[Federal Register Volume 68, Number 181 (Thursday, September 18, 2003)]
[Notices]
[Pages 54747-54757]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-23251]


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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 Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, August 22, 2003, through September 4, 2003. 
The last biweekly notice was published on September 2, 2003 (68 FR 
52233).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation

[[Page 54748]]

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.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing 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 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By October 16, 2003, 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 2.714, 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 by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide 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 to relief. A 
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for

[[Page 54749]]

leave to intervene and requests for hearing be transmitted to the 
Secretary of the Commission either by means of facsimile transmission 
to 301-415-1101 or by e-mail to [email protected]. 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 because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that copies be transmitted either by means of facsimile 
transmission to 301-415-3725 or by e-mail to [email protected]. A 
copy of the request for hearing and petition for leave to intervene 
should also be sent to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    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 
[email protected].

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

    Date of amendment request: August 19, 2003.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) to modify the requirements for the 
containment pressure control system to eliminate a problem with circuit 
fluctuation as a result of electronic noise.
    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 license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed amendment has no impact on any accident 
probabilities or consequences. The CPCS [containment pressure 
control system] functions to control the operation of the 
Containment Spray System and the Air Return System following certain 
design basis accidents. It cannot initiate any accidents by itself. 
Therefore, accident probabilities will be unaffected. Since the 
proposed change has been shown to have no effect upon any safety 
analysis results, the consequences of accidents will also be 
unaffected.
    (2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    As stated previously, the CPCS in and of itself cannot initiate 
any accident condition. No change to any method of plant operation 
is being proposed in conjunction with this amendment request. 
Therefore, no new accident types can be created.
    (3) The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed amendment will have no impact on any safety margin. 
None of the results of any existing safety analyses is affected as a 
result of the proposed change. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions. The fission product barriers include the 
fuel cladding, the reactor coolant pressure boundary, and the 
containment. None of these fission product barriers will be affected 
as a result of the proposed change. Therefore, no safety margin will 
be impacted.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, and Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, located in Mecklenburg County, North 
Carolina and York County, South Carolina

    Date of amendment request: March 24, 2003, as supplemented June 25, 
2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to relocate reactor coolant 
system cycle-specific parameter limits from the TS to the core 
operating limits reports for the Catawba and the McGuire Nuclear 
Stations.
    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:
    As required by 10 CFR 50.91(a)(1), this analysis is provided to 
demonstrate that the proposed license amendment does not involve a 
significant hazard.
    Conformance of the proposed amendment to the standards for a 
determination of no significant hazards, as defined in 10 CFR 50.92, is 
shown in the following:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The relocation of Reactor Coolant System (RCS) related 
cycle-specific parameter limits from the Technical Specifications 
(TS) to the Core Operating Limits Reports (COLR) proposed by this 
amendment request does not result in the alteration of the design, 
material, or construction standards that were applicable prior to 
the change. The proposed change will not result in the modification 
of any system interface that would increase the likelihood of an 
accident since these events are independent of the proposed change. 
The proposed amendment will not change, degrade, or prevent actions, 
or alter any assumptions previously made in evaluating the 
radiological consequences of an accident described in the UFSARS. 
Therefore, the proposed amendment does not result in the 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. This change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
No new accident causal mechanisms are created as a result of NRC 
approval of this amendment request. No changes are being made to the 
facility which should introduce any new accident causal mechanisms. 
This amendment request does not impact any plant systems that are 
accident initiators.
    (3) Does the proposed change involve a significant reduction in 
margin of safety?
    No. Implementation of this amendment would not involve a 
significant reduction in the margin of safety. Previously approved 
methodologies will continue to be used in the determination of 
cycle-specific core operating limits appearing in the COLRS. 
Additionally, previously approved RCS minimum total flow rates for 
McGuire and Catawba are retained in their respective TS

[[Page 54750]]

so as to assure that lower flow rates will not be used without prior 
NRC approval. Consequently, no safety margins will be impacted.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: August 11, 2003.
    Description of amendment request: The proposed amendment would 
relocate Technical Specification (TS) Surveillance Requirement 4.5.2.f 
(vacuum leak rate test of the watertight enclosure for decay heat 
removal system valves DH-11 and DH-12) from the TSs to the Technical 
Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Under the proposed change, initial conditions and assumptions 
remain as previously analyzed for accidents in the Davis-Besse 
Nuclear Power Station Updated Safety Analysis Report. 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.
    Under the proposed change, the manner in which the watertight 
enclosure is sealed and tested is not altered, and the operability 
requirements of the watertight enclosure for Decay Heat Removal 
System valves DH-11 and DH-12 will continue to be adequately 
addressed by testing. No different accident initiators or failure 
mechanisms are introduced by the proposed change. 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.
    Since there are no new or significant changes to the initial 
conditions contributing to accident severity or consequences, there 
are no significant reductions in a margin of safety. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear 
Generating Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: July 21, 2003.
    Description of amendment request: The amendment application 
proposes a revision to the Technical Specifications (TS) administrative 
controls for the radioactive effluent controls program. The proposed 
changes will make the Three Mile Island Nuclear Generating Station Unit 
2 (TMI-2) radioactive effluent controls program technical 
specifications consistent with the technical specifications for the 
operating facility on site--Three Mile Island Nuclear Generating 
Station, Unit 1 (TMI-1). The proposed change adopts the TMI-1 liquid 
discharge limits since both TMI-1 and TMI-2 use the same liquid 
discharge monitor and have a common discharge pathway. The gaseous 
discharge limits will also be updated to reflect the current 10 CFR 
part 20 nomenclature along with some minor editorial changes. 
Additionally, the definition of a member of the public will be made 
consistent with the definition in 10 CFR part 20.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The TMI-2 TS for radioactive liquid effluent release, TS 
6.7.4.a.2, will be revised to be consistent with the equivalent TS 
for TMI-1 (TS 6.8.4.b.(2)). The change will allow up to 10 times the 
concentrations specified in 10 CFR Part 20, Appendix B, Table 2, 
Column 2. Making the limits on the liquid effluent release 
concentrations for TMI-2 equivalent to those for TMI-1 is justified 
in that both units share a common effluent monitoring instrument and 
a common discharge path to the Susquehanna River.
    The TMI-2 TS for limits on dose rate for radioactive gaseous 
effluent, TS 6.7.4.a.7, will be changed from the limits in 10 CFR 
20, Appendix B, Table 2, Column 1, to be consistent with the 
equivalent TS for TMI-1 (TS 6.8.4.b.(7)). The revised limits will be 
as follows: (a) For noble gases: less than or equal to 500 mrem/yr 
to the total body and less than or equal to 3000 mrem/yr to the 
skin, and (b) For tritium and all radionuclides in particulate form 
with half-lives greater than 8 days: less than or equal to 1500 
mrem/yr to any organ. The TMI-2 TS will continue to specify that 
annual and quarterly doses conform to Appendix I of 10 CFR Part 50.
    The other changes are administrative and do no affect plant 
systems.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated?
    No. These changes will affect administrative controls on 
radionuclides that may be released from the site. It does not change 
the allowable off-site dose limits for any calendar year of 
operations. It does not change any plant system or the ALARA 
philosophy on discharges. Therefore, the proposed changes do not 
involve the possibility of a new or different type of accident from 
any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    No. These changes will affect the administrative controls on 
radionuclides that may be released from the site. It does not change 
the allowable off-site dose limits for any calendar year of 
operations. It does not change any plant system and will not affect 
the actual discharges from the plant. Therefore, there cannot be a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Esq., First Energy Legal 
Department, 76 South Mail Street, Akron, OH 44308.
    NRC Section Chief: Scott W. Moore.

[[Page 54751]]

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

    Date of amendment request: August 28, 2003 (superseded the July 18, 
2003, application).
    Description of amendment request: The proposed amendment will 
increase the licensed power level to 1524 megawatts thermal (MWt) or 
1.60 percent greater than the current power level of 1500 MWt. The 
requested increase in licensed rated power is the result of a 
measurement uncertainty recapture (MUR) power uprate. The information 
provided in support of this request is based on the NRC's Regulatory 
Issue Summary 2002-03, ``Guidance on the Content of Measurement 
Uncertainty Recapture Power Uprate Applications,'' dated January 31, 
2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    There are no changes as a result of the MUR power uprate to the 
design or operation of the plant that could affect system, 
component, or accident functions. All systems and components 
function as designed and the performance requirements have been 
evaluated and found to be acceptable. The reduction in power 
measurement uncertainty allows for safety analyses to continue to be 
used without modification. This is because the safety analyses 
dependent on power level were performed or evaluated at 102% of 1500 
MWt (1530 MWt) or higher. Analyses at these power levels support a 
core power level of 1524 MWt with a measurement uncertainty of 0.4%. 
Radiological consequences of USAR [Updated Safety Analysis Report] 
Chapter 14 accidents were assessed previously using the alternate 
source term methodology (Reference 10.2 [of the August 28, 2003, 
application]). These analyses were performed at 102% of 1500 MWt 
(1530 MWt) and continue to be bounding. Updated Safety Analysis 
Report (USAR) Chapter 14 analyses and accident analyses continue to 
demonstrate compliance with the relevant accident analyses' 
acceptance criteria.
    Therefore, there is no significant increase in the consequences 
of any accident previously evaluated.
    The primary loop components (reactor vessel, reactor internals, 
control element drive mechanisms, loop piping and supports, reactor 
coolant pumps, steam generators, and pressurizer) were evaluated at 
an uprated core power level of 1524 MWt and continue to comply with 
their applicable structural limits. These analyses also demonstrate 
the components will continue to perform their intended design 
functions. Changing the heatup and cooldown curves is based on 
uprated fluence values. This does not have a significant effect on 
the reactor vessel integrity. Thus, there is no significant increase 
in the probability of a structural failure of the primary loop 
components. The LBB [leak-before-break] analysis conclusions remain 
valid and the breaks previously exempted from structural 
consideration remain unchanged.
    All of the NSSS [nuclear steam system supplier] systems will 
continue to perform their intended design functions during normal 
and accident conditions. The auxiliary systems and components 
continue to comply with the applicable structural limits and will 
continue to perform their intended functions. The NSSS/BOP [balance-
of-plant] interface systems were evaluated at 1524 MWt and will 
continue to perform their intended design functions. Plant 
electrical equipment was also evaluated and will continue to perform 
their intended functions. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed change. All 
systems, structures, and components previously required for the 
mitigation of an event remain capable of fulfilling their intended 
design function at the uprated power level. The proposed change has 
no adverse effects on any safety related systems or component and 
does not challenge the performance or integrity of any safety 
related system. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Operation at 1524 MWt core power does not involve a significant 
reduction in the margin of safety. The current accident analyses 
have been previously performed with a 2% power measurement 
uncertainty or at uprated core powers that exceed the MUR uprated 
core power. System and component analyses have been completed at the 
MUR uprated core power conditions. Analyses of the primary fission 
product barriers at uprated core powers have concluded that all 
relevant design basis criteria remain satisfied in regard to 
integrity and compliance with the regulatory acceptance criteria. As 
appropriate, all evaluations have been both reviewed and approved by 
the NRC, or are currently under review (the proposed Pressure-
Temperature Limits Report). Therefore, the proposed change does not 
involve a significant reduction in margin of safety.

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

Southern California Edison Company, et al., Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit 1, San Diego County, California

    Date of amendment requests: July 25, 2003.
    Description of amendment requests: The amendment application 
requests a revision to the Unit 1 Defueled Safety Analysis Report 
(DSAR) that concerns the turbine gantry crane, turbine gantry crane 
capacity, fuel shipment and the structural descriptions of the turbine 
building. The licensee is engineering structural changes to the turbine 
building and gantry crane and replacing the turbine gantry crane 
trolley in preparation for moving spent fuel from the Unit 1 spent fuel 
pool to the Independent Spent Fuel Storage Installation (ISFSI). With 
the planned modifications listed above, the licensee will be able to 
satisfy the guidance of NUREG-0612, ``Control of Heavy Loads at Nuclear 
Power Plants,'' and NUREG-0554. ``Single-Failure Proof Cranes for 
Nuclear Power Plants,'' (regarding safe load handling paths and single-
failure proof cranes) in performing the necessary movement of Unit 1 
spent fuel to dry cask storage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The DSAR addresses fuel handling accidents. The process for 
transporting a cask is essentially unchanged from that previously 
performed. The building arrangement is such that the cask is never 
carried over the spent fuel pool. The transport height of the cask 
has been increased to a minimum of 9 inches based on the design of 
the new Ederer X-Sam single-failure proof trolley. Because the 
turbine gantry crane upgrade improves the reliability of the crane, 
a single failure will not result in loss of its capability to safely 
retain control of the hook load.
    If a portion of the new turbine gantry crane lifting device 
malfunctions or fails, the crane system is designed such that the 
load will move a limited distance downward prior to backup 
restraints becoming engaged. The increased minimum transport height 
(9 inches) is established to accommodate the

[[Page 54752]]

design features. The probability of a fuel handling accident is 
unchanged. Because the spent fuel fission product activity has 
decayed by more than ten years compared to the source term analyzed 
in the DSAR, the consequences of the analyzed fuel handling accident 
are significantly lessened.
    Therefore, the proposed DSAR change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated?
    No. By implementing use of a qualified single-failure proof 
crane for cask handling, accidental dropping of the cask is not 
postulated. The cask load will be increased to a maximum of 105 tons 
under the new single failure proof turbine gantry crane design. The 
construction of a single failure proof turbine gantry crane 
mitigates the potential for an accident, since a single failure will 
not result in the loss of its capability to safely retain control of 
the hook load.
    Therefore, performing fuel transfer in a manner consistent with 
the proposed DSAR amendment will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety?
    No. The proposed DSAR change makes use of analysis methods and 
inputs consistent with other structural and safety analyses given in 
the DSAR. The turbine gantry crane will be upgraded to comply with 
the single failure proof requirements of NUREG-0554. The safety 
margins provided by the new crane design have either remained the 
same or have been enhanced to ensure adequate margin to prevent 
failure of the crane or any lifting devices associated with the 
lifting of a spent fuel transfer cask.
    Therefore, the proposed DSAR 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: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Scott W. Moore.

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

    Date of amendment requests: August 4, 2003.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification 3.9.3, ``Containment Penetrations.'' 
Specifically, a Note will be added to the Limiting Condition for 
Operations that permits the Containment equipment hatch to be open 
during core alterations and movement of irradiated fuel in containment.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to Technical Specification 3.9.3 would allow 
the containment equipment hatch to be open during fuel movement or 
core alterations. Currently, the equipment hatch is closed with four 
bolts during fuel movement or core alterations to prevent the escape 
of radioactive material in the event of an in-containment fuel 
handling accident. The containment equipment hatch is not an 
initiator of an accident. Whether the containment equipment hatch is 
open or closed during fuel movement and core alterations has no 
affect on the probability of any accident previously evaluated.
    Allowing the containment equipment hatch to be open during fuel 
movement or core alterations does not significantly increase the 
consequences from a fuel handling accident. The calculated offsite 
doses are well within the limits of 10 CFR Part 100 and the 
calculated control room operator dose are within the limits of 10 
CFR [Part] 50 Appendix A General Design Criterion (GDC) 19. In 
addition, the calculated doses are larger than the expected doses 
because the calculation does not incorporate containment closure 
after the containment is evacuated, which is much less than the two 
hours assumed in the analysis. The proposed change should 
significantly reduce the dose to workers in containment in the event 
of a fuel handling accident by reducing the time required to 
evacuate the containment.
    The changes being proposed do not adversely affect assumptions 
contained in other plant safety analyses or the physical design of 
the plant, nor do they affect other Technical Specifications that 
preserve safety analysis assumptions. Therefore, operation of the 
facility in accordance with the proposed amendment would not involve 
a significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change to Technical Specification 3.9.3, 
``Containment Penetrations,'' affects a previously evaluated fuel 
handling accident inside containment. The new Fuel Handling Accident 
analysis continues to assume that all of the iodine and noble gases 
that become airborne escape the containment within two hours, and 
reach the exclusion area boundary and control room with no credit 
taken for containment air exhaust filtration, or for decay or 
deposition during atmospheric dispersion. The change will include 
the addition of flashing that will restrict a release of post-
accident fission products when the Containment Structure Equipment 
Hatch Shield Doors are in their closed position. In this manner, the 
closed Shield Doors will provide Containment closure. Accordingly, 
since the proposed change does not functionally alter the design of 
plant systems and the revised analysis is consistent with the Fuel 
Handling Accident analysis, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. [The containment equipment hatch is not an 
initiator of an accident.]
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The margin of safety as defined by 10 CFR Part 100 has not been 
significantly reduced. The calculated dose is well within the limits 
given in 10 CFR Part 100 as defined by Standard Review Plan 15.7.4. 
The analysis does not credit closing the Containment Structure 
Equipment Hatch Shield Doors. Accordingly, the proposed change does 
not alter the bases for assurance that safety-related activities are 
performed correctly or the basis for any Technical Specification 
that is related to the establishment of or maintenance of a safety 
margin. Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.
    Based on the above discussion, Southern California Edison has 
determined that the proposed amendment request does not (1) involve 
a significant increase in the probability or consequences of an 
accident previously evaluated, (2) create the possibility of a new 
or different kind of accident from any accident previously 
evaluated, or (3) involve a significant reduction in a margin of 
safety, therefore, the proposed change does not involve a 
significant hazards consideration as defined in 10 CFR 50.92.
    Therefore, the operation of the facility in accordance with this 
proposed change will not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California

[[Page 54753]]

Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.
    NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: July 25, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.1.8, ``Scram Discharge Volume 
(SDV) Vent and Drain Valves,'' to allow a vent or drain line with one 
inoperable valve to be isolated instead of requiring the valve to be 
restored to Operable status within 7 days.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments to revise the action for one or more SDV vent or drain lines 
with an inoperable valve, including a model safety evaluation and model 
no significant hazards consideration (NSHC) determination, using the 
consolidated line-item improvement process (CLIIP). The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on April 15, 2003 (68 FR 18295). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
July 25, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any previously evaluated.

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

    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDVs is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of an SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: August 22, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.1, ``Reactor Trip System 
Instrumentation.'' The revision adds a Surveillance Requirement for 
response time to the Source Range (SR) Neutron Flux Reactor Trip 
function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed amendment enhances the operability of the SR 
reactor trip channels by requiring response time testing. This will 
provide additional assurance that the plant will be operated within 
its design and licensing basis. The change does not involve any 
physical modifications or functional design changes to the SR 
instrumentation, and will not alter any system interfaces. The 
design standards, criteria, and material specifications applicable 
to the design and installation of the SR instrumentation still 
apply. The performance of response time testing for the SR Neutron 
Flux channels does not contribute to the initiation of any accident 
previously evaluated. Testing will be performed when the SR reactor 
trip function is not required to be operable. A response time will 
ensure that a Uncontrolled Rod Cluster Control Assembly Bank 
Withdrawal from Subcritical (RWFS) event in Modes 3, 4, or 5 remains 
bounded by the current analysis and the reactor would be shutdown 
before any significant power is generated. Thus, the probability of 
occurrence of an accident evaluated in the Updated Final Safety 
Analysis Report (UFSAR) will not increase as a result of the 
performance of response time testing. The performance of response 
time testing will not affect any radiological barriers. The testing 
will not alter any operator responses required for accident 
mitigation and will not change any assumptions made in evaluating 
radiological consequences of an accident described in the UFSAR. The 
consequences of an RWFS event occurring from Mode 3, 4, or 5 are 
less severe than from Mode 2 since reactivity levels are lower in 
the lower modes. Therefore, there is no potential for an increase in 
the consequences of any previously evaluated accident.
    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 will not require any changes to 
hardware, setpoints, or design functions. The addition of a response 
time test requirement will not change the way the system is operated 
but will impose more restrictive operability requirements for the SR 
reactor trip function. This enhancement to the operability 
requirements for a protection system function is not considered an 
accident initiator. Therefore, the activity will not create a new or 
different kind of accident from those previously evaluated in the 
UFSAR.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change does not involve any changes to 
setpoints or safety limits. The required response time is consistent 
with the current accident analysis described in UFSAR and will 
ensure that a RWFS event in Modes 3, 4, or 5 remains bounded by the 
current analysis. The addition of a response time verification 
requirement is an enhancement to the operability requirements of the 
SR reactor trip channels and does not reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority,

[[Page 54754]]

400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].

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

    Date of application of amendments: April 10, 2003, as supplemented 
by letter dated July 1, 2003.
    Brief description of amendments: The amendments revised frequencies 
associated with the Technical Specification Surveillance Requirements 
3.4.12.5 and 3.4.12.7 concerning the Low Temperature Overpressure 
Protection System.
    Date of Issuance: August 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 333, 333, and 334.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
2885).
    The supplement dated July 1, 2003, provided clarifying information 
that did not change the scope of the April 10, 2003, application nor 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 25, 2003.
    No significant hazards consideration comments received: No.
    Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas
    Date of application for amendment: June 30, 2003, as supplemented 
by letters dated August 1 and 12, 2003.
    Brief description of amendment: The amendment (1) eliminated credit 
for the Boraflex neutron absorbing material used for reactivity control 
in Region 1 of the spent fuel pool (SFP), (2) credited a combination of 
soluble boron and several defined fuel loading patterns within the 
storage racks to maintain SFP reactivity within the effective neutron 
multiplication factor (Keff) limits of 10 CFR 50.68, (3) 
increased the minimum boron concentration in the SFP to greater than 
2000 parts per million (ppm), and (4) reduced the fresh fuel assembly 
initial enrichment to less than or equal to 4.55 +/- 0.05 weight 
percent uranium-235 (U-235).
    Date of issuance: September 3, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 250.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43384).
    The August 1 and 12, 2003, supplemental letters provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 3, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 23, 2002, as 
supplemented July 2, 2003.
    Brief description of amendments: The amendments revise (1) the 
Operating Licenses to delete obsolete and expired license conditions 
and make administrative and editorial changes, and (2) the Technical 
Specifications (TSs) to make administrative and editorial changes.
    Additionally, the licensee proposed to delete the radiation 
monitoring instrumentation identification numbers from certain TSs. The 
licensee will be submitting new information to support these changes in 
a future request. The NRC staff will handle this request under separate 
cover.
    Date of issuance: August 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 279 and 261.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63695).
    The supplement dated July 2, 2003, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the Nuclear 
Regulatory Commission staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 22, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 2, 2003, as supplemented by 
letters

[[Page 54755]]

dated June 30, July 30, August 8, and 18, 2003.
    Brief description of amendment: The amendment updates the existing 
reactor coolant system pressure and temperature limit curves (TS Figure 
3.4.9-1) and extends their applicability to 32 effective full power 
years.
    Date of issuance: August 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
by September 1, 2003.
    Amendment No.: 253.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28855).
    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 amendment is contained 
in a Safety Evaluation dated August 25, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 29, 2003.
    Brief description of amendment: The amendment revises the drywell 
leakage and sump monitoring detection section of the current Technical 
Specifications (TSs). Specifically, the changes clarify the associated 
definitions and divide TS 3.6.D/4.6.D, ``Coolant Leakage,'' into two 
subsections and retitle it ``Reactor Coolant System (RCS).'' One of the 
subsections contains the Limiting Condition for Operations (LCOs) for 
RCS operational leakage, and the other subsection contains the LCOs for 
the RCS leakage detection instrumentation.
    Date of issuance: August 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 137.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18279).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 21, 2003.
    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: August 27, 2002, and its 
supplements dated May 15, June 26, and August 1, 2003.
    Brief description of amendments: The amendments revised Table 
3.3.1-1, ``Reactor Trip System Instrumentation'' of the technical 
specifications to replace the term ``minimum measured flow per loop'' 
to ``measured loop flow'' in the allowable value and nominal trip 
setpoint for the reactor coolant flow-low reactor trip function, and 
delete footnote (l). The amendments also allow an alternate method for 
the measurement of reactor coolant system (RCS) total volumetric flow 
rate through measurement of the elbow tap differential pressure on the 
RCS primary cold legs.
    Date of issuance: August 21, 2003.
    Effective date: August 21, 2003, and shall be implemented within 30 
days from the date of issuance.
    Amendment Nos.: Unit 1-161; Unit 2-162.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 7, 2003 (68 FR 
810).
    The May 15, June 26, and August 1, 2003, 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 August 21, 2003.
    No significant hazards consideration comments received: No.

Rochester Gas and Electric Corporation, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: May 3, 2001, as supplemented 
August 7, 2001, October 29, 2001, May 3, 2002, October 7, 2002, 
November 5, 2002, and June 6, 2003.
    Brief description of amendment: The amendment revised the Ginna 
Station Improved Technical Specifications to reflect design changes to 
the actuation circuitry associated with the Control Room Emergency Air 
Treatment System (CREATS). The proposed design changes consist of 
replacing the current diverse radiation monitors with two Geiger-
Mueller (GM) tubes powered from two separate safety-related power 
supplies which are configured into two redundant actuation logic trains 
using safety-grade digital instrumentation. The design changes are 
intended to increase system reliability by providing redundancy and 
reducing spurious actuations. The amendment changes limiting condition 
for operation 3.3.6 for the CREATS Actuation Instrumentation as 
follows:
    a. Adds a new Condition to require immediately placing the CREATS 
in the emergency mode of operation upon the loss of two instrument 
channels/trains.
    b. Adds a new surveillance requirement involving a CHANNEL CHECK of 
the Control Room Radiation Intake Monitors.
    c. Revises Table 3.3.6-1 to increase the number of trains of Manual 
and Automatic Initiation Circuits from one train to two trains.
    d. Extends the Completion Time of the Required Action for a loss of 
one channel/train from 1 hour to 7 days as the result of installing 
redundant channels/trains.
    e. Revises Table 3.3.6-1 to remove reference to the iodine, noble 
gas, and particulate control room radiation intake monitors. These 
monitors will be replaced by the two new GM tubes.
    f. Revises Table 3.3.6-1 to replace the column heading ``Trip 
Setpoint'' with ``Allowable Value.''
    Date of issuance: August 29, 2003.
    Effective date: August 29, 2003.
    Amendment No.: 83.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46481).
    The supplemental letters referenced above provided clarifying 
information that did not change the scope of the amendment as described 
in the original 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 August 29, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: September 24, 2002, as supplemented by 
letters dated May 20 and July 16, 2003.
    Brief Description of amendments: The changes revise Technical 
Specifications

[[Page 54756]]

(TS) 3.7.10, ``Control Room Emergency Filtration/Pressurization System 
(CREFS),'' and TS 3.7.12 ``Penetration Room Filtration (PRF) System,'' 
to establish actions to be taken for inoperable ventilation systems due 
to a degraded control room pressure boundary or PRF and spent fuel pool 
room boundary, respectively. This revision approves changes that would 
allow up to 24 hours to restore the pressure boundary to an operable 
status when two ventilation trains are inoperable due to an inoperable 
pressure boundary in MODES 1, 2, 3, and 4. In addition, a Limiting 
Condition for Operation Note would be added to allow the pressure 
boundary to be opened intermittently under administrative control 
without affecting CREFS or PRF System operability. The applicable TS 
Bases have been revised to document the TS changes and to provide 
supporting information. These changes are based on Technical 
Specifications Task Force document TSTF-287, Revision 5.
    Date of issuance: August 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 161 and 154.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68744).
    The supplements dated May 20 and July 16, 2003, provided clarifying 
information that did not change the scope of the September 24, 2002, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 22, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 14, 2003.
    Brief description of amendments: The amendments revise 
Surveillance.
    Requirement 4.6.2.1 for demonstrating operability of containment 
spray system spray nozzles to require verification of operability only 
after spray ring header maintenance that could result in nozzle 
obstructions without specifying the method of verification.
    Date of issuance: August 20, 2003.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-156; Unit 2-144.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2003 (68 FR 
37582).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 20, 2003.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: April 15, 2003.
    Description of amendment request: The amendments revised Technical 
Specification (TS) 3.7.3, ``Control Room Emergency Ventilation (CREV) 
System,'' to allow up to 24 hours to restore the control room pressure 
boundary (CRPB) to operable status when two trains of the ventilation 
system are inoperable due to an inoperable CRPB in MODES 1, 2, and 3. 
In addition, a note is included to allow the pressure boundary to be 
opened intermittently under administrative controls without affecting 
the CREV System operability. The licensee revised the applicable TS 
Bases to make them consistent with the TS changes. These changes are 
based on TS Task Force Traveler No. 287, which was approved by the NRC 
on March 16, 2000.
    Date of issuance: August 29, 2003.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 246, 283 and 241.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
Amendments revised the TSs.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28858).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 29, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 5, 2003.
    Brief description of amendments: The amendments extend from 1 hour 
to 24 hours the completion time for Condition B of Technical 
Specification 3.5.1, which defines requirements for the restoration of 
an emergency core cooling system accumulator when it has been declared 
inoperable for a reason other than boron concentration.
    Date of issuance: August 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 106 and 106.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40721).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 25, 2003.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: June 9, 2003, as supplemented 
on July 28, 2003.
    Brief Description of amendments: These amendments revise Section 6 
of the Surry Power Station Technical Specifications (TS) for Units 1 
and 2 to adopt the format for topical report references that are 
described in Industry/Technical Specifications Task Force Traveler, 
TSTF-363, Rev 0, ``Revised Topical Report References in Improved 
Technical Specification (ITS) 5.6.5, [Core Operating Limits Report] 
COLR.''
    Date of issuance: August 27, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 235 and 234.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40722).
    The July 28, 2003, supplement contained clarifying information only 
and did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 27, 2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 5th day of September 2003.


[[Page 54757]]


    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-23251 Filed 9-17-03; 8:45 am]
BILLING CODE 7590-01-P