[Federal Register Volume 69, Number 110 (Tuesday, June 8, 2004)]
[Notices]
[Pages 32070-32080]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-12671]


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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, May 14 through May 27, 2004. The last 
biweekly notice was published on May 25, 2004 (69 FR 29761).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this

[[Page 32071]]

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 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-

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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 email to 
pdr@nrc.gov.

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

    Date of amendment request: April 26, 2004.
    Description of amendment request: The proposed amendment would 
revise the Completion Time for Required Action A.1 of Technical 
Specification 3.8.7, ``Inverters--Operating,'' from the current 24 
hours for a Division 1 or 2 Nuclear System Protection System (NSPS) 
inverter inoperable to 7 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS change revises the Completion Time for Required 
Action A.1 associated with the Division 1 and 2 NSPS inverters. 
Specifically, the proposed action allows continued unit operation, 
for up to 7 days, with an inoperable Division 1 or 2 NSPS inverter.
    The proposed change does not affect the design of the NSPS 
inverters, the operational characteristics or function of the 
inverters, the interfaces between the inverters and other plant 
systems, or the reliability of the inverters. An inoperable NSPS 
inverter is not considered as an initiator of any analyzed event. In 
addition, Required Actions and the associated Completion Times are 
not initiators of any previously evaluated accidents. Extending the 
Completion Time for an inoperable NSPS inverter would not have a 
significant impact on the frequency of occurrence for any accident 
previously evaluated. The proposed change will not result in changes 
to the plant activities associated with NSPS inverter maintenance, 
but rather will allow increased flexibility in the scheduling and 
performance of preventive maintenance. Therefore, this change will 
not significantly increase the probability of occurrence of any 
event previously analyzed.
    The consequences of a previously analyzed event are dependent on 
the initial conditions assumed in the analysis, the availability and 
successful functioning of equipment assumed to operate in response 
to the analyzed event, and the setpoints at which these actions are 
initiated. With an NSPS inverter inoperable, the affected instrument 
bus is capable of being fed from its dedicated safety-related 
alternate power supply, which is powered from a Class 1E 480 VAC bus 
through a step-down transformer and an isolation transformer. In the 
event of a Loss of Offsite Power (LOOP), the affected instrument bus 
will experience a momentary loss of power until the associated 
diesel generator (DG) re-energizes the 480 VAC bus. A LOOP with an 
inoperable NSPS inverter (i.e., instrument bus being powered by its 
alternate power supply) will result in a loss of power to the 
associated instrument bus until the associate DG re-energizes the 
Class 1E 480 VAC bus. All instruments supplied by the instrument bus 
would be restored with no adverse impact to the unit because no 
other instrument channels in the opposite train would be expected to 
be inoperable or in a tripped condition during this time, with the 
exception of routine surveillances. In the event of a failure to re-
energize the 480 VAC bus or of a transformer failure, the most 
significant impact on the unit is the failure of one train of 
Engineered Safety Feature (ESF) equipment to actuate. In this 
condition, the redundant train of ESF equipment will automatically 
actuate to mitigate the accident, and the affected unit would remain 
within the bounds of the accident analyses. In addition, there would 
be no adverse impact to the unit because no other instrument 
channels in the opposite train would be expected to be inoperable or 
in a tripped condition during this time, with the exception of 
routine surveillances.
    To fully evaluate the effect of the proposed NSPS inverter 
Completion Time extension, probabilistic risk assessment (PRA) 
methods and a deterministic analysis were utilized. The Incremental 
Conditional Core Damage Probability (ICCDP) and Incremental 
Conditional Large Early Release Probability (ICLERP) for each 
inverter division are sufficiently below the regulatory guidelines 
to be able to call the risk change small. Hence, the guidelines of 
Regulatory Guide 1.177, ``An Approach for Plant-Specific, Risk-
Informed Decision-Making: Technical Specifications,'' for the 
increased inverter Completion Time have been met. Furthermore, the 
evaluation of changes in Core Damage Frequency (CDF) and Large Early 
Release Frequency (LERF) due to the expected increased inverter 
unavailability, as mitigated by the compensating measures assumed in 
the analysis, have been shown to meet the risk significance criteria 
of Regulatory Guide 1.174, ``An Approach for Using Probabilistic 
Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes 
to the Licensing Basis,'' with substantial margin. This calculation 
supports the increase in the Division 1 and 2 inverter Completion 
Times from a quantitative risk-informed perspective consistent with 
the plant operational and maintenance practices. Therefore, the 
request for extending the Completion Time will not significantly 
increase the consequences of an accident previously evaluated.
    In summary, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed 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 CPS is 
operated. There are no setpoints at which protective or mitigative 
actions are initiated that are affected by this proposed action. The 
use of the alternative Class 1E power source for the instrument bus 
is consistent with the CPS plant design. The change does not alter 
assumptions made in the safety analysis. 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 
alteration in the procedures, which ensure the unit remains within 
analyzed limits, is proposed, and 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.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a marge of safety?
    Response: No.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
or actions. There is no change in the design of the affected 
systems, no alteration of the setpoints at which alarms or actions 
are initiated, and no change in plant configuration from original 
design. With one of the required instrument buses being powered from 
the alternate class 1E power supply, there is no significant 
reduction in the margin of safety. Testing of the DGs and associated 
electrical distribution equipment provides confidence that the DGs 
will start and provide power to the associated

[[Page 32073]]

equipment in the unlikely event of a LOOP during the extended 7-day 
Completion Time.
    Applicable regulatory requirements will continue to be met, 
adequate defense-in-depth will be maintained, sufficient safety 
margins will be maintained, and any increases in risk are small and 
consistent with the NRC Safety Goal Policy Statement (Federal 
Register, Vol. 51, p. 30028 (51 FR 30028), August 4, 1986, as 
interpreted by NRC Regulatory Guides 1.174 and1.177). Furthermore, 
increases in risk posed by potential combinations of equipment out 
of service during the proposed NSPS inverter extended Completion 
Time will be managed under a configuration risk management program 
(CRMP) consistent with 10CFR50.65, ``Requirements for monitoring the 
effectiveness of maintenance at nuclear power plants.'', paragraph 
(a)(4).
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60666.
    NRC Section Chief: Anthony J. Mendiola.

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: April 30, 2004.
    Description of amendment request: The proposed amendments would 
incorporate the oscillation power range monitor (OPRM) instrumentation 
into the technical specifications (TS). The proposed changes would 
revise: (1) TS 3.3.1.3, ``Oscillation Power Range Monitor (OPRM) 
Instrumentation,'' to insert a new TS section for the OPRM 
instrumentation, (2) TS 3.4.1, ``Recirculation Loops Operating,'' to 
delete the current thermal hydraulic instability administrative 
requirements, and (3) TS 5.6.5, ``Core Operating Limits Report 
(COLR),'' to add the appropriate references for the OPRM trip setpoints 
and methodology.
    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 proposed change has no impact on any of the 
existing neutron monitoring functions.
    Activation of the OPRM scram function will replace the current 
methods that require operators to insert an immediate manual reactor 
scram in certain reactor operating regions where thermal hydraulic 
instabilities could potentially occur. While these regions will 
continue to be avoided during normal operation, certain transients, 
such as a reduction in reactor recirculation flow, could place the 
reactor in these regions. During these transient conditions, with 
the OPRM instrumentation scram function activated, an immediate 
manual scram will no longer be required. This 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 
instability 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 
operation actions. Operators will continue to monitor for 
indications of thermal hydraulic instability when the reactor is 
operating in regions of potential instability as a backup to the 
OPRM instrumentation. 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 replace procedural actions 
that were established to avoid operating conditions where reactor 
instabilities might occur with an NRC approved automatic detect and 
suppress function (i.e., OPRM).
    Potential failure in the OPRM trip function could result in 
either a 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 effects of failures 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, 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 OPRM trip function is being implemented to 
automate the detection and subsequent suppression 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. Since the MCPR 
Safety Limit will not be exceeded as a result of an instability 
event following implementation of the OPRM trip function, it is 
concluded that the proposed change does not reduce the margin of 
safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: April 13, 2004.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications to maintain hydrogen 
recombiners and hydrogen and oxygen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI, Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the technical specifications (TS) for nuclear power reactors currently 
licensed to operate. The revised 10 CFR 50.44, ``Standards for 
combustible gas control system in light-water-cooled power reactors,'' 
eliminated the requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The

[[Page 32074]]

licensee affirmed the applicability of the model NSHC determination in 
its application dated March 4, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG 1.97 Category 1, is intended for key variables that 
most directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen and oxygen monitors no 
longer meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44 the Commission found that Category 
3, as defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents. Also, as part of the 
rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 2, as defined in RG 1.97, is an appropriate categorization 
for the oxygen monitors, because the monitors are required to verify 
the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
classification of the oxygen monitors as Category 2 and removal of 
the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs, 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 and oxygen monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

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

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

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

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

    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: May 5, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for instrumentation setpoints, 
allowable values, and calibration requirements based on updated 
calculations and reviews, and add a definition of ``annual'' frequency 
for use in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their 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 new DVR [degraded voltage relay] voltage and 
minimum time delay Allowable Values are more restrictive than the 
existing TS limits. The proposed new DVR maximum time delay is based 
on the existing analytical limit, and is only increased to the 
extent permitted by the methods endorsed by Regulatory Guide (RG) 
1.105. Annual channel calibrations are already performed, and adding 
them to TS ensures from a regulatory perspective that the relay 
drift is consistent with the setpoint calculations. The proposed new 
LVR [loss of voltage relay] voltage upper Allowable Value is based 
on a comprehensive EDG [emergency diesel generator] transient 
analysis, and is only increased to the extent permitted by the 
methods endorsed by Regulatory Guide (RG) 1.105. The proposed new 
LVR time delay allowable values are more restrictive than the 
existing TS limits, and are within the existing TS range of 
allowable values. Accident initial conditions, probability, and 
assumptions remain as previously analyzed. The remaining portions of 
the amendment request are administrative changes that will have no 
effect on operations of the relays. The Degraded Voltage and Loss of 
Voltage Relays are not

[[Page 32075]]

accident initiators; therefore, a malfunction of these relays will 
have no significant effect on accident initiation frequency. The 
proposed changes do not invalidate the assumptions used in 
evaluating the radiological consequences of any accident. Therefore, 
the proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed new DVR voltage and minimum time delay 
Allowable Values are more restrictive than the existing TS limits. 
The proposed new DVR maximum time delay is based on the existing 
analytical limit, and is only increased to the extent permitted by 
the methods endorsed by Regulatory Guide (RG) 1.105. Annual channel 
calibrations are already performed, and adding them to TS ensures 
from a regulatory perspective that the relay drift is consistent 
with the setpoint calculations. The proposed new LVR voltage upper 
Allowable Value is based on a comprehensive EDG transient analysis, 
and is only increased to the extent permitted by the methods 
endorsed by Regulatory Guide (RG) 1.105. The proposed new LVR time 
delay allowable values are more restrictive than the existing TS 
limits, and are within the existing TS range of allowable values. 
Accident initial conditions and assumptions remain as previously 
analyzed. The remaining portions of the amendment request are 
administrative changes that will have no effect on operations of the 
relays.
    The proposed changes do not introduce any new or different 
accident initiators. 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. The proposed changes to the DVR Allowable Values will ensure 
an adequate margin of safety is maintained between the lowest 
allowable voltage setpoint and the highest per unit voltage required 
by safety-related equipment, while at the same time establishing an 
Allowable Value, not previously provided, that ensures a sufficient 
margin of safety between the highest allowable voltage setpoint and 
the lowest expected per unit source voltages.
    The proposed changes to the DVR Allowable Values will ensure an 
adequate margin of safety is maintained between the longest 
allowable time delay and the longest time delay assumed by the 
accident analyses, while at the same time establishing an Allowable 
Value, not previously provided, that ensures a sufficient margin of 
safety between the shortest allowable time delay and the longest 
acceleration time for 4160 Volt continuously energized Safety 
Features Actuation System motors.
    The proposed new LVR voltage upper Allowable Value is based on a 
comprehensive EDG transient analysis, and is only increased to the 
extent permitted by the methods endorsed by Regulatory Guide (RG) 
1.105. In addition, the new Allowable Value reflects improvements in 
channel uncertainties that were made possible by upgrading the 
relays to solid state units.
    The proposed new LVR time delay allowable values are more 
restrictive than the existing TS limits, and are within the existing 
TS range of allowable values.
    A new requirement to perform an annual channel calibration of 
the Degraded Voltage and Loss of Voltage Relays is proposed. This 
new requirement to demonstrate proper channel operations will not 
adversely affect a margin of safety. The remaining changes are 
administrative, and will have no effect on margin of safety. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: May 21, 2004.
    Description of amendment request: The proposed amendment would 
revise the following in the technical specifications (TSs): (1) adding 
a new figure (Figure 2-3) to the table of contents that shows the 
volume of Trisodium Phosphate (TSP) required over the operating cycle; 
(2) Section 2.3(4), ``Emergency Core Cooling System--Trisodium 
Phosphate (TSP),'' regarding volume and form of TSP; and (3) Section 
3.6(2)d.(i), ``Safety Injection and Containment Cooling Systems 
Tests,'' regarding the surveillance requirement for TSP volume. The 
amendment also proposes modifications to the corresponding Basis of TS 
2.3 and TS 3.6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    There are no changes to the design or operation of the plant 
that could affect system, component, or accident functions as a 
result of deleting the requirement for the ``dodecahydrate'' form of 
TSP, or replacing the volume of active TSP required during Operating 
Modes 1 and 2 with an amount dependent upon HZP [hot zero power] CBC 
[critical boron concentration] as shown in Figure 2-3. All systems 
and components function as designed and the performance requirements 
have been evaluated and found to be acceptable. Hydrated TSP in the 
range of 45-57% moisture content will maintain pH >= 7.0 in the 
recirculation water following a LOCA [loss-of-coolant accident]. 
This function is maintained with the proposed change. Allowing the 
required volume of active TSP to decrease over the operating cycle 
as HZP CBC decreases will ensure that the pH of the containment sump 
is >= 7.0 yet provides additional margin for EEQ [equipment 
environmental qualification] concerns as containment sump pH is less 
likely to exceed 7.5.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed change. All 
systems, structures, and components previously required for 
mitigation of an event remain capable of fulfilling their intended 
design function with this change to the TS. The proposed change has 
no adverse effects on any safety-related systems or component and 
does not challenge the performance or integrity of any safety 
related system. The proposed change has evaluated the TSP 
configuration such that no new accident scenarios or single failures 
are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Deleting the requirement for the ``dodecahydrate'' form of TSP 
and allowing the required volume of active TSP to decrease as HZP 
CBC decreases still ensures that the pH of the containment sump is 
>= 7.0. Hydrated TSP in the range of 45-57% moisture content will 
maintain pH >= 7.0 in the recirculation water following a LOCA. This 
change provides additional margin for EEQ concerns as containment 
sump pH is less likely to exceed 7.5. Therefore, this change does 
not involve a significant reduction in the margin of safety. 
Evaluations were made that indicate that the margin for pH control 
is not altered by the proposed changes. A TSP volume that is 
dependent on HZP CBC has been evaluated with respect to 
neutralization of all borated water and acid sources. These 
evaluations concluded that there would be no impact on pH control, 
and hence no reduction in the margin of safety related to post LOCA 
conditions.
    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

[[Page 32076]]

review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: March 31, 2004.
    Description of amendment request: The proposed amendment would 
revise the reactor pressure vessel pressure-temperature limits and 
extend the validity of the limits to 32 effective full power years.
    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 revised curves are based on uprated fluence projections and 
are applicable for the service period up to 32 effective full power 
years (EFPY). There are no changes being made to the reactor coolant 
system (RCS) pressure boundary or to RCS material, design or 
construction standards. The proposed heatup and cooldown curves 
define limits that continue to ensure the prevention of nonductile 
failure of the RCS pressure boundary. The design-basis events that 
were evaluated have not changed. The modification of the heatup and 
cooldown curves does not alter any assumptions previously made in 
the radiological consequence evaluations since the integrity of the 
RCS pressure boundary is unaffected. Therefore, the proposed changes 
will not significantly increase the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Revisions to the heatup and cooldown curves do not involve any 
new components or plant procedures. The proposed changes do not 
create any new single failure or cause any systems, structures, or 
components to be operated beyond their design bases. Therefore, the 
proposed license amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed figures define the limits for ensuring prevention 
of nonductile failure for the reactor coolant system based on the 
methods described in 1989 ASME Code [American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code] Section XI Appendix G, 10 
CFR 50 Appendix G, and ASME Code Cases N-640 and N-588. The effect 
of the change is to permit plant operation within different 
pressure-temperature limits, but still with adequate margin to 
assure the integrity of the reactor coolant system pressure 
boundary. 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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.

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

    Date of application for amendments: December 15, 2003.
    Brief Description of amendments: The amendments revise Technical 
Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain 
Valves,'' for the condition of having one or more SDV vent or drain 
lines with one valve inoperable.
    Date of issuance: May 17, 2004.
    Effective date: May 17, 2004.
    Amendment Nos.: 232 and 259.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12364).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 17, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 19, 2003, supplements 
dated October 23, 2003, and January 28, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specifications to modify the requirements for the containment 
pressure control system to eliminate a problem with circuit fluctuation 
as a result of electronic noise.
    Date of issuance: May 12, 2004.

[[Page 32077]]

    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 214 and 208.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54749).
    The supplements dated October 23, 2003, and January 28, 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 as published in the Federal Register on 
September 18, 2003 (68 FR 54749).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 12, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 14, 2002, as supplemented by letters 
dated June 27, 2002, July 9, 2003, and April 7 and May 12, 2004.
    Brief description of amendment: The amendment revises the Updated 
Safety Analysis Report (USAR) Appendix 3B and Sections 6.2.1.1.3.2.1, 
``Reactor Water Cleanup Break'' and 6.2.1.2 ``Containment 
Subcompartments'' to change the method of analysis for high energy line 
breaks inside and outside of containment. The change will replace the 
current THREED code for room pressure-temperature analyses with the 
GOTHIC (Generation of Thermal-Hydraulic Information for Containments) 
code.
    Date of issuance: May 20, 2004.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 139.
    Facility Operating License No. NPF-47: The amendment revised the 
USAR Appendix 3B and Sections 6.2.1.1.3.2.1 and 6.2.1.2.2.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45563). The June 27, 2002, July 9, 2003, and April 7 and May 12, 2004, 
supplemental letters provided clarifying information that did not 
expand the scope of the original Federal Register notice or the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 20, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 12, 2003, as supplemented by 
letter dated April 22, 2004.
    Brief description of amendment: The amendment changes the heater 
acceptance criteria contained in surveillance requirements 4.6.6.1d.5, 
4.7.6.1d.3, and 4.7.7d.4, performed to verify that the heat dissipated 
by the heaters is within a given band, for the shield building 
ventilation, control room ventilation, and controlled ventilation area 
systems, respectively. The changes increase the upper limit of the 
acceptance criteria from rated capacity plus 5 percent to rated 
capacity plus 10 percent and without any change for the lower limit of 
the band of rated capacity minus 10 percent.
    Date of issuance: May 24, 2004.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 194.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59217). The April 22, 2004, supplemental letter 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 amendment is contained 
in a Safety Evaluation dated May 24, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 28, 2004, as 
supplemented May 3, 2004
    Brief description of amendments: The amendments eliminated the 
requirements in BVPS-1 and BVPS-2 Technical Specifications (TSs) 
associated with hydrogen recombiners and relocate the requirements for 
hydrogen monitors to the Licensing Requirements Manuals.
    Date of issuance: May 19, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 120 days.
    Amendment Nos.: 259 and 142.
    Facility Operating License Nos. DPR-66 and NPF-73: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12370). The supplement dated May 3, 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 as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 19, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 14, 2003, as supplemented 
November 20, 2003, March 25, 2004, and April 27, 2004.
    Brief description of amendment: The amendment allows a one-time 
increase in the completion time for restoring an inoperable nuclear 
services seawater system train to operable status.
    Date of issuance: May 18, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 212.
    Facility Operating License No. DPR-72: Amendment revises the 
License and Technical Specifications.
    Date of initial notice in Federal Register: August 5, 2003 (68 FR 
42644). The November 20, 2003, March 25, 2004, and April 27, 2004, 
supplements contained clarifying information only and did not change 
the initial no significant hazards consideration determination or 
expand the scope of the initial application. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
May 18, 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: September 26, 2003, as 
supplemented January 7, 2004.
    Brief description of amendments: The amendments modify Technical 
Specifications (TS) requirements to adopt the provisions of Industry/TS

[[Page 32078]]

Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode 
Restraints.''
    Date of issuance: May 12, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: Unit 1-169; Unit 2-170.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68671). The supplemental letter dated January 7, 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 May 12, 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 amendment: July 31, 2002 (superseded 
November 24, 1999, application) and its supplements dated August 15 and 
December 23, 2003.
    Brief description of amendments: The amendments revise the 
technical specifications to relocate the pressure-temperature limits 
and low temperature overpressure protection system limit setpoints into 
a plant-specific pressure temperature limits report that will be 
administratively controlled by the technical specifications.
    Date of issuance: May 13, 2004.
    Effective date: May 13, 2004, and shall be implemented within 30 
days from the date of issuance.
    Amendment No.: Unit 1-170; Unit 2-171.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58648). The supplemental letters dated August 15 and December 23, 
2003, 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. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 13, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 2, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and 
Drain Valves,'' for the condition of having one or more SDV vent or 
drain lines with one valve inoperable.
    Date of issuance: May 25, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 240 and 183.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12372). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 25, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 23, 2004, as supplemented 
April 30, 2004.
    Brief description of amendment: The amendments allow both trains of 
control room air-conditioning system (CRACS) to be inoperable for up to 
7 days provided control room temperatures are verified every 4 hours to 
be less than or equal to 90 degrees Fahrenheit. If this temperature 
limit cannot be maintained or both CRACS trains are inoperable for more 
than seven days, the requirements of Technical Specification Section 
3.0.3 must be implemented.
    Date of issuance: May 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 292 and 282.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the technical specifications.
    Date of initial notice in Federal Register: April 14, 2004 (69 FR 
19880). The April 30, 2004, letter provided clarifying information that 
did not expand the scope of the original application or change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated May 21, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 13, 2002, as 
supplemented by letters dated May 8, 2003, December 17, 2003, February 
12, 2004, and March 9, 2004.
    Brief description of amendment: These amendments revise the 
completion time of Required Action A.1 of Technical Specification 
3.8.7, ``Inverters--Operating,'' from 24 hours to 7 days for an 
inoperable instrument bus inverter.
    Date of issuance: May 12, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 235 and 217.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18289). The May 8, 2003, December 17, 2003, February 12, 2004, and 
March 9, 2004, supplementary letters contained clarifying information 
only and did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the initial 
application. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 12, 2004.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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,

[[Page 32079]]

which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Registermedia to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 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 Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to pdr@nrc.gov.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1-800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, 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 identify 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 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
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 flaw or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment

[[Page 32080]]

under consideration. The contention must be one which, if proven, would 
entitle the petitioner 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.
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    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
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    Each contention shall be given a separate numeric or alpha 
designation within on of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer or 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).

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

    Date of amendment request: May 7, 2004.
    Brief description of amendment: The amendment restores the licensed 
thermal power from 1524 megawatts thermal (MWt), as approved in 
Amendment No. 224, to the previous value of 1500 MWt.
    Date of issuance: May 14, 2004.
    Effective date: May 14, 2004.
    Amendment No.: 227.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Operating License and the Technical Specifications. Public 
comments requested as to proposed no significant hazards consideration 
(NSHC): Yes. Omaha-World Herald. The notice provided an opportunity to 
submit comments on the Commission's proposed NSHC determination. No 
comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, State consultation, and final NSHC determination 
are contained in a safety evaluation dated May 14, 2004.
    Attorney for licensee: James R. Curtiss, Esq. Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

    Dated at Rockville, Maryland, this 28th day of May, 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-12671 Filed 6-7-04; 8:45 am]
BILLING CODE 7590-01-P