[Federal Register Volume 69, Number 168 (Tuesday, August 31, 2004)]
[Notices]
[Pages 53098-53120]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-19586]



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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, August 6 through August 19, 2004. The last 
biweekly notice was published on August 19, 2004 (69 FR 51487).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
And Opportunity For a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ 
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or

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fact. Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by email to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the 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 email to 
[email protected].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: June 22, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) 
Vent and Drain Valves,'' to allow a vent or drain line with one 
inoperable valve to be isolated instead of requiring the valve to be 
restored to Operable status within 7 days.
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on February 24, 2003 
(68 FR 8637), on possible amendments to revise the action for one or 
more SDV vent or drain lines with an inoperable valve, 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 15, 2003 (68 FR 18294). The licensee affirmed 
the applicability of the model NSHC determination in its application 
dated June 22, 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

    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with one valve inoperable instead of requiring the valve to be 
restored to operable status within 7 days. With one SDV vent or 
drain valve inoperable in one or more lines, the isolation function 
would be maintained since the redundant valve in the affected line 
would perform its safety function of isolating the SDV. Following 
the completion of the required action, the isolation function is 
fulfilled since the associated line is isolated. The ability to vent 
and drain the SDV is maintained and controlled through 
administrative controls. This requirement assures the reactor 
protection system is not adversely affected by the inoperable 
valves. With the safety functions of the valves being maintained, 
the probability or consequences of an accident previously evaluated 
are not significantly increased.

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 (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
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 proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDV is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of the SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60666.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: April 23, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) Section 6.16, ``Post-Accident 
Sampling Programs NUREG 0737 (II.B.3, II-F.1.2),'' and the related 
requirements to maintain a Post-Accident Sampling System (PASS). 
Licensees were generally required to implement PASS upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide 1.97, Revision 3, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Access 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the NRC's lessons 
learned from the accident that occurred at TMI Unit 2. Requirements 
related to PASS were imposed by Order for many facilities and were 
added to or included in the TSs for nuclear power reactors currently 
licensed to operate. Lessons learned and improvements implemented over 
the last 20 years have shown that the information obtained from PASS 
can be readily obtained through other means or is of little use in the 
assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 3, 2003 (68 FR 10052) on possible amendments 
to eliminate PASS, 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 a 
license amendment application in the Federal Register on May 13, 2003 
(68 FR 25664). The licensee affirmed the applicability of the following 
NSHC determination in its application dated April 23, 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 PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
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. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the 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 PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in 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 PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radioisotopes 
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 PASS, 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. Methodologies that are 
not reliant on PASS are designed 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 use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Thomas S. O'Neill, Associate General 
Counsel, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville, 
IL 60555.
    NRC Section Chief: Richard J. Laufer.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: July 26, 2004.
    Description of amendments request: The proposed amendments would 
delete requirements from the Technical Specifications (TS) to maintain 
hydrogen recombiners and hydrogen and oxygen monitors. 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

[[Page 53101]]

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, 
``Combustible gas control for nuclear 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 no significant hazards consideration determination in its 
application dated July 26, 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 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 1.97, Category 1, is intended for key variables that 
most directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen and oxygen monitors no 
longer meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 3, as defined in RG 1.97, is an appropriate categorization 
for the hydrogen monitors because the monitors are required to 
diagnose the course of beyond design-basis accidents. Also, as part 
of the rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 2, as defined in RG 1.97, is an appropriate categorization 
for the oxygen monitors, because the monitors are required to verify 
the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
classification of the oxygen monitors as Category 2 and removal of 
the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs [severe 
accident management guidelines], the emergency plan (EP), the 
emergency operating procedures (EOPs), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

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

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

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in 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 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. 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 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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief (Acting): Michael L. Marshall.

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: June 21, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 5.5.14, ``Technical 
Specifications (TS) Bases Control Program,'' to replace the previous 10 
CFR 50.59 term ``unreviewed safety question'' with current terminology. 
The proposed amendment would also revise TS Section 5.7.1, ``High 
Radiation Area,'' to add wording that was inadvertently deleted with 
the issuance of the Improved Standard Technical Specifications in 
Amendment No. 176.
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

[[Page 53102]]

    The proposed changes do not modify the facility or the 
procedures for operation of the facility. One change updates the 
terminology used in 10 CFR 50.59 evaluations. The change does not 
alter the requirement of the TS Bases Control Program. The 
requirement for NRC review and approval of a TS Bases change is 
still determined through the use of the 10 CFR 50.59 review process. 
The second change corrects a typographical error that occurred under 
Amendment No. 176. The wording as proposed in this correction 
restores the requirement to the phraseology approved in Amendment 
No. 152 and is consistent with existing plant procedures.
    Since there are no changes to the facility or facility 
procedures, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not modify the facility or the 
procedures for operation of the facility. One change updates the 
terminology used in 10 CFR 50.59 evaluations. The change does not 
alter the requirement of the TS Bases Control Program. The 
requirement for NRC review and approval of a TS Bases change is 
still determined through the use of the 10 CFR 50.59 review process. 
The second change corrects a typographical error that occurred under 
Amendment No. 176. The wording as proposed in this correction 
restores the requirement to the phraseology approved in Amendment 
No. 152 and is consistent with existing plant procedures.
    Since there are no changes to the facility or facility 
procedures, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed changes continue to provide the controls necessary 
to ensure changes to the TS Bases are made in conformance with 10 
CFR 50.59. The proposed changes continue to provide the controls 
necessary to ensure adequate control of High Radiation Areas. The 
proposed changes will not result in any changes to the facility or 
facility operating procedures. Therefore, the changes do not result 
in a significant reduction in the margin of safety.
    Based on the above discussion, Carolina Power & Light has 
determined that 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 L. Marshall, Acting.

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

    Date of amendment request: June 9, 2004.
    Description of amendment request: The proposed change revises 
Technical Specifications (TS) Limiting Condition for Operation (LCO) 
3.4.11, ``RCS Pressure and Temperature (P/T) Limits,'' to replace the 
P/T curves for inservice leak and hydrostatic testing, non-nuclear 
heating and cooldown, and nuclear heating and cooldown currently 
illustrated in TS Figures 3.4.11-1, 3.4.11-2, and 3.4.11-3, 
respectively.
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes deal exclusively with the Reactor Coolant 
System (RCS) Pressure and Temperature (P/T) curves, which define the 
limitations for operation and testing. Because of the design 
conservatisms used to calculate the RCS P/T limits, reactor vessel 
failure has a low probability of occurrence and is not considered as 
a design basis accident in the safety analyses of the plant. The 
proposed changes adjust the reference temperature for the limiting 
material to account for irradiation effects and provide a comparable 
level of protection as previously evaluated and approved. The 
adjusted reference temperature calculations were performed in 
accordance with the requirements of 10 CFR [Part] 50 Appendix G 
using the guidance contained in RG [Regulatory Guide] 1.99, Revision 
2, ``Radiation Embrittlement of Reactor Vessel Materials,'' to 
provide operating limits for up to 33.1 EFPY [effective full power 
years]. The proposed license amendment does not involve a change to 
operation of equipment required to mitigate any accident analyzed in 
Columbia's UFSAR [Updated Final Safety Analysis Report]. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The revised P/T curves are based on a later edition and addenda 
of the ASME Code that incorporates current industry standards for 
the curves. The revised curves are also based on an RPV [reactor 
pressure vessel] fluence that has been recalculated in accordance 
with the methodology of RG 1.190. The proposed changes do not 
involve a modification to plant equipment. There is no effect on the 
function of any plant system, and no new system interactions are 
introduced by this change. No new failure modes are introduced. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed curves conform to the guidance contained in RG 
1.190, ``Calculational and Dosimetry Methods for Determining 
Pressure Vessel Neutron Fluence,'' and RG 1.99, Revision 2, 
``Radiation Embrittlement of Reactor Vessel Materials,'' and 
maintain the safety margins specified in 10 CFR [Part] 50 Appendix 
G. 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 C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: August 5, 2004.
    Description of amendment request: The proposed change will revise 
Technical Specification (TS) 5.5.12, ``Primary Containment Leakage Rate 
Testing Program,'' to allow a one-time deferral of the Type A 
containment integrated leak rate test (ILRT). The current 10-year 
interval between Type A tests would be extended to 15 years from the 
previous time a Type A test was performed. The last Type A test was 
performed on July 20, 1994.
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed one-time extension to the Type A testing interval 
from once-per-10 years to once-per-15 years will not increase the 
probability of an accident previously evaluated. The performance of 
Type A tests is not an accident initiator. The primary containment 
Type A testing interval extension does not involve a plant

[[Page 53103]]

modification and will not cause equipment failure or accident 
initiation.
    The proposed extension to the Type A testing interval does not 
involve a significant increase in the consequences of an accident. 
The NUREG 1493 generic study of the effects of extending containment 
leakage testing concluded that Type B and C testing can identify the 
vast majority (greater than 95 percent) of potential leakage paths 
and that reducing the Type A test interval to once-per-20 years 
leads to an ``imperceptible increase in risk.'' Other testing and 
inspection programs, in addition to the Type A test, provide a high 
degree of assurance that the primary containment integrity will be 
maintained. Inspections required by the Maintenance Rule and ASME 
Code [are] periodically performed in order to identify indications 
of containment degradation that could affect containment leak 
tightness.
    Experience at Columbia demonstrates that excessive containment 
leakage paths are detectable by Type B and C local leak rate tests. 
Type B and C testing will identify containment openings, such as a 
valve, that would otherwise be detected by the Type A test. These 
factors show that a one-time Type A test interval extension from 
once-per-10 years to once-per-15 years will not involve a 
significant increase in the consequences of an accident.
    Previous Type A test results at Columbia show leakage has not 
exceeded acceptance criteria in the past, indicating a leak-tight 
containment and demonstrating the structural capability of the 
primary containment. The testing results have established that 
Columbia has had acceptable containment leakage rates with 
considerable margin.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The Columbia primary containment is designed to contain energy 
and fission products during and after a design basis accident. The 
proposed extension of the Type A testing interval will not create 
the possibility of a new or different type of accident from any 
previously evaluated. There are no changes being made to the 
physical plant or in operation of the plant that could introduce a 
new failure mode with the potential to create an accident or affect 
mitigation of an accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed extension of the Type A testing interval will not 
significantly reduce the margin of safety. The NUREG 1493 generic 
study of the effects of extending containment leakage testing found 
that a 20-year interval in Type A leakage testing leads to an 
``imperceptible increase in risk.'' NUREG 1493 found that 
generically, the design containment leakage rate contributes less 
than 0.1 percent to the overall accident risk and that the increase 
in the Type A testing interval would have a minimal effect on risk 
because the vast majority (greater than 95 percent) of all potential 
leakage paths are detected by Type B and C leakage testing.
    A Columbia plant specific probabilistic risk assessment on the 
change in the Type A test interval from once-per-10 years to once-
per-15 years determined:
     The risk impact due to a change in Large Early Release 
Frequency (LERF) is an increase of 2E-8/year that is characterized 
by Regulatory Guide 1.174 [``An Approach for Using Probabilistic 
Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes 
to the Licensing Basis''] as ``very small.''
     The total integrated plant risk increase measured by 
person-rem/year is negligible.
     The change in conditional containment failure 
probability is an increase of 0.1 percent, which is considered to 
represent a very small impact on risk.
    Deferral of Type A testing for Columbia does not increase the 
level of risk to the public due to loss of capability to detect and 
measure containment leakage or loss of containment structural 
integrity. Other containment testing methods and inspections will 
assure all limiting conditions for operation will continue to be 
met. The margin of safety inherent in existing accident analyses 
will be maintained.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: June 22, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) to maintain 
hydrogen and oxygen monitors. A notice of availability for this 
technical specification improvement using the consolidated line item 
improvement process (CLIIP) was published in the Federal Register (FR) 
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 (not installed at FitzPatrick and therefore not addressed 
by this proposed amendment) 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 FR on September 25, 2003 (68 
FR 55416). The licensee affirmed the applicability of the model NSHC 
determination in its application dated June 22, 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 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

[[Page 53104]]

variables that most directly indicate the accomplishment of a safety 
function for design-basis accident events. The hydrogen and oxygen 
monitors no longer meet the definition of Category 1 in RG 1.97. As 
part of the rulemaking to revise 10 CFR 50.44 the Commission found 
that Category 3, as defined in RG 1.97, is an appropriate 
categorization for the hydrogen monitors because the monitors are 
required to diagnose the course of beyond design-basis accidents. 
Also, as part of the rulemaking to revise 10 CFR 50.44, the 
Commission found that Category 2, as defined in RG 1.97, is an 
appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2,] and removal 
of the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOPs), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

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

    The relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen and oxygen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The 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 relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that 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 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: 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. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: June 2, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to fully adopt the alternate 
source term (AST) methodology for design-basis accident dose 
consequence evaluations in accordance with 10 CFR 50.67. Specifically, 
the amendment would revise the TS Definition regarding dose equivalent 
iodine and TS Section 5.5.10, ``Ventilation Filter Testing Program 
(VFTP).'' The AST methodology for the fuel-handling accident was 
previously approved in Amendment No. 215, dated March 17, 2003.
    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 involves the reanalysis of design basis 
radiological accidents in Containment and the Fuel Storage Building. 
The new analyses, based on the Alternate Source Term (AST), in 
accordance with 10 CFR 50.67, will replace the existing analyses 
that are based on the methodologies of [Atomic Energy Commission 
Report, ``Calculation of Distance Factors for Power and Test Reactor 
Sites,'' 1962] TID-14844. As a result of the new analyses, changes 
to the Technical Specifications are proposed which take credit for 
the new analysis results.
    The proposed changes to the Technical Specifications modify 
requirements regarding filter testing for a variety of systems 
(i.e., Containment Purge, Fuel Storage Building Emergency 
Ventilation). The analyses do not credit charcoal or HEPA [high-
efficiency particulate air] filtration for dose mitigation. The 
proposed changes reflect the plant configuration that will support 
implementation of the AST analyses.
    The AST analysis follows the guidance of the NRC Regulatory 
Guide 1.183 and uses the acceptance criteria of the NRC Standard 
Review Plan (NUREG-0800) for offsite doses and General Design 
Criteria for Control Room personnel. The accident analyses 
conservatively assume that the Containment Building and the Fuel 
Storage Building, including ventilation filtration systems for those 
buildings, do not diminish or delay the assumed fission product 
release.
    The proposed changes also revise the definition of Dose 
Equivalent Iodine (DEI) to be consistent with the assumptions of the 
analyses. The limits for DEI do not change as a result of the 
implementation of the AST analyses.
    The change from the original source term to the new proposed AST 
is a change in analysis method and assumptions and has no effect on 
accident initiators or causal factors that contribute to the 
probability of occurrence of previously analyzed accidents. Use of 
AST to analyze the dose effect of design basis accidents shows that 
regulatory acceptance criteria for the new methodology continue to 
be met. Changing the analysis methodology does not change the 
sequence or progression of the accident scenario.
    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?
    Response: No.
    The changes proposed in this license amendment request involve 
the use of a new analysis methodology and related regulatory 
acceptance criteria. In addition, certain changes to plant 
ventilation systems can be made based on the analysis results, using 
the new methodology. Use of a new analysis

[[Page 53105]]

method does not impact the design or operation of plant systems or 
components and new accident scenarios would therefore not be 
created. The proposed changes to air ventilation and filtration 
systems do not adversely affect plant equipment used to protect 
plant safety limits or the way in which that plant equipment is 
operated or maintained. As a result, no new failure modes are being 
introduced that could lead to different accidents.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The existing dose analysis methodology and assumptions 
demonstrate that the dose consequences for all design basis 
accidents are within regulatory limits for whole body and thyroid 
doses as established in 10 CFR 100 (except for the Fuel Handling 
Analysis, which is already based on the AST methodology). The 
alternate dose analysis methodology and assumptions also demonstrate 
that the dose consequences of these accidents are within the 
regulatory requirements established for the new methodology.
    The limits applicable to the alternate analysis are established 
in 10 CFR 50.67 in conjunction with the Total Effective Dose 
Equivalent (TEDE) acceptance directed in Regulatory Guide 1.183. The 
acceptance criteria for both dose analysis methods have been 
developed for the purpose of evaluating design basis accidents to 
demonstrate adequate protection of public health and safety. An 
acceptable margin of safety is inherent in both types of acceptance 
criteria.
    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: 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. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: June 3, 2004.
    Description of amendment request: The proposed amendment would 
increase the maximum authorized reactor core power level from 3067.4 
megawatt thermal (MWt) to 3216 MWt. This represents a nominal increase 
of 4.85% rated thermal power. The amendment would also revise the 
Technical Specifications (TSs) to relocate certain cycle-specific 
parameters to the Core Operating Limits Report (COLR) by adopting TS 
Task Force Traveler TSTF-339, ``Relocate Technical Specification 
Parameters to the COLR.'' In addition, the amendment would revise 
several allowable values in TS Table 3.3.1-1, ``Reactor Protection 
System (RPS) Instrumentation,'' and Table 3.3.2-1, ``Engineered Safety 
Feature Actuation System (ESFAS) Instrumentation.''
    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 evaluations and analyses associated with this proposed 
change to core power level have demonstrated that all applicable 
acceptance criteria for plant systems, components, and analyses 
(including the Final Safety Analysis Report Chapter 14 safety 
analyses) will continue to be met for the proposed increase in 
licensed core thermal power for Indian Point 3 (IP3). The subject 
increase in core thermal power will not result in conditions that 
could adversely affect the integrity (material, design, and 
construction standards) or the operational performance of any 
potentially affected system, component or analysis. Therefore, the 
probability of an accident previously evaluated is not affected by 
this change. The subject increase in core thermal power will not 
adversely affect the ability of any safety-related system to meet 
its intended safety function. Further, the radiological dose 
evaluations in support of this power uprate effort show all 
acceptance criteria are met.
    The relocation of cycle-specific core operating limits from the 
Technical Specifications to the Core Operating Limits Report (COLR), 
in accordance with TSTF-339, has no influence or impact on the 
probability or consequences of a Design Basis Accident. Adherence to 
the COLR and accepted methodologies for establishing COLR parameters 
continues to be controlled by the plant Technical Specifications. 
Relocation of cycle-specific values to the COLR while maintaining 
the limiting requirements in the Technical Specifications reduces 
administrative burden associated with processing license amendments 
for routine core reload designs.
    RPS and ESF [engineered safety feature] allowable values 
established in plant technical specifications represent acceptance 
criteria used by plant personnel in assessing the operability of 
instrumentation channels.
    Allowable values are not accident initiators and have no role in 
the probability of occurrence of an accident. Safety analyses for 
design basis accidents use certain assumptions (Safety Analysis 
Limits) regarding the actuation of RPS and ESF protective functions. 
The proposed allowable values are developed using a methodology that 
assures the accident analysis assumptions are valid and the 
consequences of previously analyzed accidents continue to meet 
established limits.
    Therefore, the proposed changes described in this license 
amendment request do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The analyses and evaluations performed for the proposed increase 
in power show that all applicable acceptance criteria for plant 
systems, components, and analyses (including FSAR [Final Safety 
Analysis Report] Chapter 14 safety analyses) will continue to be met 
for the proposed power increase in IP3 licensed core thermal power. 
The subject increase in core thermal power will not result in 
conditions that could adversely affect the integrity (material, 
design, and construction standards) or operational performance of 
any potentially affected system, component, or analyses. The subject 
increase in core thermal power will not adversely affect the ability 
of any safety-related system to meet its safety function. 
Furthermore, the conditions and changes associated with the subject 
increase in core thermal power will neither cause initiation of any 
accident, nor create any new credible limiting single failure. The 
power uprate does not result in changing the status of events 
previously deemed to be non-credible being made credible. 
Additionally, no new operating modes are proposed for the plant as a 
result of this requested change.
    The relocation of cycle-specific core operating limits from the 
Technical Specifications to the Core Operating Limits Report (COLR), 
in accordance with TSTF-339, does not involve any changes to plant 
equipment or the way is which the plant is operated. There are no 
new accident initiators or causal mechanisms being introduced by 
this proposed change. Relocation of cycle-specific values to the 
COLR while maintaining the limiting requirements in the Technical 
Specifications reduces administrative burden associated with 
processing license amendments for routine core reload designs.
    RPS and ESF allowable values established in plant technical 
specifications represent acceptance criteria used by plant personnel 
in assessing the operability of instrumentation channels. Revising 
allowable values does not involve installation of new equipment, 
modification to existing equipment, or a change in plant operation 
that could create a new or different accident scenario.
    Therefore, the proposed changes described in this license 
amendment request will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?

[[Page 53106]]

    Response: No.
    The analyses and evaluations associated with the proposed 
increase in power show that all applicable acceptance criteria for 
plant systems, components, and analyses (including FSAR Chapter 14 
safety analyses) will continue to be met for this proposed increase 
in IP3 licensed core thermal power. The subject increase in core 
thermal power will not result in conditions that could adversely 
affect the integrity (material, design, and construction standards) 
or operational performance of any potentially affected system, 
component, or analysis. The subject power uprate will not adversely 
affect the ability of any safety-related system to meet its intended 
safety function.
    Adoption of TSTF-339 allows relocation of cycle-specific 
parameters to the COLR, while maintaining limiting requirements in 
the Technical Specifications. Approved methodologies for calculating 
cycle-specific parameters are maintained in the Technical 
Specifications, and changes to the COLR are subject to the 
requirements and controls of 10 CFR 50.59. This assures that 
required margins to safety limits are maintained.
    The proposed new allowable values are developed using 
established methodologies and incorporate additional conservatism 
that assures the validity of analysis limits assumed in the 
evaluation of hypothetical accidents.
    Therefore, the proposed changes described in this license 
amendment request will 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: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: July 8, 2004.
    Description of amendment request: Delete Technical Specification 
Surveillance Requirement 4.5.2.d.1, Emergency Core Cooling System 
Subsystems -Tave >= 300 [deg]F, associated with the 
requirement to maintain an operable Automatic Closure Interlock (ACI) 
for the Shutdown Cooling (SDC) suction isolation valves.
    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 removal of the ACI function is consistent with the 
guidelines previously endorsed by the NRC in Generic Letter 88-17. 
Removal of this function results in a calculated decrease in 
intersystem Loss of Coolant Accident (ISLOCA) frequency. 
Additionally, the removal of the ACI function will result in a 
decrease in SDC system unavailability and a corresponding decrease 
in risk associated with loss of SDC events. As a result, the 
proposed change will result in a net decrease in risk and a net 
improvement in plant 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.
    The presence or omission of an ACI function is not considered an 
accident initiator nor is this function credited in any safety 
analyses for the prevention or mitigation of any accident. Alarms, 
design features, and strict administrative/procedural controls 
support correct and timely operator action to ensure the SDC system 
will not be exposed to high Reactor Coolant System (RCS) pressure.
    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 ACI function is not credited in a margin of safety analysis 
for any accident previously evaluated. Removal of the ACI function 
will result in an overall net increase in nuclear safety. 
Appropriate alarm, design features, and administrative controls will 
continue to ensure proper isolation and isolation maintenance of the 
SDC system during plant operations with elevated RCS pressures.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: July 8, 2004. This supersedes the May 
12, 2004, application in its entirety (69 FR 34699).
    Description of amendment request: The proposed amendment would 
change the reactor core analytical methods used to determine the core 
operating limits, reflect the changes allowed by Technical 
Specification (TS) Task Force (TSTF) Traveler No. 363, ``Revised 
Topical Report References in ITS [Improved Standard Technical 
Specifications] 5.6.5, COLR [Core Operating Limits Report],'' and 
delete the Index from the TSs. This request completely supersedes the 
previous request of May 12, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
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.

TS 6.9.5.1, Core Operating Limits Report (COLR)

    The proposed amendment, in part, identifies a change in the 
nuclear physics codes used to confirm the values of selected cycle-
specific reactor physics parameter limits and includes minor 
editorial changes which do not alter the intent of stated 
requirements. The proposed change also allows the use of methods 
required for the implementation of ZIRLO clad fuel rods. Inasmuch as 
the proposed change includes codes that have been previously 
approved by the NRC for CE [Combustion Engineering] cores, the 
amendment is administrative in nature and has no impact on any plant 
configuration or system performance relied upon to mitigate the 
consequences of an accident. Parameter limits specified in the COLR 
for this amendment are not changed from the values presently 
required by TSs. Future changes to the calculated values of such 
limits may only be made using NRC approved methodologies, must be 
consistent with all applicable safety analysis limits, and are 
controlled by the 10 CFR 50.59 process. Assumptions used for 
accident initiators and/or safety analysis acceptance criteria are 
not altered by this change.
    The proposed change will add an NRC approved topical report, 
WCAP-16072-P-A, to the list of referenced topical reports. The 
topical report has been previously approved by the NRC for use in 
Combustion Engineering core designs and as such, the proposed change 
is administrative in nature and has no impact on any plant 
configurations or on system performance that is relied upon to 
mitigate the consequences of an accident. In addition, prior to the 
use

[[Page 53107]]

of the ZrB2 burnable absorber coating, fuel design will 
be analyzed with applicable NRC staff approved codes and methods.
    The proposed change also implements NRC approved TSTF Traveler 
No. 363. This is an administrative change that will allow specific 
details, such as the revision number, revision date, and supplement 
number of topical reports that are referenced in the TSs, to be 
deleted and relocated in the cycle specific COLR. This proposed 
change does not result in any changes to the assumptions used to 
evaluated [evaluate] accident initiators and/or safety analysis 
acceptance criteria.

Index

    The proposed deletion of the Index is purely administrative and 
does not impact the accident analysis.
    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.

TS 6.9.5.1, Core Operating Limits Report (COLR)

    The proposed change, in part, identifies a change in the nuclear 
physics codes used to confirm the values of selected cycle-specific 
reactor physics parameter limits. The proposed change also allows 
the use of methods required for the implementation of ZIRLO clad 
fuel rods. Neither of these changes results in a change to the 
physical plant or to the modes of operation defined in the facility 
license.
    The proposed change adds a reference to the topical report that 
allows the use of ZrB2 as a burnable absorber coating on 
the fuel pellet. The topical report has been previously approved by 
the NRC for use in Combustion Engineering core designs and as such, 
the proposed change is administrative in nature and has no impact on 
any plant configurations or on system performance that is relied 
upon to mitigate the consequences of an accident. In addition, prior 
to the use of the ZrB2 burnable absorber coating, fuel 
design will be analyzed with applicable NRC staff approved codes and 
methods. This change is administrative in nature and does not create 
a new or different type of accident than previously evaluated 
because the design requirements for the facility remain the same.
    The proposed change also implements TSTF Traveler No. 363. The 
proposed change does not result in changes to the physical plant or 
to the modes of operation defined in the facility license nor does 
it involve the addition of new equipment or the modification of 
existing equipment.

Index

    The proposed deletion of the Index is purely administrative has 
no affect on existing equipment.
    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.

TS 6.9.5.1, Core Operating Limits Report (COLR)

    The proposed changes to change the nuclear physics code package 
and to add a topical report to support the use of ZIRLO do not amend 
the cycle specific parameter limits located in the COLR from the 
values presently required by the TS. The individual specifications 
continue to require operation of the plant within the bounds of the 
limits specified in COLR. Benchmarking has shown that uncertainties 
for the Westinghouse Physics code system yields are essentially the 
same or less than those obtained for the current ROCS and DIT 
[computer code] methodology. Future changes to the values of these 
limits by the licensee may only be developed using NRC approved 
methodologies, must remain consistent with all applicable plant 
safety analysis limits addressed in the Safety Analysis Report, and 
are further controlled by the 10 CFR 50.59 process. The relocation 
of the supplement numbers, revision numbers, and approval dates of 
the analytical methods listed in the COLR does not affect the margin 
of safety. The analysis will continue to be performed using NRC 
approved methodology. Safety analysis acceptance criteria are not 
being altered by this amendment.
    The proposed change will add WCAP-16072-P-A to the list of 
referenced topical reports. The topical report has been previously 
approved by the NRC for use in Combustion Engineering core designs 
and as such, the proposed change is administrative in nature and has 
no impact on any plant configurations or on system performance that 
is relied upon to mitigate the consequences of an accident. In 
addition, prior to the use of the ZrB2 burnable absorber 
coating, fuel design will be analyzed with applicable NRC staff 
approved codes and methods.

Index

    The proposed deletion of the Index, which is an administrative 
document, does not impact any TS values or safety limits.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: June 10, 2004, as supplemented by letter 
dated July 21, 2004.
    Description of amendment request: The proposed amendments would 
revise the Quad Cities Nuclear Power Station (QCNPS) technical 
specifications (TS) to change the allowable value (AV) and add 
surveillance requirements (SRs) for the main steam line (MSL) flow-high 
initiation of Group 1 primary containment isolation and control room 
emergency ventilation system isolation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    For QCNPS, Units 1 and 2, the proposed amendment will implement 
a design change that upgrades the existing MSL Flow-High 
instrumentation from pressure switches to analog trip unit devices. 
Analog trip units (ATUs) have proven to be a more reliable 
technology than the currently installed equipment. Analog trip units 
are used in various applications at QCNPS, including the Reactor 
Protection System (RPS) low water level trip function. Because the 
trip units are more reliable, the likelihood of spurious isolations 
is reduced. Further, ATUs experience less instrument drift during 
the operating cycle. The proposed change adds a 92-day trip unit 
calibration requirement for the MSL-High isolation function. The NRC 
has previously found that a 92-day calibration is appropriate for 
individual ATUs.
    Procedure revisions required by this modification are limited to 
those associated with the calibration, maintenance, and operation of 
the replacement transmitter and trip unit analog loops. All required 
design functions of the MSL high flow loop are maintained. No 
system, structure, or component will be used in a manner that is not 
already bounded by the reference design, or is inconsistent with 
analyses or descriptions in the QCNPS Updated Final Safety Analysis 
Report (UFSAR). There is no adverse effect on the performance or 
control of any design function described in the UFSAR.
    TS requirements that govern operability or routine testing of 
plant instruments are not assumed to be initiators of any analyzed 
event because these instruments are intended to prevent, detect, or 
mitigate accidents. Therefore, these changes will not involve an 
increase in the probability of occurrence of an accident previously 
evaluated. In addition, these changes will not increase the

[[Page 53108]]

consequences of an accident previously evaluated because the 
proposed change does not adversely impact structures, systems, or 
components. The planned instrument upgrade is a more reliable design 
than existing equipment. The proposed changes establish requirements 
that ensure components are operable when necessary for the 
prevention or mitigation of accidents or transients. Furthermore, 
there will be no change in the types or significant increase in the 
amounts of any effluents released offsite. For these reasons, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes support a planned instrumentation upgrade 
by incorporating SRs required to ensure operability. The change does 
not adversely impact the manner in which the instrument will operate 
under normal and abnormal operating conditions. Therefore, these 
changes provide an equivalent level of safety and will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated. The changes in methods governing 
normal plant operation are consistent with the current safety 
analysis assumptions.
    All required design functions are maintained, and the new 
setpoint is analyzed in accordance [with] an NRC-approved 
methodology for determination of setpoints and TS AVs in accordance 
with the QCNPS UFSAR, Section 7.3.2.4, ``Design Evaluation.'' 
Therefore, replacing the existing MSL high flow DPISs with analog 
trip instrumentation does not alter any UFSAR described evaluation 
methodologies, or introduce any new methodologies. These changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes support a planned instrumentation upgrade 
from differential pressure switches to ATUs. The proposed changes do 
not adversely affect the probability of failure or availability of 
the affected instrumentation. The addition of a 92-day trip unit 
calibration for MSL Flow-High is a conservative change that aligns 
the SRs for a planned instrumentation upgrade with that of similar 
instrumentation. The NRC has previously found that a 92-day 
calibration is appropriate for individual ATUs. The setpoint was 
determined using an NRC-approved methodology. The proposed changes 
do not affect the analytical limit assumed in the safety analyses 
for the actuation of the instrumentation. Therefore, it is concluded 
that the proposed changes will not result in 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: March 22, 2004 as supplemented July 23, 
2004.
    Description of amendment request: The proposed change allows 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 TS, 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 are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
4.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
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 March 22, 2004 and July 23, 
2004, supplement.
    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.


[[Page 53109]]


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

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit No. 2 (BVPS-2), Beaver County, 
Pennsylvania

    Date of amendment request: July 23, 2004.
    Description of amendment request: The proposed amendment would 
revise the BVPS-2 Technical Specifications to eliminate periodic 
response time testing requirements on selected sensors and selected 
protection channel components and permit the option of measuring or 
verifying the response times by means other than testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change to the Technical Specifications does not result in a 
condition where the design, material, and construction standards 
that were applicable prior to the change are altered. The same RTS 
[reactor trip system] and ESFAS [engineered safety features 
actuation system] instrumentation is being used; the time response 
allocations/modeling assumptions in the Updated Final Safety 
Analysis Report (UFSAR) Chapter 15 analyses are still the same; only 
the method of verifying [the] time response is changed. The proposed 
change will not modify any system interface and could not increase 
the likelihood of an accident since these events are independent of 
this change. The proposed activity will not change, degrade or 
prevent actions or alter any assumptions previously made in 
evaluating the radiological consequences of an accident described in 
the UFSAR.
    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.
    This change does not alter the performance of the pressure and 
differential pressure transmitters, process protection racks, 
Nuclear Instrumentation, and logic systems used in the Reactor Trip 
and Engineered Safety Features Actuation Systems. All sensors, 
process protection racks, Nuclear Instrumentation, and logic systems 
will still have response time verified by [a] test before placing 
the equipment into operational service and after any maintenance 
that could affect the response time. Changing the method of 
periodically verifying instrument response times for certain 
equipment (assuring equipment operability) from time response 
testing to calibration and channel checks will not create any new 
accident initiators or scenarios. Periodic surveillance of these 
instruments will detect significant degradation in the equipment 
response time characteristics.
    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 does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method for selected sensors and differential pressure 
sensors and for process protection racks, Nuclear Instrumentation, 
and logic systems is modified to allow use of actual test data or 
engineering data. The method of verification still provides 
assurance that the total system response time is within that assumed 
in the safety analysis.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: April 26, 2004.
    Description of amendment request: This proposed license amendment 
would revise the frequency of the Mode 5 Intermediate Range Monitoring 
(IRM) Instrumentation CHANNEL FUNCTIONAL TEST contained in Technical 
Specification (TS) 3.3.1.1 from 7 days to 31 days. The methodology used 
to analyze the change in testing frequency is based upon guidance 
contained in Generic Letter 91-04, ``Changes in Technical Specification 
Surveillance Intervals to Accommodate a 24-month Fuel Cycle,'' and 
Electric Power Institute (EPRI) Report TI-103335, ``Guidance for 
Instrumentation Calibration Extension/Reduction Programs.''
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed Technical Specification (TS) change involves an 
increase in the Mode 5 CHANNEL FUNCTIONAL TEST interval for Reactor 
Protection System (RPS) Intermediate Range Monitor (IRM) from 7 days 
to 31 days. The proposed TS change does not alter the design or 
functional requirements of the RPS or IRM systems. Evaluation of the 
proposed testing interval change demonstrated that the availability 
of the IRMs to prevent or mitigate the consequences of a control rod 
withdrawal event at low power levels are not significantly affected 
because of other, more frequent testing that is performed, the 
availability of redundant systems and equipment, and the high 
reliability of the IRM equipment.
    Furthermore, using the guidance of GL 91-04, a historical review 
of surveillance test results and associated maintenance records did 
not indicate evidence of any failure that would invalidate the above 
conclusions.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed TS change involves an increase in the Mode 5 IRM 
CHANNEL FUNCTIONAL TEST interval from 7 days to 31 days. Existing TS 
testing requirements ensure the operability of the IRMs. The 
proposed TS change does not introduce any failure mechanisms of a 
different type than those previously evaluated, since no physical 
changes to the plant are being made. No new or different equipment 
is being installed, and no installed equipment is being operated in 
a different manner. As a result, no new failure modes are 
introduced. In addition, the manner in which surveillance tests are 
performed remain unchanged.
    Furthermore, using the guidance in GL 91-04, a historical review 
of surveillance test results and associated maintenance records did 
not indicate evidence of any failure that would invalidate the above 
conclusions.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change will not involve a single reduction in 
the margin of safety.
    The proposed Technical Specifications (TS) change involves an 
increase in the Mode 5 CHANNEL FUNCTIONAL TEST interval for Reactor 
Protection System (RPS) Intermediate Range Monitor (IRM) from 7 days 
to 31 days. The impact on system operability is minimal, based upon 
performance of the more frequent Channel

[[Page 53110]]

Checks, continuous Control Room monitoring when the IRMs are in use, 
and the overall IRM reliability. Evaluations show there is no 
evidence of time-dependent failures that would impact the 
availability of the IRMs.
    Furthermore, using the guidance in GL 91-04, a historical review 
of surveillance test results and associated maintenance records did 
not indicate evidence of any failure that would invalidate the above 
conclusions.
    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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 28, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.9.4, ``Containment Building 
Penetrations,'' to align the language of the Surveillance Requirement 
with the Applicability Statement contained in the Limiting Condition 
for Operation.
    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. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change aligns the language of the Surveillance 
Requirement for Containment Building Penetrations with the language 
of the Applicability Statement of Technical Specification 3.9.4.
    The proposed amendment will not change the design function, or 
method of performing or controlling design functions, of structures, 
systems and components, nor will there be an effect on FPL Energy 
Seabrook programs. As a result, the proposed amendment will not 
change assumptions, or change, degrade or prevent actions described 
or assumed in accidents evaluated and described in the Seabrook 
Station UFSAR [updated final safety analysis report]. The proposed 
change to the Surveillance Requirement wording does not adversely 
affect performance of the Surveillance Requirement that verifies the 
status of Containment Building Penetrations. Since the status of the 
Containment Penetrations is not adversely affected by the proposed 
change, the radiological consequences of an event are unchanged. 
Therefore, the proposed amendment does not result in an increase in 
the radiological consequences of any accident described in the 
Seabrook Station UFSAR.
    Therefore, it is concluded that these proposed changes do not 
involve a significant increase in the probability or consequence of 
an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change aligns the language of the Surveillance 
Requirements for Containment Building Penetrations with the language 
in the Applicability Statement of the Technical Specification.
    The proposed amendment will not change the design function, or 
method of performing or controlling design functions, of structures, 
systems and components, nor will there be an effect on FPL Energy 
Seabrook programs. As a result, there are no changes associated with 
the proposed amendment that could potentially introduce new failure 
modes or accident scenarios.
    Therefore, it is concluded that these proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed change aligns the language of the Surveillance 
Requirement for Containment Building Penetrations with the language 
of the Applicability Statement of Technical Specification 3.9.4. The 
proposed amendment does not change the design function, or method of 
performing or controlling design functions, of structures, systems 
and components, nor will there be an effect on FPL Energy Seabrook 
programs. The status of containment penetrations will continue to be 
verified. The proposed change does not involve any changes to a 
margin of safety.
    Therefore, it is concluded that these proposed changes do not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: August 17, 2004.
    Description of amendment request: The licensee proposed to revise 
Section 3.3.1, ``Oxygen Concentration [of the primary containment],'' 
of the Technical Specifications (TSs) to (1) add a new action allowing 
24 hours to restore the oxygen concentration to within the limit of <4% 
by volume if the limit is exceeded when the reactor is in the power 
operating condition, and (2) incorporate the associated conforming 
changes of editorial nature. The proposed 24-hour completion time for 
restoring oxygen concentration is consistent with Improved Standard 
Technical Specifications for Boiling Water Reactors (NUREG-1433, 
Revision 3).
    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 NRC staff's analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The effect of the proposed amendment is to provide the same 
24-hour completion time to restore oxygen concentration to under the 4% 
limit should the oxygen concentration rise due to other than a reactor 
shutdown-startup evolution. The proposed amendment does not lead to, 
nor is it the result of, a plant design change. These TS changes will 
not lead to alteration of the physical design or operational procedures 
associated with the containment system, or any other plant structure, 
system, or component (SSC). All requirements needed to assure 
operability of the containment system will remain unchanged. 
Containment atmospheric oxygen concentration was not assumed to be a 
precursor of accidents, nor was it assumed to be a component in 
previously evaluated accident scenarios. Accordingly, the revised 
specifications will lead to no increase in the consequences of an 
accident previously evaluated, and no increase of the probability of an 
accident previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. As stated above, the proposed amendment involves only the 
time allowed to restore containment atmospheric oxygen concentration to 
under 4 percent by volume, and associated editorial changes. These

[[Page 53111]]

changes do not alter the physical design, safety limits, or method of 
operation associated with the operation of the plant. Accordingly, the 
changes do not introduce any new or different kind of accident from 
those previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since the licensee did not propose to 
exceed or alter a design basis or safety limit, did not propose to 
operate any component in a less conservative manner, and did not 
propose to use a less conservative analysis methodology, the proposed 
amendment will not affect in any way the performance characteristics 
and intended functions of any SSC. Therefore, the proposed amendment 
does not involve a significant reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: July 6, 2004.
    Description of amendment request: The proposed change involves the 
extension from 1 hour to 24 hours for the completion time (CT) of 
Technical Specification (TS) 3.3.a.2.B, which defines requirements for 
accumulators. Accumulators are part of the emergency core cooling 
system and consist of tanks partially filled with borated water and 
pressurized with nitrogen gas. The contents of the tank are discharged 
to the reactor coolant system (RCS) if, as during a loss-of-coolant 
accident, the coolant pressure decreases to below the accumulator 
pressure. TS 3.3.a.2.B specifies a CT to restore an accumulator to 
operable status when it has been declared inoperable for a reason other 
than the boron concentration of the water in the accumulator not being 
within the required range. This change was proposed by the Westinghouse 
Owners Group participants in the TS Task Force (TSTF) and is designated 
TSTF-370, ``Increase Accumulator Completion Time from 1 Hour to 24 
Hours.'' TSTF-370 is supported by NRC-approved Topical Report WCAP-
15049-A, ``Risk-Informed Evaluation of an Extension to Accumulator 
Completion Times,'' submitted on May 18, 1999. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on July 15, 
2002 (67 FR 46542), on possible amendments concerning TSTF-370, 
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 March 12, 2003 (68 FR 11880). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated July 6, 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 basis for the accumulator limiting condition for operation 
(LCO), as discussed in [Standard Technical Specifications] Bases 
Section 3.5.1, is to ensure that a sufficient volume of borated 
water will be immediately forced into the core through each of the 
cold legs in the event the RCS pressure falls below the pressure of 
the accumulators, thereby providing the initial cooling mechanism 
during large RCS pipe ruptures. As described in Section 9.2 of the 
WCAP-15049, ``Risk-Informed Evaluation of an Extension to 
Accumulator Completion Times,'' evaluation, the proposed change will 
allow plant operation in a configuration outside the design basis 
for up to 24 hours, instead of 1 hour, before being required to 
begin shutdown. The impact of the increase in the accumulator CT on 
core damage frequency for all the cases evaluated in WCAP-15049 is 
within the acceptance limit of 1.0E-06/yr for a total plant core 
damage frequency (CDF) less than 1.0E-03/yr. The incremental 
conditional core damage probabilities calculated in WCAP-15049 for 
the accumulator CT increase meet the criterion of 5E-07 in 
Regulatory Guides (RG) 1.174 and 1.177 for all cases except those 
that are based on design basis success criteria. As indicated in 
WCAP-15049, design basis accumulator success criteria are not 
considered necessary to mitigate large break loss-of-coolant 
accident (LOCA) events, and were only included in the WCAP-15049 
evaluation as a worst case data point. In addition, WCAP-15049 
states that the NRC has indicated that an incremental conditional 
core damage frequency (ICCDP) greater than 5E-07 does not 
necessarily mean the change is unacceptable. The proposed technical 
specification change does not involve any hardware changes nor does 
it affect the probability of any event initiators. There will be no 
change to normal plant operating parameters, engineered safety 
feature (ESF) actuation setpoints, accident mitigation capabilities, 
accident analysis assumptions or inputs. 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

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. As described in Section 9.1 of the WCAP-
15049 evaluation, the plant design will not be changed with this 
proposed technical specification CT increase. All safety systems 
still function in the same manner and there is no additional 
reliance on additional systems or procedures. The proposed 
accumulator CT increase has a very small impact on core damage 
frequency. The WCAP-15049 evaluation demonstrates that the small 
increase in risk due to increasing the accumulator allowed outage 
time (AOT) is within the acceptance criteria provided in RGs 1.174 
and 1.177. No new accidents or transients can be introduced with the 
requested change and the likelihood of an accident or transient is 
not impacted. The malfunction of safety related equipment, assumed 
to be operable in the accident analyses, would not be caused as a 
result of the proposed technical specification change. No new 
failure mode has been created and no new equipment performance 
burdens are imposed. Therefore, this 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 the Margin of Safety

    The proposed change does not involve a significant reduction in 
a margin of safety. There will be no change to the departure from 
nucleate boiling ratio (DNBR) correlation limit, the design DNBR 
limits, or the safety analysis DNBR limits. The basis for the 
accumulator LCO, as discussed in Bases Section 3.5.1, is to ensure 
that a sufficient volume of borated water will be immediately forced 
into the core through each of the cold legs in the event the RCS 
pressure falls below the pressure of the accumulators, thereby 
providing the initial cooling mechanism during large RCS pipe 
ruptures. As described in Section 9.2 of the WCAP-15049 evaluation, 
the proposed change will allow plant operation in a configuration 
outside the design basis for up to 24 hours, instead of 1 hour, 
before being required to begin shutdown. The impact of this on plant 
risk was evaluated and found to be very small. That is, increasing 
the time the accumulators will be unavailable to respond to a large 
LOCA event, assuming accumulators are needed to mitigate the design 
basis event, has a very small impact on plant risk. Since the 
frequency of a design basis large LOCA (a

[[Page 53112]]

large LOCA with loss of offsite power) would be significantly lower 
than the large LOCA frequency of the WCAP-15049 evaluation, the 
impact of increasing the accumulator CT from 1 hour to 24 hours on 
plant risk due to a design basis large LOCA would be significantly 
less than the plant risk increase presented in the WCAP-15049 
evaluation. Therefore, this change does not involve a significant 
reduction in a margin of safety.

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: July 6, 2004.
    Description of amendment request: The proposed amendment relocates 
the surveillance requirements for Item 22, ``Accumulator Level and 
Pressure,'' and Item 25, ``Portable Radiation Survey Instruments,'' 
from Table TS 4.1-1 of the Technical Specifications to licensee-
controlled documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?

NMC [Nuclear Management Company] Response for Proposed Change to 
Table TS 4.1-1, Item 22

    No. This TS change removes the accumulator water level and 
pressure channel surveillance from the TS and places them into 
licensee controlled documents. This change is consistent with 
industry and NRC [Nuclear Regulatory Commission] recognition that 
the accumulator instrumentation operability is not directly related 
to the capability of the accumulators to perform their safety 
function.
    Relocating the instrumentation surveillance requirements is an 
administrative change that will not affect equipment testing, 
availability, or operation. Therefore, the change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

NMC Response for Proposed Change to Table TS 4.1-1, Item 25

    No. Removing the surveillance requirements for portable 
radiation survey instruments from the TS is administrative and has 
no impact on plant equipment, accident initiators, or the safety 
analysis. Additionally, eliminating the monthly check and modifying 
the line item description does not impact plant equipment or 
operation. Therefore, the change does not involve an increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?

NMC Response for Proposed Change to Table TS 4.1-1, Item 22

    No. Relocating the accumulator water level and pressure 
instrument surveillance requirements to licensee controlled 
documents is an administrative change that will not change any 
equipment, require new equipment to be installed, or change the way 
current equipment operates in the plant.
    Therefore, the change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

NMC Response for Proposed Change to Table TS 4.1-1, Item 25

    No. Removing the surveillance requirements for portable 
radiation survey instruments from the TS and relocating the 
requirements to licensee controlled documents is administrative and 
has no impact on plant equipment or the way the plant equipment 
operates. Additionally, eliminating the monthly check and modifying 
the line item description does not impact plant equipment or 
operation. Portable radiation survey instruments are not accident 
initiators. Therefore, the change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?

NMC Response for Proposed Change to Table TS 4.1-1, Item 22

    No. Relocating the accumulator water level and pressure 
instrument surveillance requirements to licensee controlled 
documents is an administrative change that will not change the 
safety analyses performed for the plant nor reduce the ability of 
the accumulators to perform their safety related function. There is 
no change in the operation of the accumulators or related equipment 
and systems. Therefore, the change does not involve a reduction in 
the margin of safety.

NMC Response for Proposed Change to Table TS 4.1-1, Item 25

    No. Portable radiation survey instruments are not inputs to the 
safety analysis or to automatic plant actions. The change is 
administrative since it moves the requirements out of TS and into 
licensee controlled documents through use of the 10 CFR 50.36 
selection criteria for TS. Additionally, eliminating the monthly 
check and modifying the line item description does not impact plant 
equipment or operation. Therefore, the change does not reduce 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

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 2, 2004.
    Description of amendment request: The proposed amendment would 
implement a risk-informed process for determining allowed outage times 
for South Texas Project (STP), Units 1 and 2, Technical Specifications 
(TS). The risk-informed process involves the application of the STP, 
Units 1 and 2, Configuration Risk Management Program (CRMP). The STP 
CRMP is a procedurally controlled program utilized for the 
implementation of 50.65(a)(4) of Title 10 of the Code of Federal 
Regulations (10 CFR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change to the Technical Specifications 
involve a significant increase in the probability or consequences of 
an accident previously evaluated?
    The proposed changes to the Technical Specifications to add a 
new TS 3.13.1 and to change specific TS to apply the new TS 3.13.1 
do not involve a significant increase in the probability of an 
accident previously evaluated because the changes involve no change 
to the plant or its modes of operation. In addition, the risk-
informed configuration management program will be applied to 
effectively manage the availability of required systems, structures, 
and components to assure there is no significant increase in the 
probability of an accident. These proposed changes do not increase 
the consequences of an accident because the design-basis mitigation 
function of the affected systems is not changed and the risk-
informed configuration management program will be applied to 
effectively manage the availability of systems, structures and 
components required to mitigate the consequences of an accident. The 
application of the risk-informed configuration management program is 
considered a substantial technological improvement over current 
methods.
    Therefore, none of the proposed changes involve a significant 
increase in the

[[Page 53113]]

probability or consequences of an accident previously evaluated.
    2. Does the proposed change to the Technical Specifications 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    None of the proposed changes involve a new mode of operation or 
design configuration. There are no new or different systems, 
structures, or components proposed by these changes. Therefore, 
there is no possibility of a new or different kind of accident.
    3. Does the proposed change to the Technical Specifications 
involve a significant reduction to a margin of safety?
    Proposed new TS 3.13.1 and the associated changes to the 
specifications that apply the new TS 3.13.1 implement a risk-
informed configuration management program to assure that adequate 
margins of safety are maintained. Application of these new 
specifications and the configuration management program considers 
cumulative effects of multiple systems or components being out of 
service and does so more effectively than the current Technical 
Specifications. Therefore, application of these new specifications 
will not involve a significant reduction in a margin of safety.
    Based on the evaluation above, none of the proposed changes 
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 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: A.H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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 12, 2004.
    Description of amendment request: The proposed changes to the South 
Texas Project (STP), Units 1 and 2, Technical Specifications (TS) for 
steam generators (SGs) are based on draft TS Task Force (TSTF) Improved 
Standard TS Change Traveler TSTF-449, Rev. 2, and the Joseph M. Farley 
Nuclear Plant, Units 1 and 2, submittal dated June 28, 2004, as 
supplemented by letter dated August 5, 2004. The changes would 
implement guidance for the industry initiative on Nuclear Energy 
Institute (NEI) 97-06, ``Steam Generator Program Guidelines.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change requires a Steam Generator Program that 
includes performance criteria that will provide reasonable assurance 
that the SG tubing will retain integrity over the full range of 
operating conditions (including startup, operation in the power 
range, hot standby, cooldown, and all anticipated transients 
included in the design specification). The SG performance criteria 
are based on tube structural integrity, accident induced leakage, 
and operational leakage.
    The structural integrity performance criterion is:
    All inservice SG tubes shall retain structural integrity over 
the full range of normal operating conditions (including startup, 
operation in the power range, hot standby, and cooldown, and all 
anticipated transients included in the design specification) and 
design basis accidents. This includes retaining a safety factor of 
3.0 (3 [delta] P) against burst under normal steady state full power 
operation primary-to-secondary pressure differential and a safety 
factor of 1.4 against burst applied to the design basis accident 
primary-to-secondary pressure differentials. Apart from the above 
requirements, additional loading conditions associated with the 
design basis accidents, or combination of accidents in accordance 
with the design and licensing basis, shall also be evaluated to 
determine if the associated loads contribute significantly to burst 
or collapse. In the assessment of tube integrity, those loads that 
do significantly affect burst or collapse shall be determined and 
assessed in combination with the loads due to pressure with a safety 
factor of 1.2 on the combined primary loads and 1.0 on axial 
secondary loads.
    The accident induced leakage performance criterion is:
    The primary-to-secondary accident induced leakage rate for any 
design basis accidents, other than a SG tube rupture, shall not 
exceed the leakage rate assumed in the accident analysis in terms of 
total leakage rate for all SGs and leakage rate for an individual 
SG. Accident induced leakage is not to exceed 1 gpm [gallons per 
minute] total for all four SGs in a unit.
    The operational leakage performance criterion is:
    ``The RCS operational primary-to-secondary leakage through any 
one SG shall be limited to 150 gallons per day.''
    An SGTR [steam generator tube rupture] event is one of the 
design basis accidents analyzed as part of the plant licensing 
basis. In the analysis of an SGTR event, a bounding primary-to-
secondary leakage rate equal to the operational leakage rate limits 
in the licensing basis plus the leakage rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as MSLB [main steamline 
break], rod ejection, and reactor coolant pump locked rotor, the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). At STP these analyses assume that the 
total primary-to-secondary leakage is 1 gpm. The accident induced 
leakage criterion introduced by the proposed changes accounts for 
tubes that may leak during design basis accidents. The accident 
induced leakage criterion limits this leakage to no more than the 
value assumed in the accident analysis.
    The SG performance criteria proposed in this change to the TS 
identify the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining RCPB [reactor coolant 
pressure boundary] integrity throughout each operating cycle and in 
the unlikely event of a design basis accident. The performance 
criteria are only a part of the Steam Generator Program required by 
the proposed change to the TS. The program, defined by NEI 97-06, 
includes a framework that incorporates a balance of prevention, 
inspection, evaluation, repair, and leakage monitoring.
    The consequences of design basis accidents are, in part, 
functions of the dose equivalent I-131 in the primary coolant and 
the primary-to-secondary leakage rates resulting from an accident. 
Therefore, limits are included in the TS for operational leakage and 
for dose equivalent I-131 in primary coolant to ensure the plant is 
operated within its analyzed condition. The analysis of the limiting 
design basis accident assumes that primary-to-secondary leak rate 
after the accident is 1 gpm with no more than 500 gpd [gallons per 
day] in any one SG, and that the reactor coolant activity levels of 
dose equivalent I-131 are at the TS values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TS and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TS.
    Therefore, the proposed change does not affect the consequences 
of an SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed performance-based requirements are an improvement 
over the requirements imposed by the current TS.
    Implementation of the proposed Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the Steam Generator Program will be 
an enhancement of SG tube performance. Primary-to-secondary leakage

[[Page 53114]]

that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes are an integral part of the RCPB and, as such, are 
relied upon to maintain the primary system pressure and inventory. 
As part of the RCPB, the SG tubes are unique in that they are also 
relied upon as a heat transfer surface between the primary and 
secondary systems such that residual heat can be removed from the 
primary system. In addition, the SG tubes also isolate the 
radioactive fission products in the primary coolant from the 
secondary system. In summary, the safety function of a SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in tube 
integrity by implementing the Steam Generator Program to manage SG 
tube inspection, assessment, repair, and plugging. The requirements 
established by the Steam Generator Program are consistent with those 
in the applicable design codes and standards and are an improvement 
over the requirements in the current TS.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: A.H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: August 5, 2004.
    Brief description of amendments: The proposed change revises 
Technical Specification 3.7.10 entitled, ``Control Room Emergency 
Filtration/Pressurization System (CREFS),'' to add a new condition for 
an inoperable Control Room boundary.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This is a revision to the Technical Specifications for the 
Control Room Emergency/Filtration System which is a mitigation 
system designed to minimize in leakage and to filter the control 
room atmosphere to protect the operator following accidents 
previously analyzed. An important part of the system is the Control 
Room boundary. The Control Room boundary integrity is not an 
initiator or precursor to any accident previously evaluated. 
Therefore, the probability of any accident previously evaluated is 
not increased. The analysis of the consequences of analyzed accident 
scenarios under the control room breach conditions along with the 
compensatory actions for restoration of control room integrity 
demonstrate that the consequences of any accident previously 
evaluated are not increased. Therefore, it is concluded that this 
change does not significantly increase 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 change will not impact the accident analysis. The 
change will not alter the requirements of the Control Room 
Emergency/Filtration System or its function during accident 
conditions. The administrative controls and compensatory actions 
will ensure the control room emergency/filtration system will 
perform its safety function. No new or different accidents result 
from performing the new actions and surveillance required. The 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or a change in 
the methods governing normal plant operation. The change does not 
alter assumptions made in the safety analysis. The proposed change 
is consistent with the safety analysis assumptions and current plant 
operating practice. Therefore, the proposed change does 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 change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by these changes. The proposed change will not 
result in plant operation in a configuration outside the design 
basis for an unacceptable period of time without compensatory 
actions and administrative controls. The proposed change does not 
affect systems that respond to safely shutdown the plant and to 
maintain the plant in a safe shutdown condition. 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: July 1, 2004.
    Description of amendment request: The proposed license amendments 
would modify the Reactor Coolant System (RCS) pressure/temperature (P/
T) limit curves, the Low-Temperature Overpressure Protection System 
(LTOPS) setpoint allowable values, and the LTOPS Tenable values. In 
addition, the cumulative core burnup applicability limits for the LTOPS 
would be extended.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes modify the North Anna Units 1 and 2 RCS P/T 
limit curves, LTOPS setpoint allowable values, LTOPS Tenable and 
extend the cumulative core burnup applicability limits for the 
LTOPS. The allowable operating pressures and temperatures under the 
proposed RCS P/T limit curves are not significantly different from 
those allowed under the existing Technical Specification P/T limits. 
The revisions in the values for the LTOPS setpoint allowable values 
and LTOPS Tenable values do not significantly change the plant 
operating space. No changes to plant systems, structures or 
components are proposed, and no new operating modes are established. 
The P/T limits, LTOPS setpoint allowable values, and Tenable values 
do not contribute to the probability of occurrence or consequences 
of accidents previously analyzed. The revised licensing basis

[[Page 53115]]

analyses utilize acceptable analytical methods, and continue to 
demonstrate that established accident analysis acceptance criteria 
are met. Therefore, there is no increase in the probability or 
consequences of any accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes modify the North Anna Units 1 and 2 RCS P/T 
limit curves, LTOPS setpoint allowable values, LTOPS Tenable values 
and extend the cumulative core burnup applicability limits for the 
LTOPS. The allowable operating pressures and temperatures under the 
proposed RCS P/T limit curves are not significantly different from 
those allowed under the existing Technical Specification P/T limits. 
No changes to plant systems, structures or components are proposed, 
and no new operating modes are established. Therefore, the proposed 
changes do not create the possibility of any accident or malfunction 
of a different type previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety?
    The proposed revised RCS P/T limit curves, LTOPS setpoint 
allowable values, and LTOPS Tenable analysis bases do not involve a 
significant reduction in the margin of safety for these parameters. 
The effects of RCS pressure and temperature measurement uncertainty 
continue to be considered in the supporting analyses. The proposed 
revised RCS P/T limit curves are valid to cumulative core burnups of 
50.3 EFPY [effective full-power year] and 52.3 EFPY for North Anna 
Units 1 and 2 respectively. The proposed revised LTOPS setpoint 
allowable values and Tenable analyses support these same cumulative 
core burnup limits. The analyses demonstrate that established 
analysis acceptance criteria continue to be met. Specifically, the 
proposed P/T limit curves, LTOPS setpoint allowable values and LTOPS 
Tenable values provide acceptable margin to vessel fracture under 
both normal operation and LTOPS design basis (mass addition and heat 
addition) accident conditions. Therefore, the proposed changes do 
not result in a significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: Mary Jane Ross-Lee (Acting).

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

    Date of amendment request: July 22, 2004.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) Figure 3.5.5-1, ``Seal Injection Flow 
Limits,'' to reflect flow limits that allow a higher seal injection 
flow for a given differential pressure between the charging discharge 
header and the reactor coolant system pressure. Specifically, the 
licensee requests approval of the proposed amendment to allow for 
repositioning the seal injection throttle valves during the upcoming 
refueling outage.
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The restriction on reactor coolant pump (RCP) seal injection 
flow limits the amount of Emergency Core Cooling System (ECCS) flow 
that would be diverted from the injection path following an 
accident. This limit is based on safety analysis assumptions that 
are required because RCP seal injection flow is not isolated during 
safety injection. The intent of the Limiting Condition for Operation 
(LCO) limit on seal injection flow is to make sure that flow through 
the RCP seal water injection line is low enough to ensure sufficient 
centrifugal charging pump injection flow is directed to the Reactor 
Coolant System (RCS) via the injection points.
    There are no hardware changes nor are there any changes in the 
method by which any safety related plant system performs its safety 
function. The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
[the] plant is operated and maintained. The proposed change does not 
alter or prevent the ability of structures, systems, and components 
from performing their intended safety function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change does not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. Further, the proposed change does not increase the types 
or amounts of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed change is consistent with the 
safety analysis assumptions and resultant consequences.
    Since the change continues to ensure 100 percent of the assumed 
charging flow is available, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes nor are there any changes in the 
method by which any safety related plant system performs its safety 
function. This amendment will not affect the normal method of plant 
operation. The proposed change does not introduce any new equipment 
into the plant or alter the manner in which existing equipment will 
be operated. No performance requirements or response time limits 
will be affected. The change is consistent with assumptions made in 
the safety analysis and licensing basis regarding limits on RCP seal 
injection flow.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. The[re] will be no adverse effect or challenges 
imposed on any safety related system as a result of this amendment.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the acceptance criteria for 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection function. Increasing the total seal 
injection flow limit to 90 gpm does not significantly impact the 
assumed ECCS flow that would be available for injection into the RCS 
following an accident.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: July 23, 2004.
    Description of amendment request: The proposed amendment will 
delete the requirements from the technical

[[Page 53116]]

specifications (TS) to maintain hydrogen recombiners and hydrogen 
monitors. 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 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 proposed license amendment will revise TS 3.3.3, ``Post 
Accident Monitoring (PAM) Instrumentation,'' to delete the Note in 
Condition C. Also in TS 3.3.3, Condition D will be deleted. In TS Table 
3.3.3-1, Function 10, ``Containment Hydrogen Concentration Level,'' is 
deleted and replaced with ``Not Used.'' TS 3.6.8, ``Hydrogen 
Recombiners,'' will be deleted and the Table of Contents will be 
revised to reflect that deletion. TS 5.6.8, ``PAM Report,'' will be 
revised to reflect changing Condition G to Condition F in TS 3.3.3.
    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 July 23, 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. 
RG 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.

    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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: July 23, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.'' The Table of

[[Page 53117]]

Contents will also be revised to reflect the deletions.
    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 July 23, 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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating letter 
report of shutdown experience and operating statistics if the 
equivalent data is submitted using an industry electronic database. 
It also eliminates the TS 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.
    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 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.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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 significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, 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 email to [email protected].

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: March 23, 2004.
    Brief description of amendment: The amendment eliminates the 
Technical Specification requirements related to hydrogen monitors.
    Date of Issuance: August 9, 2004.
    Effective date: August 9, 2004 and shall be implemented within 60 
days of issuance.
    Amendment No.: 246.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 27, 2004 (69 FR 
22879).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 9, 2004.
    No significant hazards consideration comments received: No.

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

DukeEnergy Corporation, Docket Nos.50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: March 23, 2004.
    Brief description of amendments: The amendments revise the reactor 
coolant pump flywheel inspection interval from 10 years to 20 years.
    Date of issuance: August 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 216 and 210, 223 and 205.
    Renewed facility operating license Nos. NPF-35, NPF-52, NPF-9, And 
NPF-17: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 2004.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 5, 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: August 22, 2002, as supplemented 
by letters dated September 12, 2003, and February 4, February 16, March 
23, April 28, June 17, July 6, July 12, July 19, and July 29, 2004.

[[Page 53118]]

    Brief description of amendments: The amendments revised Technical 
Specification 3.8.1, ``AC Sources--Operating,'' to temporarily extend 
the Completion Times (CTs) for the Keowee hydro units (KHUs) to allow 
additional time for maintenance and upgrades. The amendments extend by 
17 days (from 45 days to 62 days) the CT when one KHU is not operable 
and extend by 120 hours (from 60 hours to 180 hours) the CT when both 
KHUs are not operable.
    Date of Issuance: August 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 339, 341, and 340.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58641).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 5, 2004.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: October 21, 2003.
    Brief description of amendment: The change removes MODE 
restrictions that prevent performance of Surveillance Requirements 
(SRs) 3.8.4.7 and 3.8.4.8 for the Division III direct current 
electrical power subsystem while in MODES 1, 2, or 3. These 
surveillances verify that the battery capacity is adequate to perform 
its required functions. The changes allow the performance of SR 3.8.4.7 
and SR 3.8.4.8 during normal plant operations rather than only during 
refueling outages.
    Date of issuance: August 12, 2004.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 141.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specfications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68662).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 12, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 9, 2004.
    Brief description of amendment: The amendment extends the 
completion time (CT) from 1 hour to 24 hours for Condition B of 
Technical Specification (TS) 3.5.1, which defines requirements for the 
emergency core cooling system accumulators. Condition B of TS 3.5.1 
specifies a CT to restore an accumulator to operable status when it has 
been declared inoperable for a reason other than the boron 
concentration of the water in the accumulator not being within the 
required range.
    Date of issuance: August 18, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 222.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19567).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 18, 2004.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: August 16, 2002, as supplemented 
March 25, 2003, April 6, and July 22, 2004.
    Brief description of amendment: This amendment deleted the existing 
requirements in Technical Specification (TS) 3.10.D.1.d from TS 3/
4.10.D, ``Multiple Control Rod Removal,'' and the associated 
Surveillance Requirement 4.10.D.1.d. This amendment added a new 
requirement to TS 3.10.D.1.d. Additionally, this amendment made an 
editorial change to correct a reference to TS 3.3.B.3 instead of TS 
3.3.B.4 in TS 3/4.10.D.1.
    Date of issuance: August 17, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 207.
    Facility Operating License No. DPR-35: Amendment revised the TSs.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75873).
    The supplements dated March 25, 2003, April 6, and July 22, 2004, 
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 Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 17, 2004.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: February 9, 2004.
    Brief description of amendment: The amendment eliminates the 
requirements in the Technical Specifications associated with hydrogen 
recombiners and hydrogen monitors.
    Date of issuance: August 12, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment No.: 222.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 2004 (69 FR 
16617).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 12, 2004.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: February 9, 2004.
    Brief description of amendment: The amendment eliminates the 
requirements in the Technical Specifications associated with hydrogen 
recombiners and hydrogen monitors.
    Date of issuance: August 5, 2004.
    Effective date: As of the date of issuance to be implemented within 
120 days from the date of issuance.
    Amendment No.: 254.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 2004 (69 FR 
16618).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 5, 2004.
    No significant hazards consideration comments received: No.

[[Page 53119]]

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

    Date of application for amendments: August 27, 2003.
    Brief description of amendments: The amendments change Technical 
Specification 4.0.3, ``Missed Surveillance Time Allowance.'' TS 4.0.3 
describes the relationship between meeting the surveillance requirement 
and operability. The amendments modify TS 4.0.3 to allow a missed 
surveillance to be completed within 24 hours or up to the limit of the 
specified interval, whichever is greater. Additionally, the amendments 
add a statement that a risk evaluation shall be performed for any 
surveillance delayed greater than 24 hours and that the risk impact 
shall be managed. The amendments also change the Bases to further 
clarify the provisions of the TS. In addition, the proposed amendments 
make format changes to improve appearance. The changes to the TS and 
its Bases are consistent with industry/Technical Specification Task 
Force TSTF-358, Revision 6, which was approved by the Nuclear 
Regulatory Commission (NRC) on October 3, 2001, and incorporated the 
NRC's comments on TSTF-358, Revision 5. TSTF-358, Revision 5, was 
approved with comment by the NRC as a part of the Consolidated Line 
Item Improvement Process in a Federal Register Notice dated September 
28, 2001.
    Date of issuance: August 9, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 282, 266.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 11, 2004 (69 FR 
26190).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 9, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 30, 2004.
    Brief description of amendments: The amendments relocate the 
requirements for hydrogen monitors to the Technical Requirements 
Manual.
    Date of issuance: August 13, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 214 and 219.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9862).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 13, 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: March 25, 2004, as supplemented 
June 2, 2004.
    Brief description of amendments: The amendments approve a change to 
the licensing basis to allow the use of the methods described in 
Framatome-ANP Topical Report BAW-10169-A, ``RSG Plant Safety Analysis--
B&W Safety Analysis Methodology for Recirculating Steam Generator 
Plants,'' dated October 1989, for calculating the mass and energy 
release rates resulting from a postulated main steamline break accident 
for input to containment analyses. These methods utilize the RELAP5/
MOD2-B&W code approved by the Nuclear Regulatory Commission staff in a 
safety evaluation report dated March 14, 1995.
    Date of issuance: August 19, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 164 and 155.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
authorized revision to the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: April 27, 2004 (69 FR 
22881).
    The June 2, 2004, supplemental letter contained clarifying 
information and did not change the initial proposed no significant 
hazards consideration determination and was within the scope of the 
original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a safety evaluation dated August 19, 2004.
    No significant hazards consideration comments received: No.

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 application for amendment: July 23, 2003.
    Brief description of amendment: Revised the near end-of-life 
Moderator Temperature Coefficient (MTC) Surveillance Requirement 
4.1.1.3.b by placing a set of conditions on core operation, which if 
met, would allow exemption from the required MTC measurement. The 
conditional exemption is determined on a cycle-specific basis by 
considering the margin predicted to the surveillance requirement MTC 
limit and the performance of other core parameters, such as beginning 
of life MTC measurements and the critical boron concentration as a 
function of cycle life.
    Date of issuance: July 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 169.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: September 30, 2003 (68 
FR 56346).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 21, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 30, 2003.
    Brief description of amendments: The amendments revised the staff 
position titles in Section 5.0 ``Administrative Controls'' of the 
Technical Specifications.
    Date of issuance: June 3, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 242 and 185.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9865).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 3, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 23rd day of August 2004.


[[Page 53120]]


    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Division of Licensing Project Management, Office of Nuclear Reactor 
Regulation.
Director,
[FR Doc. 04-19586 Filed 8-30-04; 8:45 am]
BILLING CODE 7590-01-P