[Federal Register Volume 69, Number 71 (Tuesday, April 13, 2004)]
[Notices]
[Pages 19561-19582]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-8047]


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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, March 19 through April 1, 2004. The last 
biweekly notice was published on March 30, 2004 (69 FR 16615).

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

    The Commission has made a proposed determination that the following 
amendment requests involve

[[Page 19562]]

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 fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the

[[Page 19563]]

Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; (2) courier, express mail, and expedited delivery 
services: Office of the Secretary, Sixteenth Floor, One White Flint 
North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: 
Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office 
of the Secretary, U.S. Nuclear Regulatory Commission, 
hearingdocket@nrc.gov; or (4) facsimile transmission addressed to the 
Office of the Secretary, U.S. Nuclear Regulatory Commission, 
Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 
415-1101, verification number is (301) 415-1966. A copy of the request 
for hearing and petition for leave to intervene should also be sent to 
the Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and it is requested that copies be 
transmitted either by means of facsimile transmission to 301-415-3725 
or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
attorney for the licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
pdr@nrc.gov.

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

    Date of amendment request: February 27, 2004.
    Description of amendment request: The licensee proposed to relocate 
the average power range monitor (APRM)-based stability protection 
settings for Option II stability solution to the Core Operating Limits 
Report (COLR). The Option II solution demonstrates that existing 
quadrant-based APRM trip systems will initiate a reactor scram for 
postulated reactor instability and avoid violating the minimum critical 
power ratio safety limit. Use of Option II was previously approved by 
the Nuclear Regulatory Commission staff thru Amendment No. 235, dated 
October 18, 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes will relocate the Average Power Range 
Monitor (APRM) based stability protection settings for the Option II 
stability solution from the Technical Specifications (TS) to the 
Core Operating Limits Report (COLR). The APRM based stability 
protection settings are not an initiator or a precursor to an 
accident. Furthermore, changes to the stability protection settings 
do not physically modify or change the function, or system 
interfaces, of the APRM Neutron Flux Scram and Neutron Flux Control 
Rod Block systems or components. The APRM based stability protection 
settings provide automatic protection to assure that anticipated 
coupled neutronic/thermal-hydraulic instabilities will not 
compromise established fuel safety limits. The proposed TS changes 
cannot increase the consequences of a previously evaluated accident 
because the changes do not alter any Limiting Safety System Setting, 
but only relocate the applicable stability protection settings to 
the COLR. The applicable stability protection settings will continue 
to be determined by an NRC approved methodology.
    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 proposed changes will relocate the APRM based stability 
protection settings for the Option II stability solution from the TS 
to the COLR. The APRM based stability protection settings for the 
Option II stability solution assure anticipated coupled neutronic/
thermal-hydraulic instabilities will not compromise established fuel 
safety limits. These changes do not introduce any new accident 
precursors and do not involve any alterations to plant 
configurations which could initiate a new or different kind of 
accident. The proposed changes do not affect the intended function 
of the APRM system nor do they affect the operation of the system in 
a way which would create a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will relocate the APRM based stability 
protection settings for the Option II stability solution from the TS 
to the COLR. The APRM based stability protection settings for 
protection against reactor instability assure anticipated coupled 
neutronic/thermal-hydraulic instabilities will not compromise 
established fuel safety limits. No fuel thermal limits or other 
design and licensing basis acceptance criteria are adversely 
affected. No other events are adversely affected. The margin of 
safety, as defined in the TS, for all events is maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: March 8, 2004.
    Description of amendment request: The proposed amendment would 
delete Operating License Condition 2.C.(6) ``Long Range Planning 
Program.'' The original objective of this requirement was to enable the 
licensee to better control and manage resources regarding major 
activities. The license condition does not have any direct effect on 
plant design or operation. Since imposition of this requirement on May 
27, 1988, the licensee has developed internal processes to control and 
manage work activities, thus leading the licensee to determine that 
this license condition is no longer needed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 19564]]

issue of no significant hazards consideration. The NRC staff has 
reviewed the licensee's analysis against the three standards of 10 CFR 
50.92(c). 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 subject license condition was not a factor in the 
scenario of any previously analyzed postulated design-basis accident or 
anticipated operational transient. No hardware design change is 
involved with the proposed amendment. Thus, the proposed deletion of 
the license condition would create no adverse effect on the functional 
performance of any plant structure, system, or component (SSC). All 
SSCs will continue to perform their design functions with no decrease 
in their capabilities to mitigate the previously analyzed consequences 
of postulated accidents and anticipated operational transients. 
Accordingly, the deletion of the license condition will lead to no 
increase in the consequences of an accident previously evaluated, and 
no increase in 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. The proposed amendment is not the result of a hardware 
design change, nor does it lead to the need for a hardware design 
change. There is no change in the methods the unit is operated. As a 
result, all SSCs will continue to perform as previously analyzed by the 
licensee, and previously evaluated and accepted by the NRC staff. 
Therefore, the proposed amendment will not create the possibility of a 
new or different kind of accident from any 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 proposed deletion of the 
license condition will not lead the licensee to exceed or alter a 
design basis or safety limit, and will not result in operating any 
component in a less conservative manner, 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 amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: March 8, 2004.
    Description of amendment request: The proposed amendment would 
delete Operating License Condition 2.C.(9) ``Long Range Planning 
Program.'' The original objective of this requirement was to enable the 
licensee to better control and manage resources regarding major 
activities. The license condition does not have any direct effect on 
plant design or operation. Since imposition of this requirement on May 
27, 1988, the licensee has developed internal processes to control and 
manage work activities, thus leading the licensee to determine that 
this license condition is no longer needed.
    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 has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). 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 subject license condition was not a factor in the 
scenario of any previously analyzed postulated design-basis accident or 
anticipated operational transient. No hardware design change is 
involved with the proposed amendment. Thus, the proposed deletion of 
the license condition would create no adverse effect on the functional 
performance of any plant structure, system, or component (SSC). All 
SSCs will continue to perform their design functions with no decrease 
in their capabilities to mitigate the previously analyzed consequences 
of postulated accidents and anticipated operational transients. 
Accordingly, the deletion of the license condition will lead to no 
increase in the consequences of an accident previously evaluated, and 
no increase in 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. The proposed amendment is not the result of a hardware 
design change, nor does it lead to the need for a hardware design 
change. There is no change in the methods the unit is operated. As a 
result, all SSCs will continue to perform as previously analyzed by the 
licensee, and previously evaluated and accepted by the NRC staff. 
Therefore, the proposed amendment will not create the possibility of a 
new or different kind of accident from any 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 proposed deletion of the 
license condition will not lead the licensee to exceed or alter a 
design basis or safety limit, and will not result in operating any 
component in a less conservative manner, 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 amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Edward J. Cullen, Jr., Esquire, Vice 
President, General Counsel and Secretary, Exelon Generation Company, 
LLC, 300 Exelon Way, Kennett Square, PA 19348.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendments request: December 12, 2003.
    Description of amendments request: The proposed amendment would 
delete Technical Specification (TS) Section 5.5.3, ``Post-Accident 
Sampling,'' requirements to maintain a Post-Accident Sampling System 
(PASS). Licensees were generally required to implement PASS upgrades as 
a result of 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

[[Page 19565]]

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 TS 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 August 11, 2000 (65 FR 49271) 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 October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated December 12, 2003.
    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 radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the 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 [a] 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: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: June 25, 2003.
    Description of amendment request: The proposed amendments would 
correct two inadvertent editorial changes made by Duke during the 
submittal of Technical Specification (TS) Amendment 194/175 which 
revised TS 3.3.1 (Reactor Trip System Instrumentation) and TS Amendment 
197/178 which revised TS 4.2.1 (Design Features, Fuel Assemblies).
    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 this LAR [License Amendment Request] involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    No. Approval and implementation of this LAR will have no affect 
on accident probabilities or consequences since the proposed changes 
are editorial in nature and were previously reviewed and approved by 
the NRC [Nuclear Regulatory Commission].
    2. Does this LAR create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. This LAR does not involve any physical changes to the plant. 
Therefore, no new accident causal mechanisms will be generated. The 
proposed changes are editorial in nature and were previously 
reviewed and approved by the NRC. Consequently, plant accident 
analyses will not be affected by these changes.
    3. Does this LAR involve a significant reduction in a margin of 
safety?
    No. Margin of safety is related to the confidence in the ability 
of the fission

[[Page 19566]]

product barriers to perform their design functions during and 
following accident conditions. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The performance of these barriers will not be affected by the 
proposed changes since they are editorial in nature and have been 
previously reviewed and approved by the NRC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: January 15, 2004, as supplemented by 
letter dated March 15, 2004.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications associated with the control rod 
drive (CRD) trip devices. These amendments are needed to support 
implementation of the reactor trip breaker (RTB) replacement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated[.]
    The proposed LAR [license amendment request] modifies the 
Technical Specifications [TS] to incorporate new TS requirements 
associated with the new Control Rod Drive (CRD)/Reactor Trip Breaker 
(RTB) configuration. The proposed LAR will continue to ensure that 
the CRD trip devices will be operable to ensure that the reactor 
remains capable of being tripped at any time it is critical. 
Reliable CRD reactor trip circuit breakers and associated support 
circuitry provides assurance that a reactor trip will occur when 
initiated. The new RTBs will have the same seismic and quality group 
qualifications as the existing components in the CRDCS [CRD control 
system] system [sic]. The new RTBs will enhance the reliability of 
the system by resolving age-related degradation issues and replacing 
obsolete equipment. Therefore, the proposed LAR does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated[.]
    The proposed LAR modifies the Technical Specifications to 
incorporate new TS requirements associated with the new CRD/RTB 
configuration. The systems affected by implementing the proposed 
changes to the TS are not assumed to initiate design basis 
accidents. Rather, the systems affected by the changes are used to 
mitigate the consequences of an accident that has already occurred. 
The proposed TS changes do not affect the mitigating function of 
these systems. The reliability of the mitigating systems will be 
improved by implementation of the RTB Upgrade. Consequently, these 
changes do not alter the nature of events postulated in the Safety 
Analysis Report nor do they introduce any unique precursor 
mechanisms. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed TS changes do not unfavorably affect any plant 
safety limits, set points, or design parameters. The changes also do 
not unfavorably affect the fuel, fuel cladding, RCS [reactor coolant 
system], or containment integrity. Therefore, the proposed TS 
change, which adds TS requirements associated with the CRD/RTB 
upgrade, 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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point 
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

    Date of amendment request: March 3, 2004.
    Description of amendment request: The proposed amendments would 
revise the administrative Technical Specifications (TSs) for the 
Reactor Coolant Pump Flywheel Inspection Program to extend the 
allowable inspection interval to 20 years.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 24, 2003 (68 FR 37590), on possible amendments 
to extend the inspection interval for reactor coolant pump (RCP) 
flywheels, 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 October 22, 2003 (68 
FR 60422). The licensee affirmed the applicability of the model NSHC 
determination in its application dated March 3, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    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; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or 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 type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

[[Page 19567]]

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 in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any 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 accident previously evaluated.

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

    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 impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: March 9, 2004.
    Description of amendment request: The proposed amendment would 
extend the completion time (CT) from 1 hour to 24 hours for Condition B 
of Technical Specification (TS) 3.5.1, ``Accumulators.'' The 
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. 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. This change was proposed by the Westinghouse Owners 
Group participants in the TS Task Force (TSTF) and is designated TSTF-
370. 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 March 9, 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 Bases Section 3.5.1.1, 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 WCAP-15049-A, the proposed change will 
allow plant operation with an inoperable accumulator for up to 24 
hours, instead of 1 hour, before the plant would be required to 
begin shutting down. The impact of the increase in the accumulator 
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A 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-A for the accumulator CT increase meet the criterion of 
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using 
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications,'' for all cases except those that are based on 
design basis success criteria. As indicated in WCAP-15049-A, 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-A evaluation as a worst case 
data point. In addition, WCAP-15049-A 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-A 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-A evaluation demonstrates that the small 
increase in risk due to increasing the CT for an inoperable 
accumulator 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 a 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 
3.5.1.1, is to ensure that a sufficient volume of borated

[[Page 19568]]

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 WCAP-
15049-A, the proposed change will allow plant operation with an 
inoperable accumulator for up to 24 hours, instead of 1 hour, before 
the plant would be required to begin shutting down. 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 large LOCA 
with loss of offsite power) would be significantly lower than the 
large LOCA frequency of the WCAP-15049-A 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-A evaluation.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: December 24, 2003.
    Description of amendment request: The proposed amendment would 
delete requirements in the Pilgrim Nuclear Power Station Technical 
Specifications (TSs) 3.7.A.7.c and 4.7.A.7.c, associated with hydrogen 
analyzers. The NRC staff issued a notice of opportunity for comment in 
the Federal Register on August 2, 2002 (67 FR 50374), on possible 
amendments to eliminate the hydrogen analyzers from TSs, including a 
model safety evaluation and model no significant hazards consideration 
(NSHC) determination, using the Consolidated Line Item Improvement 
Process (CLIIP). The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 25, 2003 (68 FR 
55416). The licensee affirmed the applicability of the relevant 
portions of the model NSHC determination in its application dated 
December 24, 2003.
    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 [Regulatory Guide] 1.97 Category 1, is intended for key variables 
that most directly indicate the accomplishment of a safety function 
for design-basis accident events. The hydrogen monitors no longer 
meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 3, as defined in RG 1.97, is an appropriate categorization 
for the hydrogen monitors because the monitors are required to 
diagnose the course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the SAMGs [Severe Accident 
Management Guidelines], the emergency plan (EP), the emergency 
operating procedures (EOP), and site survey monitoring that support 
modification of emergency plan protective action recommendations 
(PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

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

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

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

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

    Based upon the reasoning presented above, the requested change does 
not involve a significant hazards consideration. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: J.M. Fulton, Esquire, Assistant General 
Counsel,

[[Page 19569]]

Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: Darrell J. Roberts, Acting.

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

    Date of amendment request: March 15, 2004.
    Description of amendment request: The amendment proposes to move 
the Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical 
Specification (TS) 3.4.8.2, pressurizer heatup and cooldown limits to 
the Technical Requirements Manual (TRM), which is reviewed in 
accordance with Section 50.59 of Title 10 of the Code of Federal 
Regulations (10 CFR), ``Changes, tests, and experiments.'' The 
associated action statement, surveillance requirement, and bases are 
also proposed for relocation.
    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 probability of an accident is unchanged as a result of the 
proposed change to delete the Waterford 3 pressurizer heatup and 
cooldown rates and associated action, surveillance requirement, and 
bases from the TS. The cooldown and heatup rates are not initiators 
to any accidents or pressurizer transients discussed in the 
Waterford 3 Final Safety Analysis Report (FSAR). Therefore, the 
probability of an accident is not changed.
    The purpose of the pressurizer heatup and cooldown limits is to 
ensure that given transient events will not negatively affect the 
pressurizer structural integrity beyond Code allowables. These 
limits will be maintained within ASME [American Society of 
Mechanical Engineers] Code allowables in the TRM in accordance with 
10 CFR 50.59. Therefore, the consequences of an accident are not 
increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The limitations imposed on the pressurizer heatup and cooldown 
rates are provided to assure that the pressurizer is operated within 
the design criteria assumed for the flaw evaluation and fatigue 
analysis performed in accordance with the ASME Code Section XI, 
subsection IWB-3600 requirements. The Waterford 3 FSAR has analyzed 
the conditions that would result from a thermal or pressurization 
transient on the Waterford 3 pressurizer. The proposed deletion of 
the pressurizer heatup and cooldown rates and relocation of the 
limits to the TRM does not change the way that the pressurizer is 
designed or operated.
    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 margin of safety is established by the rules contained in 
the ASME Section III Code. Any future changes to the cooldown or 
heatup rates will be evaluated using 10 CFR 50.59 and are required 
to meet the ASME Code margins.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: March 12, 2004.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications (TS) to eliminate selected response 
time testing (RTT) requirements associated with Reactor Protection 
System instrumentation and Primary Containment Isolation 
instrumentation for Main Steam Line Isolation functions. The proposed 
changes are consistent with the Boiling Water Reactor Owners Group 
(BWROG) Licensing Topical Report ``System Analyses for the Elimination 
of Selected Response Time Testing Requirements,'' NEDO-32291A, 
Supplement 1, dated October 1999, as approved by the NRC on June 11, 
1999.
    The original Licensing Topical Report (LTR) NEDO-32291-A, dated 
October 1995, established a generic basis for elimination of many RTTs 
for instrument loops that had good performance histories and longer 
response time requirements. The justification was based on the adequacy 
of surveillance tests other than RTTs to assure that response time 
requirements were met for sensors in those loops. Supplement 1 to NEDO-
32291-A was prepared to document an analysis to extend the conclusions 
of the original study to cover the logic components in selected 
instrumentation loops that have intermediate length response time 
requirements. The intent was to demonstrate that elimination of the RTT 
requirements for the logic portions of those loops is of no safety 
significance. Supplement 1 concludes, for instrument loops meeting the 
application criteria of the Licensing Topical Report, that performance 
of ongoing TS required surveillance tests other than RTTs (i.e., 
calibration tests, functional tests, and logic system functional tests) 
provides adequate assurance that those instrument loops will meet their 
respective response time requirements.
    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 probability or consequences of an accident previously evaluated.
    The proposed amendment to the TS eliminates selected RTT 
requirements in accordance with the NRC approved BWROG LTR. 
Elimination of RTT for selected instrumentation in the Reactor 
Protection System and Primary Containment Isolation Instrumentation 
does not result in the alteration of the design, material, or 
construction standards that were applicable prior to the proposed 
change. The response time assumptions used in the accident analyses 
remain unchanged. Only the methodology used for response time 
verification is changed. All component models used in the affected 
trip channels were analyzed for a bounding response time. As 
documented in the BWROG LTR and supplement, a degraded response time 
will be detected by other TS required tests. The bounding response 
time of the relays discussed in the supplement to the LTR can be 
used in place of actual measured response times to ensure that the 
instrumentation systems will meet the response time requirements of 
the accident analysis.
    The proposed change will not result in the modification of any 
system interface that would increase the likelihood of an accident 
since these events are independent of the proposed change. In 
addition, the proposed amendment will not change, degrade, or 
prevent actions, or alter any assumptions previously made in 
evaluating the radiological consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 19570]]

    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed action does not involve physical alteration of the 
station. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
is no change being made to the parameters within which LaSalle is 
operated. There are no setpoints at which protective or mitigative 
actions are initiated that are affected by this proposed action. All 
Reactor Protection System and Primary Containment Isolation 
Instrumentation channels affected by the proposed change will 
continue to have an initial response time verified by test before 
initially placing the channel in service and after any maintenance 
that could affect response time.
    The proposed change does not alter assumptions made in the 
safety analysis. A review of the failure modes of the affected 
sensors and relays indicates that a sluggish response of the 
instruments can be detected by other TS surveillances. Changing the 
method of periodically verifying instrument response for the 
selected instrument channels will not create any new accident 
initiators or scenarios. Periodic surveillance of these instruments 
will detect significant degradation in the channel characteristic. 
This proposed action will not alter the manner in which equipment 
operation is initiated, nor will the function demands on credited 
equipment be changed. No change is being made to procedures relied 
upon to respond to an off-normal event. As such, no new failure 
modes are being introduced.
    The sensors and relays in the affected channels will be able to 
meet the bounding response times as defined and presented in the LTR 
Supplement. It has been found acceptable to use component bounding 
response times in place of actual measured response times to ensure 
that instrumentation systems will meet response time requirements of 
the accident analyses. In addition, [Exelon Generation Company, LLC] 
EGC's adherence to the conditions listed in the NRC Safety 
Evaluations for the LTR and Supplement provides additional assurance 
that the instrumentation systems will meet the response time 
requirements of the accident analyses.
    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.
    Implementation of the BWROG LTR methodologies for eliminating 
selected response time testing requirements does not involve a 
significant reduction in the margin of safety. The current response 
time limits are based on the maximum values assumed in the plant 
safety analyses. The analyses conservatively establish the margin of 
safety. The elimination of the selected response time testing does 
not affect the capability of the associated systems to perform their 
intended function within the allowed response time used as the basis 
for plant safety analyses. Plant and system response to an 
initiating event will remain in compliance within the assumptions of 
the safety analyses, and therefore, the margin of safety is not 
affected. This is based on the ability to detect a degraded response 
time of an instrument or relay by the other required TS tests, 
component reliability, and redundancy and diversity of the affected 
functions, as justified in the reviewed and approved LTR and 
Supplement.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of application for amendments: February 27, 2004.
    Description of amendment request: The proposed change to the 
Technical Specifications (TSs) supports the activation of the trip 
outputs of the previously-installed Oscillation Power Range Monitor 
(OPRM) portion of the Power Range Neutron Monitoring (PRNM) system. 
Specifically, this proposed change will revise TS Sections 3.3.1.1, 
``Reactor Protection System Instrumentation,'' and 3.4.1, 
``Recirculation Loops Operating Reporting Requirements,'' and their 
associated TS Bases, and TS Section 5.6.5, ``Core Operating Limits 
Report (COLR).'' In addition, the proposed change deletes the Interim 
Corrective Action requirements from the Recirculation Loops Operating 
TSs.
    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 modification has no impact on any of the 
previously installed PRNM functions. Plant operation in portions of 
the former restricted region may potentially cause a marginal 
increase in the probability of occurrence of an instability event. 
This potential increase in probability is acceptable because the 
OPRM function will automatically detect the condition and initiate a 
reactor scram before the Minimum Critical Power Ratio (MCPR) Safety 
Limit is reached. Consequences of the potential instability event 
are reduced because of the more reliable automatic detection and 
suppression of an instability event, and the elimination of 
dependence on the manual operator actions.
    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 modification replaces procedural actions that 
were established to avoid operating conditions where reactor 
instabilities might occur with an NRC approved automatic detect and 
suppress function.
    Potential failures in the OPRM Upscale function could result in 
either failure to take the required mitigating action or an 
unintended reactor scram. These are the same potential effects of 
failure of the operator to take the correct appropriate action under 
the current procedural actions. The net effect of the modification 
changes the method by which an instability event is detected and by 
which mitigating action is initiated, but does not change the type 
of stability event that could occur. The effects of failure of the 
OPRM equipment are limited to reduced or failed mitigation, but such 
failure cannot cause an instability event or other type of accident.
    Therefore, since no radiological barrier will be challenged as a 
result of activating the OPRM trip function, it is concluded that 
this proposed activity does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The current safety analyses assume that the 
existing procedural actions are adequate to prevent an instability 
event. As a result, there is currently no quantitative or 
qualitative assessment of an instability event with respect to its 
impact on MCPR.
    The OPRM trip function is being implemented to automate the 
detection (via direct measurement of neutron flux) and subsequent 
suppression (via scram) of an instability event prior to exceeding 
the MCPR Safety Limit. The OPRM trip provides a trip output of the 
same type as currently used for the Average Power Range Monitor 
(APRM). Its failure modes and types are identical to those for the 
present APRM output. Currently, the MCPR Safety Limit is not 
impacted by an instability event since the event is ``mitigated'' by 
manual means via the procedural actions, which prevent plant 
operating conditions where an instability event is possible. In both 
methods of mitigation (manual and automated), the margin of safety 
associated with the MCPR Safety Limit is maintained.
    Therefore, since the MCPR Safety Limit will not be exceeded as a 
result of an

[[Page 19571]]

instability event following implementation of the OPRM trip 
function, it is concluded that 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. Edward Cullen, Vice President and 
General Counsel, Exelon Generation Company, LLC, 2301 Market Street, 
S23-1, Philadelphia, PA 19101.
    NRC Section Chief: Darrell Roberts, Acting.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: February 27, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) Section 5.6.2.6, ``Post Accident 
Sampling,'' 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 TS 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 February 27, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The 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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602-1551.
    NRC Section Chief: William F. Burton, Acting.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: January 28, 2004.
    Description of amendment request: Duane Arnold Energy Center

[[Page 19572]]

implemented improved technical specifications in 1998 via Amendment 223 
using NUREG 1433, ``Standard Technical Specifications--General Electric 
Plants BWR/4,'' Revision 1, as a model. The proposed amendment would 
revise Technical Specification Sections 5.5.11, 1.4, 3.3.1.1, and 5.5.2 
to adopt the following selected NRC approved generic changes to the 
improved technical specification NUREG.
     Technical Specification Task Force (TSTF)-273, 
Revision 2, Safety Function Determination Program Clarifications.
     TSTF-284, Revision 3, Add ``Met'' versus 
``Perform'' to Specification 1.4, Frequency.
     TSTF-264, Deletion of Flux Monitors Specific 
Overlap Surveillance Requirements.
     TSTF-299, Administrative Controls Program 
5.5.2.b Test Interval Defined and Allowance for 25 Percent Extension of 
Frequency.
    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:

Adoption of TSTF-273, Revision 2, and TSTF-284, Revision 3

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves reformatting, renumbering, and 
rewording the existing Technical Specifications. The reformatting, 
renumbering, and rewording process involves no technical changes to 
the existing Technical Specifications. As such, this change is 
administrative in nature and does not affect initiators of analyzed 
events or assumed mitigation of accident or transient events. 
Therefore, 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?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change will not impose any new or eliminate any old requirements. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed change will not reduce a margin of safety because 
it has no effect on any safety analyses' assumptions. This change is 
administrative in nature. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

Adoption of TSTF-264, Revision 0

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change deletes Surveillance Requirements. 
Surveillances are not initiators to any accident previously 
evaluated. Consequently, the probability of an accident previously 
evaluated is not significantly increased. The equipment being tested 
is still required to be Operable and capable of performing the 
accident mitigation functions assumed in the accident analysis. As a 
result, the consequences of any accident previously evaluated are 
not significantly affected. Therefore, the proposed change does 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?
    Response: No.
    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. The 
remaining Surveillance Requirements are Consistent with industry 
practice and are considered to be sufficient to prevent the removal 
of the subject Surveillances from creating a new or different type 
of accident. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Response: No.
    The deleted Surveillance Requirements do not result in a 
significant reduction in the margin of safety. As provided in the 
justification, the change has been evaluated to ensure that the 
deleted Surveillance Requirements are not necessary for verification 
that the equipment used to meet the LCO [limiting condition for 
operation] can perform its required functions. Thus, appropriate 
equipment continues to be tested in a manner and at a frequency 
necessary to give confidence that the equipment can perform its 
assumed safety function. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

Adoption of TSTF-299, Revision 0

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change provides more stringent requirements for 
operation of the facility. These more stringent requirements do not 
result in operation that will increase the probability of initiating 
an analyzed event and do not alter assumptions relative to 
mitigation of an accident or transient event. The more restrictive 
requirements continue to ensure process variables, structures, 
systems, and components are maintained consistent with the safety 
analyses and licensing basis. Therefore, this change does 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?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change does impose different requirements. However, these changes 
are consistent with the assumptions in the safety analyses and 
licensing basis. Thus, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed change provides additional restrictions which 
enhance plant safety. This change maintains requirements within the 
safety analyses and licensing basis. Therefore, this change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Morgan Lewis, 1111 
Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Section Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: February 27, 2004.
    Description of amendment request: The proposed amendment would 
remove license condition 2.C.(2)(b) to perform large transient testing 
as part of the extended power uprate (EPU) power ascension testing 
program at the Duane Arnold Energy Center (DAEC).
    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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The requested licensing action would remove the current 
requirement to perform specific large transient tests as part of the 
DAEC EPU power ascension testing program. No other changes are 
proposed. Therefore,

[[Page 19573]]

the probability of an accident previously evaluated is not 
significantly increased.
    The proposed action will not affect any System, Structure, or 
Component designed for the mitigation of previously analyzed events. 
The proposed change does not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Thus, the proposed change will not increase the consequences of any 
previously evaluated accident.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The requested licensing action would remove the current 
requirement to perform specific large transient tests as part of the 
DAEC EPU power ascension testing program. No other changes are 
proposed. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety.
    Performance of these specific large transient tests is not 
necessary to ensure acceptable plant operation at the higher thermal 
power level. Simple, integrated systems tests are performed in lieu 
of the complex, challenging large transient tests. Other required 
testing of the specific SSCs that have been modified for EPU ensures 
that the plant will respond as expected during any abnormal 
operating event, including these specific transients. Thus, the 
proposed elimination of the large transient tests will not 
significantly reduce any margin of safety from that previously 
approved for EPU operation at the DAEC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Morgan Lewis, 1111 
Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: January 30, 2004.
    Description of amendment request: The proposed amendment would 
revise Monticello Nuclear Generating Plant (MNGP) Technical 
Specifications (TS) to (1) clarify the permissive set point for the 
source range monitor (SRM) detector not-fully-inserted rod block 
bypass, (2) correct a typographical error in the surveillance 
requirement for suppression pool temperature monitoring, (3) clarify 
the set point for the pressure suppression chamber-reactor building 
vacuum breakers instrumentation, (4) clarify the operating force 
requirements for the pressure suppression chamber--drywell vacuum 
breakers surveillance test, and (5) make corrections resulting from 
License Amendments (LAs) 130 and 132.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The SRM Detector-not-fully-inserted rod block bypass set point, 
the Pressure Suppression Chamber--Reactor Building Vacuum Breakers 
actuation instrumentation set point requirement and the Pressure 
Suppression Chamber--Drywell Vacuum Breakers surveillance test 
requirements are being clarified in the MNGP TS to ensure these 
functions will adequately support safe operation of the facility. 
Typographical errors are being corrected along with corrections 
resulting from omissions and an oversight from previous LAs. The 
proposed TS changes do not introduce new equipment or new equipment 
operating modes, nor do the proposed changes alter existing system 
relationships. The changes do not affect plant operation, design 
function or any analysis that verifies the capability of a SSC 
[structure, system or component] to perform a design function. 
Further, the proposed changes do not increase the likelihood of the 
malfunction of any structure, system or component (SSC) or impact 
any analyzed accident. Consequently, the probability of an accident 
previously evaluated is not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The SRM Detector-not-fully-inserted rod block bypass set point, 
the Pressure Suppression Chamber--Reactor Building Vacuum Breakers 
actuation instrumentation set point requirement and the Pressure 
Suppression Chamber--Drywell Vacuum Breakers surveillance test 
requirements are being clarified in the MNGP TS to ensure these 
functions will adequately support safe operation of the facility. 
Typographical errors are being corrected along with corrections 
resulting from omissions and an oversight from previous LAs. The 
changes do not create the possibility of new credible failure 
mechanisms, or malfunctions. These changes do not modify the design 
function or operation of any SSC. Further the changes do not involve 
physical alterations of the plant; no new or different type of 
equipment will be installed. The proposed changes do not introduce 
new accident initiators. Consequently, the changes cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    The SRM Detector-not-fully-inserted rod block bypass set point, 
the Pressure Suppression Chamber--Reactor Building Vacuum Breakers 
actuation instrumentation set point requirement and the Pressure 
Suppression Chamber--Drywell Vacuum Breakers surveillance test 
requirements are being clarified in the MNGP TS to ensure these 
functions will adequately support safe operation of the facility. 
Typographical errors are being corrected along with corrections 
resulting from omissions and an oversight from previous LAs. These 
changes do not exceed or alter a design basis or a safety limit for 
a parameter established in the MNGP Updated Safety Analysis Report 
(USAR) or the MNGP facility license. Consequently, the changes do 
not result in a significant reduction in the margin of safety.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: February 10, 2004.
    Description of amendment request: The proposed change involves the 
extension from 1 hour to 24 hours of the completion time (CT) for 
Action (a) of Technical Specification (TS) 3.5.1.1, 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

[[Page 19574]]

decreases to below the accumulator pressure. Action (a) of TS 3.5.1.1 
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. This change was proposed by the Westinghouse Owners 
Group participants in the TS Task Force (TSTF) and is designated TSTF-
370. 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 included in its 
application several minor changes to make the plant specific TS more 
consistent with the STS and TSTF-370. The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated February 10, 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 Basis Section 3.5.1.1, 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 WCAP-15049-A, the proposed change will 
allow plant operation with an inoperable accumulator for up to 24 
hours, instead of 1 hour, before the plant would be required to 
begin shutting down. The impact of the increase in the accumulator 
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A 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-A for the accumulator CT increase meet the criterion of 
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using 
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications,'' for all cases except those that are based on 
design basis success criteria. As indicated in WCAP-15049-A, 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-A evaluation as a worst case 
data point. In addition, WCAP-15049-A 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-A 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-A evaluation demonstrates that the small 
increase in risk due to increasing the CT for an inoperable 
accumulator 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 Basis Section 
3.5.1.1, 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 WCAP-15049-
A, the proposed change will allow plant operation with an inoperable 
accumulator for up to 24 hours, instead of 1 hour, before the plant 
would be required to begin shutting down. 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 large LOCA 
with loss of offsite power) would be significantly lower than the 
large LOCA frequency of the WCAP-15049-A 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-A evaluation.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: February 20, 2004.
    Description of amendment request: The proposed amendments would 
revise Vogtle Electric Generating Plant, Units 1 and 2 Administrative 
Controls Section 5.2.2.g of Technical Specification to limit the 
requirement of the Shift Technical Advisor function to Modes 1-4 in 
accordance with NUREG 0737.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to TS [Technical Specification] 5.2.2.g does 
not significantly

[[Page 19575]]

increase the probability or consequences of an accident previously 
evaluated in the FSAR [Final Safety Analysis Report]. This revision 
does not have any effect on the probability of any accident 
initiators. The consequences of accidents previously evaluated in 
the FSAR are not adversely affected by this proposed change because 
the STA [Shift Technical Advisor] is not credited for mitigation of 
any accidents. The proposed change which requires the STA function 
to be available while in Modes 1-4 is in accordance with the 
requirements of NUREG 0737, Item I.A.1.1. Consequently, the 
probability or consequences of an accident previously evaluated are 
not significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    The proposed change to TS 5.2.2.g does not create the 
possibility of a new or different kind of accident from any 
previously evaluated. No new accident scenarios, failure mechanism, 
or limiting single failures are introduced as a result of the 
proposed change. The proposed Technical Specifications change does 
not challenge the performance or integrity of any safety-related 
systems. The proposed change to TS 5.2.2.g is in accordance with 
NUREG 0737.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change to TS 5.2.2.g will not reduce a margin of 
safety because it has no direct effect on any safety analyses 
assumptions. The STA function is to evaluate plant conditions and 
provide advice to the shift supervisor during plant transients and 
accidents. The proposed change limits the requirements for the STA 
function to Modes 1-4 in accordance with NUREG 0737. The STA 
function is not credited for the mitigation of any accidents 
previously evaluated.

    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. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: February 26, 2004.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 5.6.6, ``Reactor Coolant System 
(RCS) Pressure and Temperature Limits Report (PTLR)'', to reference the 
NRC-approved methodology for developing Pressure-Temperature limits and 
Cold Overpressure Protection System setpoints and the methodology used 
to justify eliminating the reactor vessel closure head/vessel flange 
requirements. The proposed amendment would also revise TS 3.4.12, 
``Cold Overpressure Protection System (COPS)'', to change the Reactor 
Coolant System vent size.
    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?
    No. The proposed changes to the Technical Specifications [TS] 
and PTLRs [Pressure and Temperature Limits Reports] do not affect 
any plant equipment, test methods, or plant operation, and are not 
initiators of any analyzed accident sequence. Operation in 
accordance with the proposed TS will ensure that all analyzed 
accidents will continue to be mitigated by the SSCs [systems, 
structures and components] as previously analyzed.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes do not introduce any new equipment, 
create new failure modes for existing equipment, or create any new 
limiting single failures. The changes to the P-T [pressure-
temperature] limits and COPS [Cold Overpressure Protection Systems] 
setpoints will ensure that appropriate fracture toughness margins 
are maintained to protect against reactor vessel failure during both 
normal and low temperature operation. The changes to the P-T limits 
and COPS setpoints are consistent with the methodology approved by 
the NRC [Nuclear Regulatory Commission] in WCAP-14040, Rev. 4. Plant 
operation will not be altered, and all safety functions will 
continue to perform as previously assumed in accident analyses.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    No. The proposed changes will not adversely affect the operation 
of plant equipment or the function of any equipment assumed in the 
accident analysis. The utilization of ASME [American Society of 
Mechanical Engineers] Code Case N-640 maintains the relative margin 
of safety commensurate with that which existed at the time that ASME 
B&PV [Boiler and Pressure Vessel] Code, Section XI, Appendix G was 
approved in 1974 and will ensure an acceptable 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. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: John A. Nakoski.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant (BFN), Unit 1, Limestone County, Alabama

    Date of amendments request: March 9, 2004 (TS 434).
    Description of amendments request: The proposed amendment would 
lower the current Reactor Vessel Water Level--Low, Level 3 Allowable 
Value in the Unit 1 Technical Specifications for several instrument 
functions to reduce the likelihood of unnecessary reactor scrams and 
the resultant engineered safety feature actuations by increasing the 
operating range between the normal reactor vessel water level and Level 
3 trip functions.
    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?
    No. The Reactor Vessel Water Level--Low, Level 3 functions are 
in response to water level transients and are not involved in the 
initiation of accidents or transients. Therefore, reducing the BFN, 
Unit 1, Level 3 Allowable Value does not increase the probability of 
an accident previously evaluated.
    Additionally, the results of the safety evaluation associated 
with the lowering of the Level 3 Allowable Value concludes that the 
previously evaluated transient and accident consequences are not 
significantly affected by the change. Therefore, the proposed 
amendment does not involve a significant increase in the probability 
of consequences or 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?
    No. The proposed amendment to lower the BFN, Unit 1, Reactor 
Vessel Water Level--Low, Level 3 Allowable Value does not involve a 
hardware change and the purpose of the Level 3 function is not 
affected. The Level 3 functions will continue to fulfill their 
design objective. The proposed changes do not create the possibility 
of any new failure mechanisms. No new external threats or release 
pathways are created. Therefore, reduction of the Allowable Value 
does not result in the possibility of a new or different kind of 
accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?

[[Page 19576]]

    No. The results of the safety evaluation associated with the 
reducing the BFN, Unit 1, Reactor Vessel Water Level--Low, Level 3 
Allowable Value concluded that transient and accident consequences 
remain within the required acceptance criteria. Therefore, the 
margin of safety is not reduced for any event evaluated.

    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: William F. Burton, Acting.

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

    Date of amendment request: March 5, 2004.
    Description of amendment request: The proposed amendments would 
delete Technical Specifications (TSs) 3.6.4.1, ``Hydrogen Monitors,'' 
and 3.6.4.2, ``Electric Hydrogen Recombiners-W.'' The proposed changes 
support Title 10, Code of Federal Regulations, Part 50, Section 44 (10 
CFR 50.44), ``Standards for Combustible Gas Control system in Light-
Water-Cooled Power Reactors'' and are consistent with the Industry/
Technical Specification Task Force (TSTF) Standard TS Change Traveler, 
TSTF-447, ``Elimination of Hydrogen Recombiners and change to Hydrogen 
and Oxygen Monitors.''
    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:

    TVA has reviewed the proposed no significant hazards 
consideration determination published on September 25, 2003, (68 FR 
55416) as part of the consolidated line item improvement process 
(CLIIP). TVA has concluded that the proposed determination presented 
in the notice is applicable to SQN, and the determination is hereby 
incorporated by reference to satisfy the requirements of 10 CFR 
50.91(a).

    The United States Nuclear Regulatory Commission (NRC) staff has 
reviewed the licensee's analysis and, based on this review, it appears 
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: William F. Burton, Acting.

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

    Date of amendment request: March 5, 2004.
    Description of amendment request: The proposed amendments would 
delete surveillance requirement (SR) 4.9.2.c and SRs 4.10.3.2 and 
4.10.4.2 from the Technical Specifications (TSs). SR 4.9.2.c requires 
channel functional tests for each Source Range neutron flux monitor 
within 8 hours prior to initial core alterations. SRs 4.10.3.2 and 
4.10.4.2 require channel functional tests for each Power Range and 
Intermediate Range neutron flux monitor within 12 hours prior to the 
initiation of a physics test. In addition, the proposed changes include 
revisions to the associated TS bases (3/4.9.2, 3/4.10.3, and 3/4.10.4).
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10, Code of Federal Regulations, 
Part 50, Section 91(a) (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?
    No. The proposed amendment removes the requirement to perform an 
additional CHANNEL FUNCTIONAL TEST (CFT) on the Intermediate and 
Power Range functions within 12 hours of performing a PHYSICS TEST. 
The Intermediate and Power Range instrumentation is determined to be 
OPERABLE by periodic SRs which must be confirmed to be within 
frequency prior to making the reactor critical. The proposed 
amendment also removes the requirement to perform an additional CFT 
on the Source Range monitors. The Source Range instrumentation is 
determined to be OPERABLE by periodic SRs, which must be confirmed 
to be within frequency prior to Mode 6, prior to CORE ALTERATIONS, 
and must remain OPERABLE. A CFT for the Source Range, Intermediate 
Range, or Power Range instrumentation is not a precursor to, or 
assumed to be an initiator of any analyzed accident. Therefore, this 
change does not involve a significant increase in the probability of 
an accident previously evaluated.
    Regarding a significant increase in the consequences of an 
accident, several factors must be considered. First the PHYSICS 
TESTS are performed in accordance with the TSs in Mode 2. Therefore, 
the power level of the reactor is limited to 5 percent or less. 
Along with this, the reactor trip function of the Intermediate Range 
detectors will be unaffected by the proposed amendment and 
therefore, will be available to mitigate a reactivity transient at 
low power. Further, the trip setpoint for the Power Range monitors 
are decreased during startup. This setpoint reduction provides an 
additional measure to limit a reactivity excursion. 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?
    No. The proposed changes permit the conduct of normal operating 
evolutions during limited periods when additional controls over 
reactivity margin are imposed by the TSs. The proposed change does 
not introduce any new equipment into the plant or significantly 
alter the manner in which existing equipment will be operated. The 
proposed changes are not based on a change in the design or 
configuration of the plant. The changes to operating allowances are 
minor and are only applicable during certain conditions. The 
operating allowances are consistent with those acceptable at other 
times. The proposed changes delete the requirements for the 
performance of a CFT for the Source Range, Intermediate Range, and 
Power Range instrumentation within 8 hours of initiating CORE 
ALTERATIONS for the Source Range monitors and within 12 hours of 
starting a PHYSICS TEST for the Intermediate Range and Power Range 
instrumentation. Since the proposed changes only allow activities 
that are presently approved and routinely conducted, no possibility 
exists for a new or different kind of accident from those previously 
evaluated. 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?
    No. As stated previously, the proposed change deletes the 
requirement to perform an additional CFT for the Source Range, 
Intermediate Range, and Power Range instrumentation within 8 hours 
of initiating CORE ALTERATIONS for the Source Range monitors and 
within 12 hours of starting a PHYSICS TEST for the Intermediate 
Range and Power Range instrumentation. The Source Range, 
Intermediate Range, and Power Range instrumentation channels are 
determined to be OPERABLE by meeting the requirements of the 
periodic surveillance. These SRs are not affected by the proposed 
amendment. The proposed changes do not involve a significant 
reduction in a margin of safety because the ability to monitor the 
reactor during the applicable operating conditions and modes of 
operation will be maintained. The proposed changes do not affect 
these operating restrictions and the margin of safety which assures 
the ability to

[[Page 19577]]

monitor the reactor is not affected. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.

    The United States Nuclear Regulatory Commission (NRC) staff has 
reviewed the licensee's analysis and, based on this review, it appears 
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: William F. Burton, Acting.

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

    Date of amendment request: March 5, 2004.
    Description of amendment request: The proposed amendments would 
change Technical Specification (TS) 4.0.5.c. Specifically, the proposed 
change would extend the examination frequency for the reactor coolant 
pump (RCP) motor flywheel from a 10-year interval to an interval not to 
exceed 20 years. This proposed change is consistent with the Industry/
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-421, ``Revision to RCP Flywheel 
Inspection Program (WCAP-15666).''
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10, Code of Federal Regulations, 
Part 50, Section 91(a) (10 CFR 50.91(a)), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    TVA has reviewed the proposed no significant hazards 
consideration determination published on June 24, 2003 (68 FR 
37590), as part of the consolidated line item improvement process 
(CLIIP). TVA has concluded that the proposed determination presented 
in the notice is applicable to SQN, and the determination is hereby 
incorporated by reference to satisfy the requirements of 10 CFR 
50.91(a).

    The United States Nuclear Regulatory Commission (NRC) staff has 
reviewed the licensee's analysis and, based on this review, it appears 
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: William F. Burton, Acting.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia

    Date of amendment request: March 4, 2004.
    Description of amendment request: The proposed amendments would 
delete the note in Improved Technical Specification Surveillance 
Requirement 3.4.12.7 that permitted the performance of the Channel 
Operational Test within 12 hours of entering a mode in which the power-
operated relief valves (PORVs) are required to be operable for low 
temperature overpressure protection (LTOP).
    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 changes involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The proposed changes to perform a Channel Operational Test on 
each required PORV at least 31 days prior to entering the LTOP Mode 
will continue to ensure verification and adjustment, if required, of 
its lift setpoint. Changes will not affect the probability of 
occurrence of any accident previously analyzed: nor alter the design 
assumptions, conditions, and configuration of the facility or the 
manner in which the plant is operated and maintained. Therefore, the 
proposed changes do not involve a significant increase in the 
consequences of any previously analyzed accident.
    2. Do changes create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    The proposed changes to perform a Channel Operational Test on 
each required PORV at least 31 days prior to entering the LTOP Mode 
will not create any new accident or event initiators. No systems, 
structures, or components are being physically modified such that 
the design function is being altered. The proposed changes do not 
impose any new or different requirements for the performance of the 
Channel Operational Test. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from 
those previously analyzed.
    3. Do changes involve a significant reduction in a margin of 
safety?
    The proposed changes do not involve any change to the safety 
analysis limits. The level of safety of facility operation is 
unaffected by the proposed changes since there is no change in the 
intent for the performance of the Channel Operational Test. 
Therefore, it is concluded that the margin of safety will not be 
reduced by the implementation of the changes.

    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: John A. Nakoski.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of application for amendment: February 25, 2004.
    Brief description of amendments: The amendment would extend the 
implementation date for Amendment Nos. 261 and 238 for Calvert Cliffs 
Units 1 and 2, respectively, to July 1, 2004. The changes to the 
reactor pressure vessel pressure-temperature limits cooldown rates that 
were approved by Amendment Nos. 261 and 238 are more conservative than 
the plants existing rates and result in a longer cooldown period. The 
existing cooldown rates are acceptable through the end of 2004.
    Date of publication of individual notice in Federal Register: March 
5, 2004 (69 FR 10487).
    Expiration date of individual notice: May 5, 2004.

[[Page 19578]]

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 e-mail to pdr@nrc.gov.

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

    Date of application for amendment: December 23, 2003, as 
supplemented by letter dated January 30, 2004.
    Brief description of amendment: The amendment modified Technical 
Specification (TS) requirements for mode change limitations to adopt 
the TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility in 
Mode Restraints.''
    Date of issuance: March 29, 2004.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 241.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: January 20, 2004 (69 
FR 2738).
    The January 30, 2004, letter provided clarifying information within 
the scope of the original application and did not change the staff's 
initial proposed no significant hazards consideration determination. 
The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 29, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 7, 2003, and its 
supplement dated December 18, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Section 5.5.6, ``Containment Tendon Surveillance 
Program,'' for consistency with the requirements of 10 CFR 50.55a(g)(4) 
for components classified as Code Class CC. The amendments also delete 
the provisions of Surveillance Requirement (SR) 3.0.2 from this TS. In 
addition, the amendments revise TS 5.5.16, ``Containment Leakage Rate 
Testing Program,'' to add exceptions to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Testing Program.'' Also, the 
paragraphs in Section 5.5.16 have been sequenced to more clearly 
separate the requirements of the program. This is considered an 
administrative change and is consistent with the guidance in NUREG-
1432, ``Standard Technical Specifications Combustion Engineering 
Plants,'' Revision 2.
    Date of issuance: March 19, 2004.
    Effective date: March 19, 2004, and shall be implemented within 90 
days of the date of issuance.
    Amendment Nos.: Unit 1-151, Unit 2--151, Unit 3--151.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68659) The December 18, 2003, supplemental letter provided revised 
technical specification pages to reflect changes that were approved in 
Amendment No. 149, did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 19, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 17, 2003 as 
supplemented by letter dated February 20, 2004.
    Brief description of amendments: The amendments revise the 
technical specifications to support the replacement of part-length 
control element assemblies (CEAs) with a new design, referred to as 
part-strength CEAs. The two designs are geometrically very similar and 
contain essentially the same amount and type of neutron absorber in the 
lower half of the assemblies, which is the region of the CEAs inserted 
into the reactor core during normal operations.
    Date of issuance: March 23, 2004.
    Effective date: March 23, 2004, and shall be implemented within 60 
days of the date of issuance.
    Amendment Nos.: Unit 1--152, Unit 2--152, Unit 3--152.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68657). The February 20, 2004, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination.

[[Page 19579]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 23, 2004.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-325, Brunswick Steam 
Electric Plant, Unit 1, Brunswick County, North Carolina

    Date of application for amendment: October 31, 2003, as 
supplemented March 4, March 12, and March 19, 2004.
    Brief description of amendment: The amendment revised the Minimum 
Critical Power Ratio Safety Limit contained in Technical Specification 
2.1.1.2.
    Date of issuance: March 26, 2004.
    Effective date: Effective as of the date of issuance and shall be 
implemented prior to startup for Unit 1, Cycle 15, operation.
    Amendment No.: 231.
    Facility Operating License Nos. DPR-71: Amendment changes the 
Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
693). The March 4, March 12, and March 19, 2004, supplemental letters 
provided clarifying information that did not change the scope of the 
proposed amendment as described in the original notice of proposed 
action published in the Federal Register and did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 2004.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: February 14, 2003, as 
supplemented by letters dated November 10 and December 10, 2003, and 
January 30, 2004.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 5.6.3.d to allow an increase in the decay heat load 
from 1.0 MBTU/hr to 7.0 MBTU/hr for fuel stored in Spent Fuel Pools C 
and D at Shearon Harris Nuclear Power Plant, Unit 1.
    Date of issuance: March 26, 2004.
    Effective date: March 26, 2004.
    Amendment No.: 115.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12948). The November 10 and December 10, 2003, and January 30, 2004, 
supplements provided clarifying information that did not change the 
scope of the proposed amendment as described in the original notice of 
proposed action published in the Federal Register and did not change 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 2004.
    No significant hazards consideration comments received: No.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear 
Plant, Charlevoix County, Michigan

    Date of application for amendment: November 20, 2002, and August 6, 
2003, as supplemented by letters dated December 1, 2003, and February 
20, 2004.
    Brief description of amendment: The amendment revises the Big Rock 
Point License and Defueled Technical Specifications to remove reactor 
operational and administrative requirements that are no longer 
applicable due to the transfer of all spent fuel from the spent fuel 
pool into dry cask storage at the Big Rock Point Independent Spent Fuel 
Storage Installation.
    Date of issuance: March 19, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 125.
    Facility Operating License No. DPR-6: Amendment revises the 
Defueled Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2800), and November 25, 2003 (68 FR 66133). The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
March 19, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 4, 2003, as supplemented 
May 13 and September 18, 2003, and February 12 and March 10, 2004.
    Brief description of amendment: The amendment revised selected 
sections of the Technical Specifications (TSs) based upon a re-analysis 
of fuel handling accidents (FHAs). The revised analysis is based upon 
selective implementation of the alternative source term methodology of 
Regulatory Guide 1.183, and in accordance with Title 10 of the Code of 
Federal Regulations, Section 50.67. Specifically, the amendment 
revised: TS 3.7.8, ``Plant Systems, Control Room Envelope 
Pressurization System;'' TS 3.9.4, ``Refueling Operations, Containment 
Building Penetrations;'' TS 3.9.9, ``Refueling Operations, Containment 
Purge and Exhaust Isolation System,'' and TS 3.9.12, ``Refueling 
Operations, Fuel Building Exhaust Filter System.''
    Date of issuance: March 17, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 219.
    Facility Operating License No. NPF-49: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: March 4, 2003 (68 FR 
40711). The May 13 and September 18, 2003, and February 12 and March 
10, 2004, supplements contained clarifying information and did not 
change the staff's initial proposed finding of no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2004.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: December 5, 2003, as 
supplemented on February 9, 2004.
    Brief description of amendment: The amendment revised the Safety 
Limit Minimum Critical Power Ratio values in Technical Specification 
1.1.A.1 to incorporate the results of the cycle-specific core reload 
analysis for Vermont Yankee Nuclear Power Station Cycle 24 operation.
    Date of Issuance: March 22, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 217.
    Facility Operating License No. DPR-28: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2741). The supplement dated February 9, 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 this amendment is contained 
in a Safety Evaluation dated March 22, 2004.

[[Page 19580]]

    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: March 26, 2003, as supplemented 
on July 24, 2003.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) regarding reactor pressure vessel (RPV) fracture 
toughness and material surveillance requirements (SRs). Specifically, 
the amendment revised the pressure-temperature limits for the RPV as 
specified in TS Figures 3.6.1, 3.6.2, and 3.6.3. In addition, the 
amendment deleted TS 4.6.A.5, which specifies plant-specific RPV 
material SRs. These plant-specific SRs are being replaced by 
implementing the Boiling Water Reactor Vessel and Internals Project 
(BWRVIP) RPV integrated surveillance program (ISP). The details of the 
BWRVIP ISP will be added to the Vermont Yankee Nuclear Power Station 
Updated Final Safety Analysis Report.
    Date of Issuance: March 29, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 218.
    Facility Operating License No. DPR-28: Amendment revised the TSs.
    Date of initial notice in Federal Register: April 29, 2003 (68 FR 
22747). The supplement dated July 24, 2003, 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 this amendment is contained 
in a Safety Evaluation dated March 29, 2004.
    No significant hazards consideration comments received: No.

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

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: June 11, 2003, as supplemented 
on December 5, December 30, 2003, and February 18, 2004.
    Brief description of amendments: The amendments revise technical 
specification 3.7.8 to permit a one-time extension from 72 hours to 144 
hours for the completion time required to restore a unit specific 
essential service water train to operable status.
    Date of issuance: March 18, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 136/136, 130/130.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 30, 2003.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 18, 2004.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: September 8, 2003.
    Brief description of amendments: The amendments modified Technical 
Specifications requirements to adopt the provisions of Industry/
Technical Specification Task Force (TSTF) change 359, ``Increase 
Flexibility in Mode Restraints.''
    Date of issuance: March 12, 2004.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 169 and 132.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68668).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 12, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 15, 2002, as supplemented by 
letter dated January 14, 2004.
    Description of amendment request: The amendment revises the 
Technical Specifications (TSs) to relocate the boron concentration 
limits and ``Safety Limits'' figures to the Core Operating Limits 
Report. Some limiting conditions and actions are revised to be 
consistent with the Improved Standard Technical Specifications.
    Date of issuance: March 23, 2004.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 96.
    Facility Operating License No. NPF-86: The amendment revises the 
TS.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36931). The January 14, 2004, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the amendment beyond the scope of 
the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 23, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 25, 2003.
    Brief description of amendment: The amendment revises the Technical 
Specification (TS) for Limiting Condition for Operation requirement 
3.5.1 to incorporate TS Task Force Traveler 318 to allow one low 
pressure coolant injection pump inoperable in each of the two emergency 
core cooling system divisions.
    Date of issuance: March 31, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 203.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59218).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 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: March 27, 2003, as supplemented 
on November 3, 2003, and January 28, 2004.
    Brief description of amendments: The amendment revises Technical 
Specification Surveillance Requirement 3.2.4.2, ``Rod Group Alignment 
Limits.'' The revision expands the alignment limits on allowable rod 
cluster control assembly, or rod, deviation from demanded position. The 
change applies in Mode 1, when operating at greater than 85 percent of 
rated thermal power.
    Date of issuance: March 29, 2004.

[[Page 19581]]

    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 212 and 217.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 29, 2003 (68 FR 
22749).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 11, 2003.
    Brief description of amendments: The amendments revise the 
technical specifications to allow use of the power distribution 
monitoring system (PDMS) for power distribution measurements as 
described in Topical Report WCAP-12462-P-A, ``BEACON: Core Monitoring 
and Support System.''
    Date of issuance: March 31, 2004.
    Effective date: March 31, 2004, and shall be implemented within 180 
days from the date of issuance.
    Amendment Nos.: Unit 1--164; Unit 2--166.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40717).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 31, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of application for amendments: September 18, 2003, as 
supplemented December 8, 2003, and February 24, 2004.
    Description of amendment request: The amendments revised the 
pressure-temperature limit curves in Technical Specification (TS) 
3.4.9.
    Date of issuance: March 10, 2004.
    Effective date: March 10, 2004.
    Amendment Nos.: 288 & 247.
    Facility Operating License No. DPR-52 and DPR-68: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: October 28, 2003 (68 FR 
61480). The December 8, 2003, and February 24, 2004, letters provided 
clarifying information that did not change the scope of the original 
request or the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 10, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: March 24, 2003, as supplemented 
December 4, 2003, and February 12, 2004.
    Brief description of amendment: The amendment revises the design 
and licensing basis failure modes and effects analysis for specific 
valves in the essential raw cooling water system, component cooling 
water system, and control air system to address a condition in which 
containment integrity, accident flood levels, and sump boron 
concentrations subsequent to a high-energy line break could not be 
automatically ensured, and, therefore, manual actions are required.
    Date of issuance: March 29, 2004.
    Effective date: As of the date of issuance and shall be implemented 
in conjunction with the next update to the Updated Final Safety 
Analysis Report required by 10 CFR 50.71(e).
    Amendment No.: 51.
    Facility Operating License No. NPF-90: Amendment revises the 
Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18287). The supplemental letters provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2004.
    No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket No. 50-445, Comanche Peak Steam 
Electric Station, Unit No. 1, Somervell County, Texas

    Date of amendment request: July 21, 2003, as supplemented by 
letters dated January 8, January 21, and March 8, 2004.
    Brief description of amendments: The Amendment revises the 
Technical Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' to allow the use of Westinghouse (Westinghouse Electric 
Station LLC) leak limiting Alloy 800 sleeves for repair of degraded SG 
tubes.
    Date of issuance: March 24, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 112.
    Facility Operating License No. NPF-87: The amendments revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 2003. 
Supplemental letters dated January 8, January 21, and March 8, 2004 
provided clarifying information that did not change the scope of the 
original Federal Register notice or the original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 24, 2004.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: December 8, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 5.5.6, ``Containment Tendon Surveillance 
Program,'' for consistency with the requirements of 10 CFR 50.55a(g)(4) 
for components classified as Code Class CC. The amendment also deletes 
the provisions of Surveillance Requirement (SR) 3.0.2 from this TS. In 
addition, the amendment revises TS 5.5.16, ``Containment Leakage Rate 
Testing Program,'' to add exceptions to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Testing Program.''
    Date of issuance: March 17, 2004.
    Effective date: March 17, 2004, and shall be implemented within 90 
days from the date of issuance.
    Amendment No.: 160.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
700).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2004.
    No significant hazards consideration comments received: No.

[[Page 19582]]

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

    Date of amendment request: October 17, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 5.5.6, ``Containment Tendon Surveillance 
Program,'' for consistency with the requirements of 10 CFR 50.55a(g)(4) 
for components classified as Code Class CC. The amendment also deletes 
the provisions of Surveillance Requirement (SR) 3.0.2 from this TS. In 
addition, the amendment revises TS 5.5.16, ``Containment Leakage Rate 
Testing Program,'' to add exceptions to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Testing Program.''
    Date of issuance: March 17, 2004.
    Effective date: March 17, 2004, and shall be implemented within 90 
days from the date of issuance.
    Amendment No.: 152.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64140).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2004.

    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 5th day of April 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-8047 Filed 4-12-04; 8:45 am]
BILLING CODE 7590-01-P