[Federal Register Volume 69, Number 91 (Tuesday, May 11, 2004)]
[Notices]
[Pages 26184-26197]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-10305]


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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, April 16 through April 29, 2004. The last 
biweekly notice was published on April 27, 2004 (69 FR 22877).

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 26185]]

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 26186]]

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: March 23, 2004.
    Description of amendment request: The licensee requested to revise 
the Technical Specifications (TSs), deleting the requirements for the 
Independent Onsite Safety Review Group (IOSRG) and locating them intact 
to a licensee-controlled document, the company-wide Quality Assurance 
Topical Report (QATR). The requirements are in the administrative 
section of the TSs and include IOSRG organization, function 
description, member qualifications, and recordkeeping. The relocation 
is proposed per the guidance of Nuclear Regulatory Commission (NRC) 
Administrative Letter 95-06. In addition, the licensee proposed to 
correct the reference for facility activities audits from a site-
specific document to the company-wide QATR.
    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 reviewed the licensee's analysis and has 
performed its own analysis as follows:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendment does not affect assumptions contained in 
the current licensing basis plant safety analyses, will not lead to 
physical changes of a plant structure, system, or component (SSC), and 
will not alter the method of operation of any SSC. The IOSRG 
requirements and conduct of IOSRG activities were not factors in any 
previously analyzed accident or transient scenarios, and thus, the 
elimination of IOSRG requirements from the TSs will have no effect on 
the probability of occurrence and consequences of any previously 
analyzed accident or transient.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed amendment is not the result of a design change or 
method of operation change, and will not lead to such changes. Hence 
no, new or different kind of accident can be created from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed amendment does not involve any change to current 
analysis models, assumptions, limiting conditions for operation, 
operational parameters, action statements, and surveillance 
requirements. Hence, there is no reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
its own analysis, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the proposed amendment involves no significant hazards 
consideration.
    Attorney for licensee: 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.

    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 23, 2004.
    Description of amendment request: The licensee requested to revise 
the Technical Specifications (TSs), deleting the requirements for the 
Independent Onsite Safety Review Group (IOSRG) and locating them intact 
to a licensee-controlled document, the company-wide Quality Assurance 
Topical Report (QATR). The requirements are in the administrative 
section of the TSs and include IOSRG organization, function 
description, member qualifications, and recordkeeping. The relocation 
is proposed per the guidance of Nuclear Regulatory Commission (NRC) 
Administrative Letter 95-06. In addition, the licensee proposed to 
correct the reference for facility activities audits from a site-
specific document to the company-wide QATR.
    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 reviewed the licensee's analysis and has 
performed its own analysis as follows:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendment does not affect assumptions contained in 
the current licensing basis plant safety analyses, will not lead to 
physical changes of a plant structure, system, or component (SSC), and 
will not alter the method of operation of any SSC. The IOSRG 
requirements and conduct of IOSRG activities were not factors in any 
previously analyzed accident or transient scenarios, and thus, the 
elimination of IOSRG requirements from the TSs will have no effect on 
the probability of occurrence and consequences of any previously 
analyzed accident or transient.

[[Page 26187]]

    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed amendment is not the result of a design change or 
method of operation change, and will not lead to such changes. Hence, 
no new or different kind of accident can be created from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed amendment does not involve any change to current 
analysis models, assumptions, limiting conditions for operation, 
operational parameters, action statements, and surveillance 
requirements. Hence, there is no reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
its own analysis, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the proposed amendment involves no significant hazards 
consideration.
    Attorney for licensee: 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.

    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: March 4, 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 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 severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen 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

[[Page 26188]]

reactor core conditions and the direction of degradation while 
effectively responding to the event in order to mitigate the 
consequences of the accident. The intent of the requirements 
established as a result of the TMI, Unit 2 accident can be 
adequately met without reliance on safety-related hydrogen monitors. 
Category 2 oxygen monitors are adequate to verify the status of an 
inerted containment.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TS will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: April 15, 2004.
    Description of amendment request: The proposed amendment would 
change the reactor coolant system (RCS) pressure/temperature (P/T) 
limits in the technical specifications (TSs) by providing a single 
maximum cooldown rate instead of a variable cooldown rate and by 
revising the cooldown curve with one that is slightly more restrictive.
    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 occurrence of an accident previously 
evaluated for ANO-2 [Arkansas Nuclear One, Unit 2] is not altered by 
the proposed amendment to the TSs. The accidents remain the same as 
currently analyzed in the ANO-2 Safety Analysis Report (SAR) as a 
result of the change to the cooldown P/T limits. The new P/T 
cooldown limits were based on NRC [Nuclear Regulatory Commission] 
accepted methodologies along with ASME [American Society of 
Mechanical Engineers] Code [Boiler and Pressure Vessel Code] 
alternatives. The proposed change does not impact the integrity of 
the reactor coolant pressure boundary (RCPB) (i.e., there is no 
change to the operating pressure, materials, loadings, etc.) as a 
result of this change. In addition, there is no increase in the 
potential for the occurrence of a loss of coolant accident. The 
proposed P/T cooldown limit curve is not considered to be an 
initiator or contributor to any accident currently evaluated in the 
ANO-2 SAR. The revised P/T cooldown limits ensure the long term 
integrity of the RCPB. For each analyzed transient and steady state 
condition, the allowable pressure was determined as a function of 
reactor coolant temperature considering postulated flaws in the 
reactor vessel beltline, inlet nozzle, outlet nozzle, and closure 
head flange.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the P/T limits will not create a new 
accident scenario. The requirements to have P/T protection are part 
of the ANO-2 licensing basis. The proposed change in the P/T 
cooldown limits is based on NRC approved methodologies performed by 
Framatome ANP. This methodology complies with NRC and ASME 
requirements for protecting the RCS. Therefore, the revised P/T 
cooldown limits provide protection of the RCS from limiting 
transients during normal cooldown.
    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 revision of the P/T limits and curves will ensure that ANO-2 
continues to operate within the operating margins of the ASME Code. 
The application of ASME Code Cases N-640 and N-588 presents 
alternative procedures for calculating P/T temperatures and 
pressures. These Code Cases allow certain assumptions to be 
conservatively reduced. However, the procedures allowed by these 
Code Cases still provide sufficient conservatism and ensure an 
adequate margin of safety in the development of P/T operating and 
pressure test limits to prevent non-ductile fractures.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: March 31, 2004.
    Description of amendment request: This license amendment request 
(LAR) proposes to eliminate the Technical Specification Surveillance 
Requirements (SRs) that require each Main Steam Safety/Relief Valve (S/
RV) to open during the manual actuation portion of testing the valves. 
In accordance with 10 CFR 50.55a, ``Codes and Standards,'' paragraph 
(a)(3), this request also includes Relief Request VR-13. VR-13 is a 
request for relief from the requirements of ASME/American National 
Standards Institute (ANSI) Operation and Maintenance (OM) of Nuclear 
Power Plants, OM-1995, Appendix I, Section 3.4.1(d) that after 
isolation, the S/RVs are manually opened and closed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed [License Amendment Request] LAR modifies TS 
3.4.4.3, SR 3.5.1.7, and SR 3.6.1.6.1 to allow the uncoupling of the 
S/RV stem from the S/RV actuator during manual actuation. The 
proposed LAR does not change the manner in which the S/RVs are 
intended to operate.
    The performance of S/RV testing provides assurance that the S/
RVs are capable of depressurizing the Reactor Pressure Vessel (RPV). 
This will protect the RPV from over pressurization and allows the 
combination of the Low Pressure Coolant Injection (LPCI) system and 
the Low Pressure Core Spray (LPCS) system to inject into the RPV as 
designed. The proposed testing requirements are sufficient to 
provide confidence that the S/RVs, [Automatic Depressurization 
System] ADS valves, and the [Low-Low Set] LLS valves will perform 
their intended design safety functions.
    Therefore, the proposed LAR does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed LAR changes TS [Surveillance Requirements] SR 
3.4.4.3, SR 3.5.1.7, and SR 3.6.1.6.1. The changes to these SRs do 
not effect the assumed accident performance of the S/RVs, nor any 
plant structure, system or component previously evaluated. The LAR 
does not install any new equipment, nor does it cause existing 
equipment to be operated in a new or

[[Page 26189]]

different manner. The S/RVs continue to be bench-tested to verify 
the safety and relief modes of valve operation. The changes will 
allow the testing of the manual actuation electrical circuitry, 
solenoid and air control valve, and the actuator without causing the 
S/RV to open. No setpoints are being changed which would alter the 
dynamic response of plant equipment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from that previously evaluated.
    3. The proposed change will not involve a single reduction in 
the margin of safety.
    The proposed LAR will allow the uncoupling of the S/RV stem from 
the other components associated with the manual actuation testing of 
the S/RVs. The proposed changes will allow the testing of the manual 
actuation electrical circuitry, solenoid and air control valve, and 
the actuator without causing the S/RV to open. The S/RVs will 
continue to be manually actuated by the bench-test of the valve 
control system and setpoint testing program prior to installation in 
the plant. The changes do not effect the valve setpoint or 
operational criteria that directs the S/RVs to be manually opened 
during plant transients. There are no changes which alter the 
setpoints at which protective actions are initiated.
    Therefore, the proposed change does not involve a significant 
reduction in any margins 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.
    FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry 
Nuclear Power Plant (PNPP), Unit 1, Lake County, Ohio
    Date of amendment request: April 5, 2004.
    Description of amendment request: This license amendment request 
(LAR) proposes to modify the existing Minimum Critical Power Ratio 
(MCPR) Safety Limit contained in Technical Specification 2.1.1.2. 
Specifically, the change modifies the MCPR Safety Limit values, as 
calculated by Global Nuclear Fuel (GNF), by decreasing the limit for 
two recirculating loop operation from 1.10 to 1.08, and decreasing the 
limit for single recirculation loop operation from 1.11 to 1.10. The 
change resulted from a core reload analysis performed during the PNPP 
Fuel Cycle 10.
    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.
    Perry Nuclear Power Plant (PNPP) Updated Safety Analysis Report 
(USAR) Section 4.2, ``Fuel System Design,'' states the PNPP fuel 
system design bases are provided in the General Electric Topical 
Report, NEDE-24011-P-A, ``General Electric Standard Application for 
Reactor Fuel (GESTAR II).'' The Minimum Critical Power Ratio (MCPR) 
Safety Limit is one of the limits used to protect the fuel in 
accordance with the design basis. The MCPR Safety Limit establishes 
a margin to the onset of transition boiling. The basis of the MCPR 
Safety Limit remains the same, ensuring that greater than 99.9 % of 
all fuel rods in the core avoid transition boiling. The methodology 
used to determine the MCPR Safety Limit values is contained within 
GESTAR II and is NRC approved. The change does not result in any 
physical plant modifications or physically affect any plant 
components. As a result, there is no increase in the probability of 
occurrence of a previously analyzed accident.
    The fundamental sequences of accidents and transients have not 
been altered. The Safety Limit MCPR is established to avoid fuel 
damage in response to anticipated operational occurrences. 
Compliance with a MCPR Safety Limit greater than or equal to the 
calculated value will ensure that less than 0.1% of the fuel rods 
will experience boiling transition. This in turn ensures fuel damage 
does not occur following transients due to excessive thermal 
stresses on the fuel cladding. The MCPR Operating Limits are set 
higher (i.e., more conservative) than the Safety Limit such that 
potentially limiting plant transients prevent the MCPR from 
decreasing below the MCPR Safety Limit during the transient. 
Therefore, there is no impact on any of the limiting USAR Appendix 
15B transients. The radiological consequences remain the same as 
previously stated in the USAR. Therefore, the consequences of an 
accident do not increase over previous evaluations in the USAR.
    Therefore, the proposed LAR does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The MCPR Safety Limit basis is preserved, which is to ensure 
that transition boiling does not occur in at least 99% of the fuel 
rods in the core as a result of the postulated limiting transient. 
The values are calculated in accordance with GESTAR II. The GESTAR 
II analyses have been accepted by the NRC. The MCPR Safety Limit is 
one of the limits established to ensure the fuel is protected in 
accordance with the design basis. The function, location, operation, 
and handling of the fuel remain unchanged. No changes in the design 
of the plant or the method of operating the plant are associated 
with these revised safety limit valves. Therefore, no new or 
different kind of accident from any previously evaluated is created.
    3. The proposed change will not involve a single reduction in 
the margin of safety.
    This change revises the PNPP MCPR Safety Limit values. The new 
MCPR Safety Limit values reflect changes due to Cycle 10 core 
design, but do not alter the design or function of any plant system, 
including the fuel. The new MCPR Safety Limit values were calculated 
using NRC-approved methods described in GESTAR II. The proposed MCPR 
Safety Limit values continue to satisfy the fuel design safety 
criteria which ensures that transition boiling does not occur in at 
least 99.9% of the fuel rods in the core as a result of the 
postulated limiting transient. Therefore, the proposed values for 
the MCPR Safety Limit do not involve a significant reduction in a 
safety margin.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.
    FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry 
Nuclear Power Plant (PNPP), Unit 1, Lake County, Ohio
    Date of amendment request: April 5, 2004.
    Description of amendment request: This license amendment request 
(LAR) proposes to modify the existing Minimum Critical Power Ratio 
(MCPR) Safety Limit contained in Technical Specification 2.1.1.2. 
Specifically, the change modifies the MCPR Safety Limit values, as 
calculated by Global Nuclear Fuel (GNF), by decreasing the limit for 
two recirculating loop operation from 1.10 to 1.08, and decreasing the 
limit for single recirculation loop operation from 1.11 to 1.10. The 
change resulted from a core reload analysis performed during the PNPP 
Fuel Cycle 10.
    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.
    Perry Nuclear Power Plant (PNPP) Updated Safety Analysis Report 
(USAR) Section 4.2, ``Fuel System Design,'' states the PNPP fuel 
system design bases are provided in the General Electric Topical 
Report, NEDE-

[[Page 26190]]

24011-P-A, ``General Electric Standard Application for Reactor Fuel 
(GESTAR II).'' The Minimum Critical Power Ratio (MCPR) Safety Limit 
is one of the limits used to protect the fuel in accordance with the 
design basis. The MCPR Safety Limit establishes a margin to the 
onset of transition boiling. The basis of the MCPR Safety Limit 
remains the same, ensuring that greater than 99.9 % of all fuel rods 
in the core avoid transition boiling. The methodology used to 
determine the MCPR Safety Limit values is contained within GESTAR II 
and is NRC approved. The change does not result in any physical 
plant modifications or physically affect any plant components. As a 
result, there is no increase in the probability of occurrence of a 
previously analyzed accident.
    The fundamental sequences of accidents and transients have not 
been altered. The Safety Limit MCPR is established to avoid fuel 
damage in response to anticipated operational occurrences. 
Compliance with a MCPR Safety Limit greater than or equal to the 
calculated value will ensure that less than 0.1% of the fuel rods 
will experience boiling transition. This in turn ensures fuel damage 
does not occur following transients due to excessive thermal 
stresses on the fuel cladding. The MCPR Operating Limits are set 
higher (i.e., more conservative) than the Safety Limit such that 
potentially limiting plant transients prevent the MCPR from 
decreasing below the MCPR Safety Limit during the transient. 
Therefore, there is no impact on any of the limiting USAR Appendix 
15B transients. The radiological consequences remain the same as 
previously stated in the USAR. Therefore, the consequences of an 
accident do not increase over previous evaluations in the USAR.
    Therefore, the proposed LAR does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The MCPR Safety Limit basis is preserved, which is to ensure 
that transition boiling does not occur in at least 99% of the fuel 
rods in the core as a result of the postulated limiting transient. 
The values are calculated in accordance with GESTAR II. The GESTAR 
II analyses have been accepted by the NRC. The MCPR Safety Limit is 
one of the limits established to ensure the fuel is protected in 
accordance with the design basis. The function, location, operation, 
and handling of the fuel remain unchanged. No changes in the design 
of the plant or the method of operating the plant are associated 
with these revised safety limit valves. Therefore, no new or 
different kind of accident from any previously evaluated is created.
    3. The proposed change will not involve a single reduction in 
the margin of safety.
    This change revises the PNPP MCPR Safety Limit values. The new 
MCPR Safety Limit values reflect changes due to Cycle 10 core 
design, but do not alter the design or function of any plant system, 
including the fuel. The new MCPR Safety Limit values were calculated 
using NRC-approved methods described in GESTAR II. The proposed MCPR 
Safety Limit values continue to satisfy the fuel design safety 
criteria which ensures that transition boiling does not occur in at 
least 99.9% of the fuel rods in the core as a result of the 
postulated limiting transient. Therefore, the proposed values for 
the MCPR Safety Limit do not involve a significant reduction in a 
safety margin.

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

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

    Date of amendment requests: August 27, 2003.
    Description of amendment requests: The proposed amendments would 
amend Unit 1 and Unit 2 Technical Specifications (TS) 4.0.3. TS 4.0.3 
describes the relationship between meeting the surveillance requirement 
and operability. The proposed change will modify TS 4.0.3 to allow a 
missed surveillance to be completed within 24 hours or up to the limit 
of the specified interval, whichever is greater. Additionally, a 
statement that a risk evaluation shall be performed for any 
surveillance delayed greater than 24 hours and that the risk impact 
shall be managed is being added to the TS.
    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?
    No.
    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
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. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this 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 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. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. The 
format changes are intended to improve readability and appearance 
and do not alter any requirements. Thus, this change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No.
    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the limiting condition for 
operation is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function. The format changes are 
intended to improve readability and appearance and do not alter any 
requirements. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    Based upon the reasoning presented above, the requested change 
does not involve a significant hazards consideration.


[[Page 26191]]


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment requests: February 14, 2004.
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TS) governing containment 
penetrations and the Containment Purge and Exhaust Isolation System, 
which are applicable during CORE ALTERATIONS and movement of irradiated 
fuel, such that those TSs are only applicable during the movement of 
recently irradiated fuel.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes incorporate line item improvements that are 
based on assumptions in the postulated fuel handling accident (FHA) 
analysis. These proposed changes remove the applicability of the 
Technical Specifications (TS) governing containment penetrations and 
the Containment Purge and Exhaust Isolation System when handling 
fuel assemblies that have decayed for a sufficient period of time. 
The containment penetration and Containment Purge and Exhaust 
Isolation System are not initiators to any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The only previously 
analyzed accident affected by the proposed change is an FHA. The 
current, Nuclear Regulatory Commission (NRC)-approved analysis of an 
FHA does not assume any holdup of the postulated radioactivity 
release by the containment building nor does it assume the operation 
of the Containment Purge and Exhaust Isolation System. As a result, 
the proposed change does not affect the assumed mitigation or 
consequences of that event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes incorporate line item improvements that are 
based on assumptions in the postulated FHA analysis. These proposed 
changes remove the applicability of the TS governing containment 
penetrations and the Containment Purge and Exhaust Isolation System 
when handling fuel assemblies that have decayed for a sufficient 
period of time. The proposed changes do not involve the addition or 
modification of equipment nor do they alter the design of the plant. 
The revised operations are consistent with the FHA analysis and do 
not require any new or different ways of operating the plant 
equipment.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes incorporate line item improvements that are 
based on assumptions in the postulated FHA analysis. These proposed 
changes remove the applicability of the TS governing containment 
penetrations and the Containment Purge and Exhaust Isolation System 
when handling fuel assemblies that have decayed for a sufficient 
period of time. The calculated offsite and Control Room doses 
resulting from an FHA are not affected by this change as the 
proposed TS changes are revised to be consistent with the 
assumptions used in these analyses. As a further measure, [Indiana 
Michigan Power Company] I&M has committed to maintaining a single 
normal or contingency method to promptly close containment 
penetrations following an FHA. These prompt methods will enable the 
ventilation systems to draw the release from a postulated FHA such 
that it can be treated and monitored. This will provide a further 
margin of safety beyond that assumed in the accident analysis.
    Therefore, the proposed change 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 requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107
    NRC Section Chief: L. Raghavan.

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

    Date of amendment requests: February 14, 2004.
    Description of amendment requests: The proposed amendments would 
modify the Technical Specification (TS) 3.9.2 limiting condition for 
operation, to delete TS Surveillance Requirements (SRs) 4.9.2.a and b 
for the Source Range Neutron Flux Monitor channel functional test, to 
revise SR 4.9.2.c for the channel check test, and to add a requirement 
to perform a channel calibration every 18 months as well as revise TS 
4.10.4.2 and 4.10.3.2 (Units 1 and 2 respectively) for Intermediate and 
Power Range channel functional test.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment replaces the Technical Specification (TS) 
3.9.2 limiting condition for operation (LCO) requirement for an 
audible indication in the containment (both units) and control room 
(Unit 2) with a requirement that a source range audible count rate 
circuit be operable. This involves no physical changes to the plant, 
and maintains the capability to alert the operators to changes in 
core reactivity. Thus, neither the probability of an accident nor 
the consequences are significantly increased.
    The proposed amendment revises the TS SR for the Power Range, 
Intermediate Range, and the Source Range Neutron Flux Monitors to 
reduce redundant testing. Surveillance testing is not an initiator 
to any accident previously evaluated. As a result, the proposed 
changes will not result in a significant increase in the probability 
of any accident previously evaluated.
    The Power Range, Intermediate Range, and the Source Range 
Neutron Flux Monitors are used to detect and mitigate accidents 
previously evaluated. However, the LCOs continue to require the 
subject flux monitors to be operable and the remaining testing is 
sufficient to ensure the flux monitors are capable of performing 
their detection and mitigation functions. Thus, the consequences of 
an accident are not significantly changed.
    Based on the above, [Indiana Michigan Power Company] I&M 
concludes that 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 and accident previously evaluated?
    Response: No.
    The proposed amendment replaces the TS 3.9.2 LCO requirement for 
an audible indication in the containment (both units) and control 
room (Unit 2) with a requirement that a source range audible count 
rate circuit be operable.

[[Page 26192]]

    The change does not make any physical changes to the plant. 
Thus, the change does not create the possibility of a new or 
different kind of accident.
    The proposed amendment revises the TS SR for the Power Range, 
Intermediate Range, and the Source Range Neutron Flux Monitors to 
reduce redundant testing. The proposed changes do not change the 
design function or operation of any plant equipment. No new failure 
mechanisms, malfunctions, or accident initiators are being 
introduced by the proposed changes. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment replaces the TS 3.9.2 LCO requirement for 
an audible indication in the containment (both units) and control 
room (Unit 2) with a requirement that a source range audible count 
rate circuit be operable. The source range audible count rate 
circuit will continue to perform its function of alerting the 
operators to changes in core reactivity.
    The proposed amendment revises the TS Surveillance Requirement 
(SR) for the Power Range, Intermediate Range, and the Source Range 
Neutron Flux Monitors to reduce redundant testing. The elimination 
of redundant testing does not reduce the reliability of the tested 
flux monitors. The flux monitors continue to be tested in a manner 
and at a frequency necessary to provide confidence that the 
equipment can perform its assumed safety function.
    Therefore, there is no significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107
    NRC Section Chief: L. Raghavan.

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

    Date of amendment requests: April 6, 2004.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) design features for fuel assemblies 
and new fuel storage criticality limitations. In addition, the licensee 
requests approval of the criticality analysis methodology supporting 
the spent fuel storage rack and new fuel storage rack in accordance 
with 10 CFR 50.59(c)(2)(viii).
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed Technical Specification (TS) changes allow the 
zirconium-based alloy, M5, to be used in addition to Zircaloy-4 and 
ZIRLO in Donald C. Cook Nuclear Plant fuel assemblies. TS changes 
are also proposed to allow Gadolinia to be used in fuel assemblies 
in the new fuel storage racks to ensure adequate reactivity margin. 
In addition, methodology changes were proposed for a criticality 
analysis supporting new and spent fuel rack design criteria. M5 is a 
Nuclear Regulatory Commission (NRC)-approved alloy for fuel cladding 
and Gadolinia is an NRC-approved fuel burnable absorber used in the 
maintenance of reactivity margin in the new fuel storage rack. The 
use of NRC-approved cladding and fuel absorbers and methodology 
changes to criticality analyses to support TS design criteria for 
the spent and new fuel storage racks are not initiators of any 
accident previously evaluated. As a result, the probability of any 
accident previously evaluated is not significantly increased. M5 
cladding has been shown to meet all 10 CFR 50.46 acceptance 
criteria. Analysis has shown that the use of Gadolinia assures 
sufficient reactivity margin to prevent a criticality accident in 
the new fuel storage rack. Changes in methodology for criticality 
analyses were performed to demonstrate TS requirements are met or to 
support proposed TS changes and do not affect plant equipment. 
Therefore, the consequences of an accident are not significantly 
increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability of occurrence or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to use the M5 alloy is based on an NRC-
approved topical report which demonstrates that the material 
properties of the M5 alloy are not significantly different from 
those of Zircaloy-4. The design and performance criteria continue to 
be met and no new failure mechanisms have been identified. 
Therefore, M5 fuel rod cladding and fuel assembly structural 
components will perform similarly to those fabricated from Zircaloy-
4, thus precluding the possibility of the fuel becoming an accident 
initiator and causing a new or different type of accident.
    The proposed TS change to use Gadolinia to ensure adequate 
reactivity margin for higher enrichment fuel assemblies prevents 
reactivity limits from being exceeded. An NRC-approved topical 
report demonstrates that Gadolinia is acceptable for use in fuel 
assemblies. The proposed change only modifies the type of fuel 
burnable absorber and does not affect any permanent plant equipment 
or plant operating procedures, and can not be an initiator of an 
accident.
    The proposed criticality analysis supports TS design criteria 
for spent and new fuel racks. The analysis evaluates reactivity 
margin based on conservative assumptions on fuel assembly design and 
burnup and does not affect any plant equipment. The criticality 
analysis can not be an initiator of an accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed TS change to allow the use of fuel rods clad with 
the M5 alloy does not change the reactor fuel reload design and 
safety limits. For each cycle reload core, the fuel assembly design 
and core configuration are evaluated using NRC-approved reload 
design methods, including consideration of the core physics analysis 
peaking factors and core average linear heat rate effects. The 
design basis and modeling techniques for fuel assemblies with 
Zircaloy-4 and ZIRLO clad fuel rods remain valid for fuel assemblies 
with M5 clad fuel rods. Use of the M5 alloy as cladding material has 
no effect on the criticality analysis for the spent fuel storage 
racks and the new fuel storage racks. Furthermore, it has no effect 
on the thermal-hydraulic and structural analysis for the spent fuel 
pool. Therefore, the design and safety analysis limits specified in 
the TS are maintained with this proposed change.
    The proposed TS change to use Gadolinia as a fuel burnable 
absorber for fuel assemblies with higher enrichments of Uranium-235 
to ensure proper reactivity control in the spent fuel storage rack 
is consistent with the current method of reducing reactivity of high 
enrichment fuel assemblies. Each method reduces the equivalent 
uranium enrichment to below that found acceptable by the NRC for 
safe storage of new fuel.
    The proposed criticality analyses use NRC-approved codes with a 
methodology different than previously approved by the NRC. The 
criticality analysis results for the spent fuel storage rack flooded 
with unborated water condition and for the new fuel storage rack 
moderated by aqueous foam condition remain less than the limiting TS 
values. Analysis results for the new fuel storage rack flooded with 
unborated water condition are consistent with previous analysis 
results.
    Therefore, the proposed change 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 requests involve no significant hazards consideration.

[[Page 26193]]

    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: April 13, 2004.
    Brief description of amendments: The requested amendments will 
revise the Technical Specification 3.3.2, ``Engineered Safety Features 
Actuation System (ESFAS) Instrumentation,'' to revise the trip setpoint 
allowable value for Refueling Water Storage Tank (RWST) Low-Low Level 
(ESFAS function 7.b) for Unit 2 to be the same as it is for Unit 1. 
Also, the frequency of calibration of the RWST water level transmitters 
will be revised from once in 9 months to once in 18 months.
    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 by focusing on the three standards set forth in 10 CFR 
50.92. The licensee's analysis of three standards is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change in the trip setpoint allowable value for 
Unit 2 Refueling Water Storage Tank (RWST) Low-Low Level has no 
impact on the probability of any accident previously evaluated. 
Since none of the accident analyses are affected by this change, the 
consequences of all previously evaluated accidents remain unchanged.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes. 
There are no changes in the method by which any safety-related plant 
system performs its safety function. Overall protection system 
performance will remain within the bounds of the previously 
performed accident analyses and the protection systems will continue 
to function in a manner consistent with the plant design basis. The 
proposed changes do not affect the probability of any event 
initiators. The proposed changes do not alter any assumptions or 
change any mitigation actions in the radiological consequence 
evaluations in the Final Safety Analysis Report (FSAR).
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. There will be no impact on the overpower limit, the 
Departure from Nucleate Boiling Ratio (DNBR) limits, the Heat Flux 
Hot Channel Factor (FQ), the Nuclear Enthalpy Rise Hot 
Channel Factor (F[Delta]H), the Loss of Coolant Accident Peak 
Centerline Temperature (LOCA PCT), peak local power density, or any 
other margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    Therefore the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri
    Date of amendment request: April 8, 2004.
    Description of amendment request: The proposed amendment revises TS 
5.5.7, ``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 April 8, 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 continued 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.
    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

[[Page 26194]]

(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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.

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

    Date of application request: April 8, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TSs would be 
eliminated, several notes or specific exceptions are revised or deleted 
to reflect the related changes to LCO 3.0.4, and Surveillance 
Requirement (SR) 3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated April 8, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant

[[Page 26195]]

to 10 CFR 51.22(b), no environmental impact statement or environmental 
assessment need be prepared for these amendments. If the Commission has 
prepared an environmental assessment under the special circumstances 
provision in 10 CFR 51.12(b) and has made a determination based on that 
assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to pdr@nrc.gov.

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

    Date of application for amendment: December 20, 2002, as 
supplemented on May 30, September 10, and November 3, 2003.
    Brief description of amendment: The amendment authorized the 
revision of the OCNGS Updated Final Safety Analysis Report (UFSAR) to 
reflect implementation of the Boiling Water Reactor Vessel and 
Internals Project reactor pressure vessel Integrated Surveillance 
Program (ISP) as the basis for demonstrating compliance with the 
requirements of Appendix H, ``Reactor Vessel Material Surveillance 
Program Requirements,'' to Title 10 of the Code of Federal Regulations, 
Part 50.
    Date of Issuance: April 27, 2004.
    Effective date: The amendment is effective immediately. The ISP 
shall be implemented prior to the next scheduled reactor vessel 
surveillance capsule removal. The UFSAR is to be revised to reflect use 
of the ISP in accordance with the schedule of 10 CFR 50.71(e).
    Amendment No.: 242.
    Facility Operating License No. DPR-16: Amendment revised the 
Operating License DPR-16.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5669).
    The May 30, September 10, and November 3, 2003, letters 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 Commission's related evaluation of 
this amendment is contained in a Safety Evaluation dated April 27, 
2004. No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 12, 2003, as supplemented 
December 5, 2003, February 23, 2004, March 26, 2004 and April 6, 2004.
    Brief description of amendments: These amendments extend several 
Required Action completion times for inoperable diesel generators 
identified in Technical Specification 3.8.1, ``AC Sources Operating.''
    Date of issuance: April 13, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 265 and 242.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2003 (68 FR 
37576). The licensee's December 5, 2003, February 23, 2004, March 26, 
2004, and April 6, 2004, letters provided additional information that 
clarified the application, did not change 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 these amendments is 
contained in a Safety Evaluation dated April 13, 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: June 27, 2003.
    Brief description of amendments: The amendments modified Technical 
Specification 4.0.5.f and associated Bases, and Bases Section 3/4.4.8, 
with regard to the commitment to perform piping inspections in 
accordance with Generic Letter 88-01, by adding the words ``or in 
accordance with alternate measures approved by the NRC staff.''
    Date of issuance: As of date of issuance and shall be implemented 
within 30 days.
    Effective date: April 20, 2004.
    Amendment Nos.: 171 and 133.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 19, 2003 (68 FR 
49817).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 20, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 30, 2003.
    Brief description of amendment: By letter dated January 30, 2003, 
FirstEnergy Nuclear Operating Company, (FENOC), the licensee for Perry 
Nuclear Power Plant (PNPP), Unit 1, submitted a request for Nuclear 
Regulatory Commission review and approval of a license amendment to 
modify the basis for their compliance with the requirements of Appendix 
H to Title 10 Part 50 of the Code of Federal Regulations (Appendix H to 
10 CFR Part 50), ``Reactor Vessel Material Surveillance Program 
Requirements.'' In the license amendment submittal, FENOC requested 
that they be approved to implement the Boiling Water Reactor Vessel and 
Internals Project reactor pressure vessel integrated surveillance 
program as the basis for demonstrating the compliance of PNPP, Unit 1, 
with the requirements of Appendix H to 10 CFR Part 50.
    Date of issuance: April 15, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 128.
    Facility Operating License No. NPF-58: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
696).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 14, 2003.

[[Page 26196]]

    Brief description of amendment: By letter dated January 14, 2003, 
FirstEnergy Nuclear Operating Company, the licensee for Perry Nuclear 
Power Plant, Unit 1, submitted a request for Nuclear Regulatory 
Commission review and approval of a license amendment to modify the 
Technical Specifications (TS) 5.1.1, 5.4.1, and 5.5.1 to replace the 
requirement for the plant manager to approve administrative procedures 
and the Offsite Dose Calculation Manual. The plant manager approval 
signature will be replaced with the signature of a procedurally 
authorized individual who would be the more appropriate authority for 
approval of the activity. Additionally, a change is proposed to revise 
License Condition 2.F, to replace the 30-day reporting period with a 
direct reference to the 10 CFR 50.73 subsection that contains the 
reporting period. The License Condition already references 10 CFR 50.73 
for use in reporting plant issues.
    Date of issuance: April 23, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 129.
    Facility Operating License No. NPF-58: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15761).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 23, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 25, 2003, as supplemented 
by your letters dated June 16, 2003, January 14, February 23, and April 
7, 2004.
    Brief description of amendments: The amendments revise the 
technical specifications (TSs) to include implementation of relaxed 
axial offset control of the reactor core through changes in TS 3.2.1 
and TS 3.2.3; relocation of selected operating parameters from TS 2.0, 
TS 3.1.8 and TS 3.3.1 to the Core Operating Limit Report (COLR) and the 
revised pressurizer pressure-low allowable value in TS Table 3.3.1-1. 
The TS changes also include, in TS 5.6.5, the topical reports 
documenting the Nuclear Regulatory Commission-approved methodologies 
that are used to support COLR implementation.
    Date of issuance: April 28, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 162 and 153.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 29, 2003 (68 FR 
22750).
    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 April 28, 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, Units 1 and 2, San Luis Obispo County, 
California

    Date of application for amendment: May 29, 2003, as supplemented by 
letters dated November 5, 2003 and December 23, 2003.
    Brief description of amendments: The amendment revises Technical 
Specification 3.8.1, ``AC Sources-Operating,'' to extend the completion 
times for the required actions associated with restoration of an 
inoperable diesel generator (DG). Specifically, the changes extend the 
completion times for restoring an inoperable DG from 7 days to 14 days.
    Date of issuance: April 20, 2004.
    Effective date: April 20, 2004, and shall be implemented within 180 
days of the date of issuance.
    Amendment No.: Unit 1-166; Unit 2-167.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2003 (68 FR 
37581).
    The supplemental letters dated November 5, 2003 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 April 20, 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: February 28, 2003, as 
supplemented by letters dated October 30, 2003, December 2, 2003, and 
January 23, 2004.
    Brief description of amendments: The amendments revise the Diablo 
Canyon Power Plant Technical Specifications (TS) to add a surveillance 
requirement to the Power Range Neutron Flux Rate--High Positive Rate 
Trip function.
    Date of issuance: April 22, 2004.
    Effective date: April 22, 2004, and shall be implemented within 180 
days from the date of issuance.
    Amendment Nos.: Unit 1--167; Unit 2--168.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18283).
    The October 30, 2003, December 2, 2003, and January 23, 2004, 
supplemental letters 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 amendments is contained 
in a Safety Evaluation dated April 22, 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: June 5, 2003.
    Brief description of amendment: Revise the required actions in 
Technical Specification (TS) 3.6.1.9 when a containment purge or 
exhaust isolation valve is found inoperable as a result of leakage in 
excess of the limit. The changes allow alternate methods to ensure flow 
path isolation to the environment consistent with the methods allowed 
for containment isolation valves in TS 3.6.3, ``Containment Isolation 
Valves.''
    Date of issuance: April 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 290 & 280.
    Facility Operating License No. DPR-77: Amendment revises the TSs.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40719).

[[Page 26197]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 21, 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: August 22, 2003, as supplemented 
March 19, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification 3.3.1, ``Reactor Trip System Instrumentation.'' The 
revision adds a Surveillance Requirement for response time to the 
Source Range Neutron Flux Reactor Trip function.
    Date of issuance: April 19, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 52.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54753). The supplemental letter 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 April 19, 2004.

    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 30th day of April 2004.
    For the Nuclear Regulatory Commisison.

Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-10305 Filed 5-10-04; 8:45 am]
BILLING CODE 7590-01-P