[Three Mile Island] Action Plan Requirements,'' and Regulatory Guide 
1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power Plants to 
Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to combustible gas control were imposed by Order 
for many facilities and were added to or included in the TSs for 
nuclear power reactors currently licensed to operate. The revised 10 
CFR 50.44, ``Standards for Combustible Gas Control System in Light-
Water-Cooled Power Reactors,'' eliminated the requirements for hydrogen 
recombiners and relaxed safety classifications and licensee commitments 
to certain design and qualification criteria for hydrogen and oxygen 
monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated May 21, 2004.
    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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not

[[Page 57993]]

contribute to the conditional probability of a large release up to 
approximately 24 hours after the onset of core damage. In addition, 
these systems were ineffective at mitigating hydrogen releases from 
risk-significant accident sequences that could threaten containment 
integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. Category 1 in RG 1.97 is intended for key variables that 
most directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen and oxygen monitors no 
longer meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 3, as defined in RG 1.97, is an appropriate categorization 
for the hydrogen monitors because the monitors are required to 
diagnose the course of beyond design-basis accidents. Also, as part 
of the rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 2, as defined in RG 1.97, is an appropriate categorization 
for the oxygen monitors, because the monitors are required to verify 
the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2,] and removal 
of the hydrogen and oxygen monitors from TSs will not prevent an 
accident management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and the site survey monitoring that support modification 
of emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TSs does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Previously Evaluated

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TSs will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen and oxygen monitor equipment was intended to mitigate a 
design-basis hydrogen release. The hydrogen recombiner and hydrogen 
and oxygen monitor equipment are not considered accident precursors, 
nor does their existence or elimination have any adverse impact on 
the pre-accident state of the reactor core or post accident 
confinement of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TSs, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TSs will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

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

    Date of amendment request: July 20, 2004.
    Description of amendment request: The proposed amendments would 
revise Administrative Controls Section 5.3.1 to replace the specific 
designation for the Health Physics Superintendent with a reference to 
the senior individual in charge of Health Physics, and to add 
flexibility to the qualification requirements for unit staff positions. 
This change supports Southern Nuclear Company's ongoing initiative to 
achieve fleet standardization.
    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?
    The proposed change to Technical Specifications Administrative 
Controls Section 5.3.1 involves the use of a more generic 
designation for the unit staff position responsible for Health 
Physics without reducing the level of authority required for that 
position. The proposed change also allows the flexibility to use an 
NRC accredited program for qualifying personnel to fill unit staff 
positions, which represents an acceptable alternative to the 
qualification requirements for these positions as currently 
specified in the Technical Specifications. Since the proposed 
changes are administrative in nature, they do not involve any 
physical changes to any structures, systems, or components, nor will 
their performance requirements be altered. The proposed changes also 
do not affect the operation, maintenance, or testing of the plant. 
Therefore, the response of the plant to previously analyzed 
accidents will not be affected. Consequently, 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 previously evaluated?
    The proposed changes to the Technical Specifications will have 
no adverse impact on the overall qualification of the unit staff. 
The alternative use of an accredited program that has been endorsed 
by the NRC will ensure the educational requirements and power plant 
experience for each unit staff position are properly satisfied and 
will continue to fulfill applicable regulatory requirements. Also, 
since no change is being

[[Page 57994]]

made to the design, operation, maintenance, or testing of the plant, 
no new methods of operation or failure modes are introduced by the 
proposed changes. Therefore, the possibility of a new or different 
kind of accident from any previously evaluated is not created.
    (3) Does the proposed change involve a significant decrease in 
the margin of safety?
    The proposed changes to the Technical Specifications will have 
no adverse impact on the onsite organizational features necessary to 
assure safe operation of the plant. Lines of authority for plant 
operation are unaffected by the proposed changes. Also, the adoption 
of the more generic designation of the individual responsible for 
Health Physics will reduce the regulatory burden of having to devote 
limited resources to process a license amendment whenever a title 
change for this position is implemented. Accordingly, this reduction 
in regulatory burden and the option to use an accredited program 
endorsed by NRC to qualify the unit staff will improve plant 
efficiency without compromising plant safety. Therefore, the 
proposed changes do not involve a significant decrease in the margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: May 21, 2004.
    Description of amendment request: The proposed amendment would 
delete the requirements from the Technical Specifications (TS) to 
maintain hydrogen recombiners and hydrogen monitors. A notice of 
availability for this improvement using the consolidated line item 
improvement process was published in the Federal Register on September 
25, 2003 (68 FR 55416). Licensees were generally required to implement 
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile 
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI Unit 2. Requirements related to 
combustible gas control were imposed by Order for many facilities and 
were added to or included in the TSs for nuclear power reactors 
currently licensed to operate. The revised 10 CFR 50.44, ``Standards 
for Combustible Gas Control System in Light-Water-Cooled Power 
Reactors,'' eliminated the requirements for hydrogen recombiners and 
relaxed safety classifications and licensee commitments to certain 
design and qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination (NSHC) for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated May 21, 2004.
    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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TSs will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TSs, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to

[[Page 57995]]

approximately 24 hours after the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

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

    Date of amendment request: May 21, 2004.
    Description of amendment request: The proposed amendment would 
delete the requirements from the Technical Specifications (TSs) to 
maintain hydrogen recombiners and hydrogen monitors. A notice of 
availability for the TS improvement using the consolidated line item 
improvement process was published in the Federal Register on September 
25, 2003 (68 FR 55416). Licensees were generally required to implement 
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile 
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
combustible gas control were imposed by Order for many facilities and 
were added to or included in the TSs for nuclear power reactors 
currently licensed to operate. The revised 10 CFR 50.44, ``Standards 
for Combustible Gas Control System in Light-Water-Cooled Power 
Reactors,'' eliminated the requirements for hydrogen recombiners and 
relaxed safety classifications and licensee commitments to certain 
design and qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination (NSHC) for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated May 21, 2004.
    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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44, the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TSs will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and the site survey monitoring that support modification 
of emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TSs, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TSs, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TSs, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TSs 
will not result in a significant reduction in

[[Page 57996]]

their functionality, reliability, and availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: August 26, 2004.
    Description of amendment request: The license amendment request 
proposes revising the Technical Specifications (TSs) to delete the TS 
requirements related to Hydrogen Analyzers and Hydrogen Recombiners 
consistent with NRC-approved TS Task Force (TSTF) Traveler number TSTF-
447, Revision 1, ``Elimination of Hydrogen Recombiners and Change to 
Hydrogen and Oxygen Monitors.'' The TS requirements related to Hydrogen 
Analyzers and Hydrogen Recombiners are contained in TS Tables 3.3-10 
and 4.3-10 and TSs 3.6.4.1 and 3.6.4.2. The availability of this TS 
improvement was announced in the Federal Register on September 25, 
2003, as part of the Consolidated Line Item Improvement Process 
(CLIIP).
    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 analysis endorses the NRC staff's generic no 
significant hazards consideration determination for TSTF-447 which was 
published in the Federal Register on September 25, 2003 (68 FR 55416) 
as follows:

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

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

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. Category 3 hydrogen monitors are adequate 
to provide rapid assessment of current reactor core conditions and 
the direction of degradation while effectively responding to the 
event in order to mitigate the consequences of the accident. The 
intent of the requirements established as a result of the TMI, Unit 
2 accident can be adequately met without reliance on safety-related 
hydrogen monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the request for amendments 
involves no significant hazards consideration.
    Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant

[[Page 57997]]

Hazards Consideration Determination, and Opportunity for a Hearing in 
connection with these actions was published in the Federal Register as 
indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see: (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (First Floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, (301) 415-4737 or by e-mail to <a href="/cdn-cgi/l/email-protection#bfcfdbcdffd1cddc91d8d0c9"><span class="__cf_email__" data-cfemail="24544056644a56470a434b52">[email protected]</span></a>.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendment: August 6, 2002, as supplemented 
December 12, 2002, July 24, 2003, and March 1, May 20, and August 11, 
2004.
    Brief description of amendment: The amendments replace the 
Technical Specifications 3.9.4 and 3.9.5 requirements to close all 
containment penetrations providing direct access from the containment 
atmosphere to outside temperature with a set of more detailed and less 
restrictive requirements.
    Date of issuance: September 13, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 268 and 244.
    Renewed Facility Operating License No. DPR-53: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63690).
    The December 12, 2002, July 24, 2003, March 1, 2004, and May 20, 
2004, letters provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination. The August 11, 2004, 
letter withdrew the licensee's requested changes to Technical 
Specification 3.9.3.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 13, 2004.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: December 12, 2003.
    Brief description of amendments: The amendments delete Technical 
Specification Section 5.5.3, ``Post-Accident Sampling.''
    Date of issuance: September 15, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 269 and 245.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19564).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 15, 2004.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: May 7, 2002, as supplemented 
April 7, 2003 and July 19, 2004.
    Brief description of amendment: The amendment relocates the 
boration system Technical Specification (TS) requirements to the 
Technical Requirements Manual and the boron dilution analysis 
restrictions within the TSs. The amendment also revises the TS limiting 
condition for operation action and the surveillance requirements 
associated with the emergency core cooling, containment spray and 
cooling and auxiliary feedwater systems.
    Date of issuance: September 9, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 283.
    Facility Operating License No. DRP-65: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: June 11, 2002 (67 FR 
40021). The April 7, 2003, and July 19, 2004, supplements contained 
clarifying information and did not change the staff's initial proposed 
finding of no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 9, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 7, 2002, as supplemented 
November 5, 2003.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) related to safety system settings. Specifically, 
the amendment revises: (1) TS 1.0 ``Definitions;'' (2) TS 2.2.1 
``Limiting Safety System Settings--Reactor Trip System Instrumentation 
Setpoints;'' (3) TS 3.3.1 ``Reactor Trip System Instrumentation;'' (4) 
TS 3.3.2 ``Engineered Safety Features Actuation System 
Instrumentation;'' (5) TS 3.7.7 ``Control Room Emergency Ventilation 
System;'' and (6) TS 3.8.3.1 ``Onsite Power Distribution--Operating.''
    Date of issuance: September 14, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 220.
    Facility Operating License No. DRP-49: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63692). The November 5, 2003, supplement contained clarifying 
information and did not change the staff's initial proposed finding of 
no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: May 25, 2004.

[[Page 57998]]

    Brief description of amendments: The amendments revised the 
licensing basis in the Updated Final Safety Analysis Report (UFSAR) to 
support installation of a low-pressure injection (LPI) cross connect 
inside containment. The changes to the UFSAR revise the licensing basis 
for selected portions of the core flood and LPI/Decay Heat Removal 
piping to allow exclusion of the dynamic effects associated with 
postulated rupture of that piping by application of leak-before-break 
technology. The amendments also revise the Technical Specifications 
(TSs) to delete TSs that will no longer apply when the LPI cross 
connect modification has been implemented.
    Date of issuance: September 2, 2004.
    Effective date: As of the date of issuance and shall be implemented 
during the fall 2004 refueling outage of Unit 3.
    Amendment Nos.: 340, 342, and 341.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: July 6, 2004 (69 FR 
40673). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 2, 2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: January 20, 2004, as 
supplemented by letters dated May 19, July 13, and August 16, 2004.
    Brief description of amendments: The amendments change the Prairie 
Island technical specification (TS) on containment to implement a 
portion of TSs Task Force Traveler 5, ``Revise containment requirements 
during handling irradiated fuel and core alterations.'' The amendments 
also selectively implement an alternative source term per Title 10 of 
the Code of Federal Regulations, Section 50.67 to perform the 
radiological consequences analysis of the design-basis fuel handling 
accident which supports the proposed TS changes.
    Date of issuance: September 10, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 166 and 156.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 2004 (69 FR 
29769 ).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 10, 2004.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of application for amendment: June 23, 2004.
    Brief description of amendment: The amendment removes a restriction 
from the Humboldt Bay Power Plant Unit 3 license thereby permitting 
Pacific Gas and Electric to engage in active decommissioning of the 
facility.
    Date of issuance: September 10, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 35.
    Facility Operating License No. DPR-7: This amendment revises the 
license.
    Date of initial notice in Federal Register: August 3, 2004 (69 FR 
46587).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 10, 2004.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: June 28, 2004, as supplemented by 
letter dated August 5, 2004.
    Brief Description of amendments: The amendments revise TS 3.4.13, 
``RCS [Reactor Coolant System] Operational Leakage,'' TS 5.5.9, ``Steam 
Generator [SG] Tube Surveillance Program,'' and TS 5.6.10, ``Steam 
Generator Tube Inspector Report.'' They also add a new TS 3.4.17, 
``Steam Generator Tube Integrity.'' These changes facilitate 
implementation of industry initiative NEI [Nuclear Energy Institute] 
97-08, ``Steam Generator Program Guidelines,'' which allows a 
comprehensive, performance-based approach to managing SG performance at 
Farley Nuclear Plant, Units 1 and 2.
    Date of issuance: September 10, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 163 and 156.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 2004 (69 FR 
46950). The supplemental letter dated August 5, 2004, provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determinations.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 10, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 6, 2002, as 
supplemented by letters dated December 19, 2002, March 28, June 24, 
September 3, and October 22, 2003.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to (1) relocate the pressure temperature limit 
curves and low temperature overpressure protection system limits to the 
Pressure and Temperature Limits Report (PTLR), (2) reference the PTLR 
in the affected TSs limiting conditions for operation and bases, 
including the addition of the PTLR to the definitions section of the 
TSs, and the addition of a new TS 6.9.1.15 to the administrative 
controls section of the TSs, (3) relocate TS 3.4.9.2, Pressurizer, to 
the Sequoyah Technical Requirements Manual and (4) revise TS 3.4.9.1, 
Pressure/Temperature Limits, Reactor Coolant System, and TS 3.4.12, Low 
Temperature Over Pressure Protection Systems, to incorporate standard 
TSs requirements from NUREG-1431, Revision 2, ``Standard Technical 
Specifications--Westinghouse Plants.''
    Date of issuance: September 15, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 294 and 284.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66015). The supplemental letters provided clarifying information that 
did not expand the scope of the original application or change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 57999]]

Safety Evaluation dated September 15, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 17th day of September, 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-21345 Filed 9-27-04; 8:45 am]
BILLING CODE 7590-01-P


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