[Federal Register Volume 69, Number 101 (Tuesday, May 25, 2004)]
[Notices]
[Pages 29761-29772]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-11507]


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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

[[Page 29762]]

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 30, through May 13, 2004. The last 
biweekly notice was published on May 11, 2004 (69 FR 26184).

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

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

[[Page 29763]]

participate fully in the conduct of the hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, Hearingdocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by e-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.

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

    Date of amendment request: March 23, 2004.
    Description of amendment request: The amendments would revise 
Technical Specification 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 March 23, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines [contained] in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    The proposed change does not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, or configuration of the facility, or the manner in which 
the plant is operated and maintained; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for

[[Page 29764]]

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 NSHC.
    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: Stephanie M. Coffin, Acting.

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

    Date of amendment request: April 19, 2004.
    Description of amendment request: The proposed change revises 
Limiting Condition for Operation (LCO) 3.7.3, ``Control Room Emergency 
Filtration System,'' to provide specific conditions and required 
actions that address degraded control room boundary.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed Technical Specifications (TS) change involves the 
Control Room Emergency Filtration (CREF) System and associated 
control room boundary, which provide a radiological controlled 
environment from which the plant can be operated following a design 
basis accident (DBA). The CREF system and the control room boundary 
are not assumed to be initiators of any analyzed accident and do not 
affect the probability of accidents. The proposed change adds a Note 
to LCO 3.7.3 that allows the control room boundary to be opened 
intermittently under administrative controls. A new Condition B is 
also added to LCO 3.7.3 to specify a Completion Time of 24 hours to 
restore an inoperable control room boundary to OPERABLE status 
before requiring the plant to perform an orderly shutdown. The 24-
hour Completion Time is reasonable based on the low probability of a 
DBA occurring during this time period and Energy Northwest's 
commitment to implement, via administrative controls, appropriate 
compensatory measures consistent with the intent of 10 CFR 50, 
Appendix A, General Design Criteria (GDC) 19. These compensatory 
measures will serve to minimize the consequences of an open control 
room boundary and ensure the CREF system can continue to perform its 
function. As such, these changes will not affect the function or 
operation of any other systems, structures or components. Therefore, 
the proposed TS change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change adds a Note to LCO 3.7.3 that allows the 
control room boundary to be opened intermittently under 
administrative controls. A new Condition B is also added to LCO 
3.7.3 to specify a Completion Time of 24 hours to restore an 
inoperable control room boundary to OPERABLE status before requiring 
the plant to perform an orderly shutdown. The CREF system and the 
control room boundary are designed to protect the habitability of 
the control room. The CREF system and the control room boundary are 
not accident initiators and do not affect the probability of 
accidents. This change is administrative in nature and does not 
involve any physical changes to the plant. Therefore, the proposed 
TS change does not create the possibility of a new or different kind 
of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change adds a Note to LCO 3.7.3 that allows the 
control room boundary to be opened intermittently under 
administrative controls. A new Condition B is also added to LCO 
3.7.3 to specify a Completion Time of 24 hours to restore an 
inoperable control room boundary to OPERABLE status before requiring 
the plant to perform an orderly shutdown. The 24-hour Completion 
Time is reasonable based on the low probability of a DBA occurring 
during this time period and Energy Northwest's commitment to 
implement, via administrative controls, appropriate compensatory 
measures consistent with the intent of 10 CFR 50, Appendix A, GDC 
19. These compensatory measures will serve to minimize the 
consequences of an open control room boundary and assure that the 
CREF system can continue to perform its function. Therefore, the 
proposed TS 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 request involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 21, 2003, as supplemented 
February 10, 2004.
    Description of amendment request: The amendment would modify the 
Technical Specifications (TSs) to delete TS 3.6.4.4, ``Shield Building 
Annulus Mixing System,'' in its entirety, revise the Main Steam 
Isolation Valve (MSIV) leakage limits contained within TS Surveillance 
Requirement 3.6.1.3.10, and delete reference to TS 3.6.4.4 within TS 
3.10.1, ``Inservice Leak and Hydrostatic Testing Operation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    As discussed above, the proposed changes are to delete the 
annulus mixing function and deletion of the single MSIV leakage rate 
limit. A review of the safety analysis report indicates that 
operation (or mis-operation) of the annulus mixing system, or any 
component of the annulus mixing system is not considered an 
initiator of any accident evaluated in the Updated Safety Analysis 
Report. The deletion of the single MSIV leakage limit of 50 scfh in 
effect establishes a maximum leakage limit of 150 scfh which is the 
current total MSIV leakage limit. The elimination of the single MSIV 
acceptable leakage rate limit does not impact any event initiator. 
As the proposed changes do not involve any accident initiators, 
there is no increase in the probability of an accident previously 
evaluated.
    The annulus mixing system and the main steam isolation valves 
operate following an LOCA [loss-of-coolant accident] to mitigate the 
consequences of an accident. Elimination of the annulus mixing 
system and the single MSIV leakage limit will lead to some increase 
in the dose consequences of a LOCA. The current LOCA dose 
consequences evaluation for RBS was revised to account for the 
elimination of the annulus mixing system and for increasing the 
single MSIV leakage to 150 scfh (applying the total MS-PLCS Division 
limit to the single MSIV). The results of the revised evaluation 
with the proposed changes show an increase in the calculated dose 
consequences, however, the calculated doses were still within the 
acceptance limits of 10 CFR 50.67. Thus, while there is an increase 
in the dose consequences of an accident previously identified, the 
increase is not deemed to be significant.
    Therefore, the proposed change does not involve a significant 
increase in the

[[Page 29765]]

probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not add any equipment, nor is any 
equipment replaced with equipment with different performance 
characteristics. Thus, no new initiators are added, and therefore, 
no new accident types are created as a result of this change. The 
proposed changes affect performance characteristics assumed in the 
LOCA dose consequences evaluation, however, the nature of the 
accidents evaluated in the safety analysis report are not changed.
    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.
    With respect to dose consequences for the LOCA event, the margin 
of safety is considered to be that provided by meeting the 10 CFR 
50.67 limits. The revised dose consequences evaluation, which 
includes the proposed changes, continues to demonstrate that the 
doses at the exclusion area boundary, the low population zone, and 
the control room are within the acceptance limits in 10 CFR 50.67. 
Therefore, there is no reduction in the 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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 16, 2004.
    Description of amendment request: The amendment would change 
Technical Specification (TS) 3.6.5.1.3, regarding drywell bypass 
leakage testing (DWBT). The change would allow for a one-time extension 
of the interval (from 10 to 15 years) for performance of the next DWBT.
    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 to TS SR 3.6.5.1.3 adds a one-time 
extension to the current interval for the DWBT. The current interval 
of ten years, based on past performance, would be extended on a one-
time basis to 15-years from the date of the last test. The proposed 
extension to the DWBT cannot increase the probability of an accident 
since there are no design or operating changes involved and the test 
is not an accident initiator. The proposed extension of the test 
interval does not involve a significant increase in the consequences 
since analysis has shown that, the proposed extension of the DWBT 
frequency has a minimal impact on plant risk. 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 extension to the interval for the DWBT does not 
involve any design or operational changes that could lead to a new 
or different kind of accident from any accidents previously 
evaluated. The tests are not being modified, but are only being 
performed after a longer interval. 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. 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.
    An evaluation of extending the DWBT surveillance frequency from 
once in 10 years to once in 15 years has been performed using 
methodologies based on the ILRT [integrated leak rate testing] 
methodologies. This evaluation assumed that the DWBT frequency was 
being adjusted in conjunction with the ILRT frequency. This analysis 
used realistic, but still conservative, assumptions with regard to 
developing the frequency of leakage classes associated with the 
DWBT. The results from this conservative analysis indicates that the 
proposed extension of the DWBT frequency has a minimal impact on 
plant risk and 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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 7, 2004.
    Description of amendment request: The proposed changes will revise 
the Waterford 3 Technical Specifications (TS) to clarify the actions of 
TS 3.4.5.1, Reactor Coolant System (RCS) Leakage; some of the 
surveillance requirements (SRs) of TS 3.4.5.2, RCS Operational Leakage; 
and delete duplication in TS 3.3.3.1, Radiation Monitoring 
Instrumentation. The proposed change is based on NUREG-1432, ``Standard 
Technical Specifications Combustion Engineering Plants,'' Revision 2, 
dated April 30, 2001. Also, the proposed change will delete the 
containment atmosphere gaseous radioactivity monitoring system from the 
TS because this monitor does not meet the requirements of Regulatory 
Guide 1.45, Revision 0, ``Reactor Coolant Pressure Boundary Leakage 
Detection Systems,'' and Title 10 of the Code of Federal Regulations 
(10 CFR), Part 50, Appendix A, General Design Criteria 30, ``Quality of 
Reactor Coolant System Pressure Boundary.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed revisions do not involve any physical change to 
plant design. The less restrictive changes proposed in this 
amendment request include relocation of information to the UFSAR 
[updated final safety analysis report], addition of a TS 3.0.4 
exception, utilization of the diversity and redundancy of the 
Waterford 3 leakage detection instrumentation, allowing diversity in 
the contingency actions, deletion of SRs, and addition of an allowed 
outage time when two of three required leakage detection 
instrumentation is inoperable. The less restrictive changes will not 
affect the capability of Waterford 3 to detect RCS leakage. At least 
one RCS leakage detection instrumentation is always required to 
remain operable, and other leakage detection indication, while not 
credited specifically for RCS leakage detection, is still available 
and required to be operable per other TS

[[Page 29766]]

requirements (i.e., Containment Temperature and Containment 
Pressure). Also contingency actions are required (i.e., RCS 
Inventory Balance, containment grab samples, flow switch 
verification) when any of the RCS leakage detection instrumentation 
is inoperable. Performance of the RCS inventory balance is the most 
accurate method of determining and quantifying leakage. The RCS 
inventory balance is being added as a contingency and replacement 
for monitoring instrumentation that has continuous indication and 
alarms in the control room.
    The more restrictive changes proposed by this revision do not 
adversely affect the capability of Waterford 3 RCS leakage detection 
instrumentation to detect RCS leakage. The deletion of the 
containment atmosphere gaseous radioactivity monitor is considered a 
more restrictive change. This monitor does not meet the leakage 
detection requirements of Regulatory Guide 1.45 and does not meet 
the requirements for retention specified in 10 CFR 50.36. Deletion 
of this monitor will reduce the diversity of the Waterford 3 
instrumentation for monitoring the containment atmosphere and 
require the plant to enter an Action statement when the containment 
atmosphere particulate monitor is inoperable. Requiring performance 
of an RCS inventory balance when the containment sump monitor is 
inoperable provides contingency actions when the plant is in a 
degraded RCS leakage detection condition.
    The administrative changes proposed by this revision do not 
adversely affect the capability of Waterford 3 RCS leakage detection 
instrumentation to detect RCS leakage. Relocating the requirements 
associated with the RCS Leak Detection System from various TS to 
Specification 3.4.5.1 and adding requirement to shutdown when all 
required RCS leakage detection instrumentation are inoperable are 
administrative in nature. The relocation of information from one TS 
to another consolidates information and causes less contusion in the 
control room by having all requirements for the leakage detection 
instrumentation in one TS. The addition of a specific action to 
shutdown when all three leakage detection instrumentation are 
inoperable versus an implied requirement to enter TS 3.0.3 is being 
performed to be similar to the STS [Standard Technical 
Specifications].
    None of the above less restrictive, more restrictive, or 
administrative changes affects the accident analyses. Since the 
proposed changes only affect the requirements for the detection of 
RCS leakage, the probability that an accident previously evaluated 
will occur remains unchanged. The proposed changes do not prevent 
nor limit the diversity of acceptable detection of RCS leakage. 
These changes also do not affect the mitigation capability of any 
accident previously evaluated. The consequences of an accident 
previously evaluated are not affected since the mitigation of 
previously evaluated accidents is not affected and leak rate 
information will remain available to station personnel.
    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 aforementioned revisions do not involve any physical change 
to plant design. None of the proposed changes affect[s] the accident 
analyses. The RCS water inventory balance is more accurate than 
normal leak detection methods in regard to actual RCS leak rates, 
and therefore is an excellent alternative when other leak detection 
components may become inoperable. The proposed changes do not 
prevent acceptable detection of RCS leakage by diverse methods. The 
detection of a RCS leak can not cause an accident. Likewise, 
detecting a RCS leak, while in its beginning stages, does not create 
the possibility of a new or different kind of accident than any 
previously analyzed. Therefore, a new or different kind of accident 
than that previously analyzed does not result due to the proposed 
changes of this submittal.
    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 aforementioned revisions do not involve any physical change 
to plant design. The proposed changes do not adversely affect the 
ability of the RCS leakage detection system to detect RCS leakage. 
The ability of the RCS leakage detection instrumentation to detect 
leakage within the requirements of Regulatory Guide 1.45 is actually 
improved. The containment atmosphere gaseous monitor is being 
deleted from TS, because, it does not meet the requirements of 
Regulatory Guide 1.45 to detect a 1.0 gpm [gallon per minute] RCS 
leakage within 1 hour. Extending the AOT [allowed outage time] when 
two of three leakage detection systems is inoperable does not 
decrease the margin of safety because one instrument remains 
operable, other instrumentation capable of indicating RCS leakage is 
available, and an RCS inventory balance is required to be performed 
on an increased frequency. The RCS inventory balance is more 
accurate than normal leak detection methods in regard to actual RCS 
leak rates, and therefore is an excellent alternative when other 
Ieak detection components may become inoperable. Maintaining diverse 
and accurate RCS leak detection methods available and capable of 
prompt leakage detection helps to ensure RCS leaks will be detected 
within an acceptable period of time and, therefore, the proposed 
changes do not significantly reduce the margin to 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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: April 29, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3/4.4.10, ``Reactor Coolant 
System--Structural Integrity, ASME Code Class 1, 2, and 3 Components,'' 
to relocate Surveillance Requirement (SR) 4.4.10.1.b which requires 
that the reactor vessel internals vent valves be tested and inspected, 
to the Technical Requirements Manual (TRM). The Davis-Besse Nuclear 
Power Station (DBNPS) TRM is a licensee-controlled document that is 
incorporated by reference into the DBNPS Updated Safety Analysis Report 
(USAR). Changes to the DBNPS TRM are performed in accordance with the 
regulatory requirements of 10 CFR 50.59.
    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. The proposed surveillance requirement relocation 
from the Technical Specifications to the USAR TRM does not alter the 
design, operation, or testing of any structure, system, or 
component. No preciously analyzed accident scenario is changed. 
Initiating conditions and assumptions remain as previously analyzed. 
Therefore, the proposed changes 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 surveillance requirement relocation 
from the Technical Specifications to the USAR TRM does not alter the 
design, operation, or testing of any structure, system or component. 
The proposed change does not introduce any new or different accident 
initiators. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.

[[Page 29767]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed surveillance requirements relocation 
from the Technical Specifications to the USAR TRM does not affect 
the capabilities of the Reactor Vessel Internals Vent Valves. 
Therefore, the proposed change will not affect 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.

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

    Date of amendment request: May 3, 2004.
    Description of amendment request: The proposed amendment would 
change the facility as described in the Updated Safety Analysis Report 
(USAR) for the emergency diesel generators (EDGs). Specifically, the 
proposed change would describe a departure from Safety Guide 9, 
``Selection of Diesel Generator Set Capacity for Standby Power 
Supplies,'' for the frequency and voltage transient during the EDG 
automatic loading sequence.
    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. The proposed amendment alters the design 
requirements for the Emergency Diesel Generators (EDGs). 
Specifically, the proposed amendment affects the requirements for 
EDG voltage and frequency response following a loss of offsite 
power. The EDGs function to mitigate the consequences of accidents 
when offsite power is not available. The EDGs are not an initiator 
of any analyzed accident.
    The effect of this change on the capability of the EDGs, the 
onsite electric power system, and essentially powered equipment to 
perform their required safety functions has been evaluated, and the 
proposed change does not significantly impact the capability of 
these systems to perform their required accident mitigation 
functions. No previous analyzed accident scenario is affected by the 
proposed change.
    The proposed change does not affect the initiation of any 
analyzed accident. The accident mitigation functions for affected 
equipment are maintained. 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 amendment affects the USAR 
requirements for EDG voltage and frequency response following a loss 
of offsite power. The effect of this change on the capability of the 
EDGs, the onsite electric power system, and essentially powered 
equipment to perform their required safety functions has been 
evaluated, and the proposed change does not significantly impact the 
capability of these systems to perform their required safety 
functions. The assumptions of the current accident analyses are 
maintained and no new or different accident initiators are created. 
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 amendment affects the USAR 
requirements for EDG voltage and frequency response following a loss 
of offsite power. The effect of this change on the capability of the 
EDGs, the onsite electric power system, and essentially powered 
equipment to perform their required safety functions has been 
evaluated, and it is concluded the proposed change does not impact 
the capability of these systems to perform their required safety 
functions. However, since the proposed change does make changes to 
the controlling values for EDG voltage and frequency transient 
response that are less restrictive than those presently described in 
the USAR, this is considered a reduction in a margin of safety.
    The magnitude of voltage and frequency drops which would result 
in failure of the EDGs, the onsite power system, or essentially 
powered equipment have not been determined due to the limitations of 
the transient assessment model and the nonlinear phenomena 
associated with that postulated failure. However, based on (1) a 
computer model and testing of the diesel engine, engine speed 
control governor and actuator, the synchronous generator and 
excitation system that demonstrate the EDGs are capable of starting, 
accelerating, and carrying the required loads, (2) a comprehensive 
evaluation of the impact of the transient voltage and frequency 
response on plant equipment and safety functions, (3) the momentary 
duration of the voltage and frequency dips, and (4) based on 
engineering judgement, the proposed change is not considered to have 
a significant effect on the 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.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: April 23, 2004.
    Description of amendment request: The proposed amendments would 
revise several Technical Specification (TS) Allowed Outage Times for TS 
3.3.3, Accident Monitoring, to be consistent with the Completion Times 
in the related Specification in NUREG-1431, Revision 2, ``Standard 
Technical Specifications Westinghouse Plants (the Improved Standard 
Technical Specifications, or ISTS).''
    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. 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 changes revise the Actions and allowed outage times 
of the accident monitoring instrumentation. The accident monitoring 
instrumentation is not an initiator of any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased by these proposed changes. 
The Technical Specifications continue to require the accident 
monitoring instrumentation to be operable. Therefore, the accident 
monitoring instrumentation will continue to provide sufficient 
information on selected plant parameters to monitor and assess these 
variables following an accident. The consequences of an accident 
during the extended allowed outage time are the same as the 
consequences during the current allowed outage time. As a result, 
the consequences of any accident previously evaluated are not 
significantly increased by these proposed changes. Therefore, the 
proposed amendments do not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different

[[Page 29768]]

kind of accident from any previously evaluated.
    The proposed changes do not alter the design, physical 
configuration, or mode of operation of the plant. The accident 
monitoring instrumentation is not an initiator of any accident 
previously evaluated. No changes are being made to the plant that 
would introduce any new accident causal mechanisms. The proposed 
changes do not affect any other plant equipment. Therefore, 
operation of the facility in accordance with the proposed amendments 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes do not change the operation, function, or 
modes of the plant or equipment operation. The proposed changes do 
not change the level of assurance that the accident monitoring 
instrumentation will be available to perform its function. The 
proposed changes provide a more appropriate time to restore the 
inoperable channel(s) to operable status, and only apply when one or 
more channels of a required instrument are inoperable. The 
additional time to restore an inoperable channel to operable status 
is appropriate based on the low probability of an event requiring an 
accident monitoring instrument during the interval, providing a 
reasonable time for repair, and other means which may be available 
to obtain the required information. Therefore, operation of the 
facility in accordance with the proposed amendments would 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: William F. Burton, Acting.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: March 15, 2004.
    Description of amendment request: Maine Yankee Atomic Power Company 
(Maine Yankee) is requesting that the U.S. Nuclear Regulatory 
Commission (NRC) release the remaining land under License No. DPR-36, 
with the exception of land where the Independent Spent Fuel Storage 
Installation is located. Maine Yankee submitted detailed information on 
dismantlement activities and final status survey results for the Spray 
Building and Spray Pipe with the amendment request, and proposes to 
submit dismantlement and survey information for the remaining land area 
in four additional submittals. Maine Yankee is seeking review and 
approval of the amendment; however, Maine Yankee is requesting that the 
NRC condition the effective date of the license amendment to correspond 
with the NRC's approval of the final information submittal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The requested license amendment involves release of land 
presently considered part of the Maine Yankee plant site under 
license DPR-36. The release of this land will occur after all 
demolition activities are completed and final status surveys have 
been performed to document the final radiological conditions of the 
land. When the release occurs, the only remaining radiological 
hazard at the site will be contained in the Independent Spent Fuel 
Storage Installation (ISFSI). Therefore, the focus of the analysis 
is on the potential impact on the probability and consequences of 
accidents associated with the ISFSI.
    The accident conditions evaluated for the spent fuel storage 
casks include the following: accident pressurization, mis-loading of 
fuel canisters, drop of the vertical concrete casks, explosion, 
fires, maximum anticipated heat load, earthquakes, floods, 
lightening strikes, tornado and tornado driven missiles, tip over of 
vertical concrete cask, and full blockage of vertical concrete cask 
air inlets and outlets. The release of the non-ISFSI land from the 
license will not affect the probability of any of these accidents. 
Maine Yankee will retain sufficient control over activities 
performed on the Owner Controlled Area through rights granted in the 
legal land conveyance documents to ensure that there is no impact on 
consequences from postulated accidents. Therefore, the proposed 
release of the land will not affect the consequences of any of these 
postulated accidents.
    The proposed action, therefore, does not increase either the 
probability or the consequences of any accidents that have been 
considered.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The requested amendment involves release of land presently 
considered part of the Maine Yankee plant site under license DPR-36. 
When the amendment becomes effective, demolition activities will be 
complete and all systems, structures and components will have been 
removed from the land. The requested release of the land does not 
create the possibility of a new or different kind of accident that 
could affect the ISFSI that has not been considered in the design, 
installation or operation of the ISFSI. As noted above, Maine Yankee 
will retain control over activities performed in the Owner 
Controlled Area for the ISFSI to assure that no new hazards are 
introduced that could create the potential for a new or different 
kind of accident. Therefore, the proposed amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety defined in the statements of consideration 
for the final rule on the Radiological Criteria for License 
Termination is described as the margin between the 100 mrem/yr 
public dose limit established in 10 CFR 20.1301 for licensed 
operation and the 25 mrem/yr dose limit to the average member of the 
critical group at a site considered acceptable for unrestricted use. 
This margin of safety accounts for the potential effect of multiple 
sources of radiation exposure to the critical group. Additionally, 
the State of Maine, through legislation, has imposed a 10 mrem/yr 
all pathways dose limit, with no more than 4 mrem/yr attributable to 
drinking water sources.
    The License Termination Plan (LTP) prepared by Maine Yankee 
establishes conservative criteria for residual radiation levels 
following completion of demolition activities at the site. The LTP 
demonstrates that when these conservative criteria are met, the dose 
to the average member of the critical group will be below the 
regulatory criteria established by the State of Maine, and, 
therefore, well below the dose limits established by the NRC. The 
proposed release of the site lands, once the criteria established in 
the LTP have been met will, therefore, not result in any reduction 
in the margin of safety.

Conclusion

    Based on the above, Maine Yankee concludes that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power 
Company, 321 Old Ferry Road, Wiscasset, Maine 04578
    NRC Section Chief: Claudia M. Craig.

[[Page 29769]]

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

    Date of amendment request: April 19, 2004.
    Description of amendment request: The licensee proposed to revise 
the Technical Specifications (TSs) to establish an operating cycle (24-
month) calibration surveillance frequency for the Intermediate Range 
Monitor (IRM) instrumentation, which would replace the current ``prior 
to startup and normal shutdown'' Surveillance Requirement (SR). The 
proposed changes also included associated conforming changes. In 
addition, the licensee proposed to relocate the Limiting Conditions for 
Operation (LCOs) and SRs for selected control rod withdrawal block 
instrumentation to the Updated Final Safety Analysis Report (UFSAR), a 
licensee-controlled document.
    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 changes are limited to: (1) establishing a 24-month 
calibration frequency for the IRM instrumentation in lieu of the 
current ``prior to startup and normal shutdown'' requirement and 
incorporating the associated conforming changes, and (2) the 
relocation of certain instrumentation requirements from the TSs that 
do not satisfy the screening criteria for retention in the TSs. The 
proposed changes do not introduce any new modes of plant operation, 
make any physical changes to the plant, or alter any operational 
setpoints in a manner which could degrade the performance of, or 
increase the challenges to, any safety system assumed to function in 
the accident analysis. In addition, evaluations of the proposed 
changes pursuant to NRC and industry guidance demonstrate that the 
availability and reliability of equipment and systems required to 
prevent or mitigate the radiological consequences of an accident are 
not significantly affected. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes establish a 24-month IRM calibration 
frequency in lieu of the current ``prior to startup and normal 
shutdown'' requirement and relocate certain instrumentation 
requirements to the UFSAR. As such, the proposed changes do not 
eliminate any requirements or impose any new requirements, and 
adequate controls of existing requirements are maintained. 
Furthermore, since the proposed changes do not make any physical 
changes to the plant, no new accident initiators or failure 
mechanisms are introduced, and the accident assumptions and initial 
conditions will remain unchanged. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident [previously] evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes establish a 24-month IRM calibration 
frequency in lieu of the current ``prior to startup and normal 
shutdown'' requirement and relocate certain instrumentation 
requirements to the UFSAR. Although the proposed changes result in 
changes to surveillance intervals, the impact, if any, on system 
availability is small based on (1) other more frequent testing that 
is performed, (2) the existence of redundant equipment, and (3) 
overall system reliability. Consistent with the findings of previous 
industry evaluations, the NMP1 [Nine Mile Point Nuclear Station, 
Unit No. 1] plant-specific analyses have shown no evidence of time-
dependent failures that would impact the availability of the 
affected systems. Furthermore, plant-specific evaluations and the 
adoption of the calculated IRM setpoint Allowable Values ensure that 
the setpoint margins are maintained for a 24-month (30-month 
maximum) calibration frequency. The proposed relocated requirements 
are consistent with the Improved Standard TSs (NUREG-1433 and NUREG-
1434) and 10 CFR 50.36, and will be maintained in accordance with 10 
CFR 50.59. Accordingly, the proposed changes will have no 
significant impact on the condition or performance of structures, 
systems, and components relied upon for accident mitigation. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: January 20, 2004.
    Description of amendment request: This License Amendment Request 
(LAR) proposes selective scope application of the alternate source term 
(AST) for the fuel handling accident (FHA) in accordance with the 
provisions of 10 CFR 50.67. Nuclear Management Company requests the 
Nuclear Regulatory Commission (NRC) review and approval of the AST FHA 
methodology for application to the Prairie Island Nuclear Generating 
Plant. This LAR also proposes revisions to Technical Specifications 
(TS) associated with ensuring that safety analyses assumptions are met 
for a postulated FHA in containment. Based on the AST FHA analyses, 
this LAR proposes to modify TS 3.9.4, ``Containment Penetrations,'' to 
apply during the handling of recently irradiated fuel and require all 
containment penetrations to be closed during handling of recently 
irradiated fuel; and also proposes to remove the requirements of TS 
3.3.5, ``Containment Ventilation Isolation Instrumentation'' relating 
to movement of irradiated fuel assemblies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Technical Specification changes require containment 
integrity during movement of recently irradiated fuel. With this 
change, the Technical Specifications selectively implement 10 CFR 
50.67 alternative source term methodologies for a fuel handling 
accident and implement portions of the approved industry improved 
Standard Technical Specification traveler, TSTF-51, ``Revise 
containment requirements during handling irradiated fuel and core 
alterations' as it applies to TS 3.9.4, ``Containment 
Penetrations.'' This change also removes requirements for 
containment ventilation isolation instrumentation during handling 
irradiated fuel from TS 3.3.5, ``Containment Ventilation Isolation 
Instrumentation'' since the containment purge and inservice purge 
system penetrations which are isolated by this instrumentation will 
be required to be isolated during movement of recently irradiated 
fuel. With the proposed 10 CFR 50.67 alternative source term 
methodologies, these filtration systems are not assumed to function 
during a fuel handling accident involving fuel which is not recently 
irradiated.
    This amendment does not alter the methodology or equipment used 
directly in fuel handling operations. None of the containment 
integrity features including the containment equipment hatch, 
personnel air locks or any other containment penetration

[[Page 29770]]

are used to handle fuel. Therefore, containment integrity and 
ventilation systems, and spent fuel pool ventilation systems are not 
accident initiators and therefore these changes do not increase the 
probability of a previously evaluated accident.
    The total effective dose equivalent (TEDE) doses from the 
analysis supporting this amendment request have been compared to 
equivalent total effective dose equivalent (TEDE) doses estimated 
with the guidelines of Regulatory Guide 1.183 Footnote 7. The new 
values are shown to be comparable to the results of the previous 
analysis.
    A fuel handling accident analysis utilizing alternative source 
term methodologies allowed by 10 CFR 50.67 demonstrated that the 
dose consequences of a postulated fuel handling accident remain 
within the limits of 10 CFR 50.67 without taking credit for 
containment closure or ventilation systems assuming the fuel has not 
recently been in a critical reactor. The alternative source term 
fuel handling accident analysis also demonstrated that the more 
restrictive dose guidelines of Regulatory Guide 1.183 are also met 
without taking credit for these mitigation features. Since the 
alternative source term fuel handling accident analysis results are 
within the regulatory limits and regulatory guidelines without 
taking credit for these mitigation features, revising this Technical 
Specification for containment closure does not involve a significant 
increase in the consequences of a previously evaluated accident.
    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.
    The proposed Technical Specification changes require containment 
integrity during movement of recently irradiated fuel. With this 
change, the Technical Specifications selectively implement 10 CFR 
50.67 alternative source term methodologies for a fuel handling 
accident and implement portions of the approved industry improved 
Standard Technical Specification traveler, TSTF-51, ``Revise 
containment requirements during handling irradiated fuel and core 
alterations'' as it applies to TS 3.9.4, ``Containment 
Penetrations.'' This change also removes requirements for 
containment ventilation isolation instrumentation during handling 
irradiated fuel from TS 3.3.5, ``Containment Ventilation Isolation 
Instrumentation'' since the containment purge and inservice purge 
system penetrations which are isolated by this instrumentation will 
be required to be isolated during movement of recently irradiated 
fuel. With the proposed 10 CFR 50.67 alternative source term 
methodologies, these filtration systems are not assumed to function 
during a fuel handling accident involving fuel which is not recently 
irradiated.
    The proposed Technical Specification changes do not involve 
plant design, hardware, system operation, or procedures involved 
with actual handling of irradiated fuel. The proposed changes 
include application of new methodology for fuel handling accident 
analysis and revises requirements for equipment operability during 
movement of irradiated fuel assemblies. These changes do not create 
the possibility for a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed Technical Specification changes require containment 
integrity during movement of recently irradiated fuel. With this 
change, the Technical Specifications selectively implement 10 CFR 
50.67 alternative source term methodologies for a fuel handling 
accident and implement portions of the approved industry improved 
Standard Technical Specification traveler, TSTF-51, ``Revise 
containment requirements during handling irradiated fuel and core 
alterations' as it applies to TS 3.9.4, ``Containment 
Penetrations.'' This change also removes requirements for 
containment ventilation isolation instrumentation during handling 
irradiated fuel from TS 3.3.5, ``Containment Ventilation Isolation 
Instrumentation'' since the containment purge and inservice purge 
system penetrations which are isolated by this instrumentation will 
be required to be isolated during movement of recently irradiated 
fuel. With the proposed 10 CFR 50.67 alternative source term 
methodologies, these filtration systems are not assumed to function 
during a fuel handling accident involving fuel which is not recently 
irradiated.
    The assumptions and input used in the fuel handling accident 
analysis are conservative. The design basis fuel handling accident 
has been defined to identify conservative conditions. The source 
term and radioactivity releases have been calculated pursuant to 
Regulatory Guide 1.183, Appendix B and with conservative assumptions 
concerning prior reactor operations. The control room atmospheric 
dispersion factor has been calculated with conservative assumptions 
associated with the release. These conservative assumptions and 
input ensure that the radiation doses cited in this license 
amendment request are the upper bounds to radiological consequences 
of a fuel handling accident in containment or the spent fuel pool. 
The analysis shows that there is a significant margin between the 
offsite radiation doses calculated for the postulated fuel handling 
accident using the alternate source term and the dose limits of 10 
CFR 50.67 and acceptance criteria of Regulatory Guide 1.183. The 
proposed changes will not degrade the plant protective boundaries, 
will not cause a release of fission products to the public, and will 
not degrade the performance of any structures, systems, and 
components important to safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

[[Page 29771]]

Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to pdr@nrc.gov.

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: April 17, 2003, as supplemented 
July 29, 2003.
    Brief description of amendments: These amendments revise the 
Required Actions requiring suspension of operations involving positive 
reactivity additions and various notes that preclude reduction of boron 
concentration.
    Date of issuance: May 6, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 266 and 243.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28841).
    The July 29, 2003, letter clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register on May 27, 2003 (68 
FR 28841).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated May 6, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 5, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications to adopt the provisions of Industry/Technical 
Specification Task Force change TSTF-359, ``Increase Flexibility in 
Mode Restraints.''
    Date of issuance: April 29, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 213, 207.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 17, 2004 (69 
FR 7520)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 29, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 5, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications to adopt the provisions of Industry/Technical 
Specification Task Force change TSTF-359, ``Increase Flexibility in 
Mode Restraints.''
    Date of issuance: April 29, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 221, 203.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 17, 2004 (69 
FR 7520)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 29, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 9, 2004, as 
supplemented by letter dated March 2, 2004.
    Brief description of amendment: The amendment removed the 
pressurizer heatup and cooldown limits, and the associated action and 
surveillance requirements, from the Technical Specifications and placed 
them in the Technical Requirements Manual.
    Date of issuance: May 4, 2004.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 253.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9860).
    The March 2, 2004, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 4, 2004.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Units Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: July 18, 2002, as supplemented 
November 14, 2002, and December 11, 2003.
    Brief description of amendments: The amendments relocate Technical 
Specification (TS) 3/4 9.7 regarding the Spent Fuel Storage Pool 
Building cranes and TS 3/4 9.13 (Unit 1) and TS 3/4 9.12 (Unit 2) 
regarding spent fuel cask cranes to the respective units' Updated Final 
Safety Analysis Report.
    Date of Issuance: April 28, 2004
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 190 and 134
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50954). The November 14, 2002, and December 11, 2003, supplements did 
not affect the original proposed no significant hazards determination, 
or expand the scope of the request as noticed in the Federal Register.
    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.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: May 22, 2002, as supplemented by 
letters dated December 5, 2002, and February 11, 2004.
    Brief description of amendment: The amendment revised Technical 
Specification 6.9.1.11.b to add two NRC-approved topical reports to the 
Core Operating Limits Report methodology list, and delete superseded 
reports. Also, the method of listing topical reports was revised to be 
consistent with Technical Specifications Task Force 363, which has been 
approved by the NRC.
    Date of Issuance: May 6, 2004.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 191.
    Facility Operating License No. DPR-67: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42827).

[[Page 29772]]

The supplemental letters provided clarifying information that was 
within the scope of the initial notice and did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 6, 2004.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: August 22, 2003, as supplemented 
by letters dated January 12 and March 11, 2004.
    Brief description of amendment: The amendment revised Section 
3.7.1, ``Service Water (SW) System and Ultimate Heat Sink (UHS),'' by 
adding a new Condition G to allow continued operation with short-term 
elevated UHS temperatures.
    Date of issuance: May 7, 2004.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 113.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 30, 2003 (68 
FR 56344).
    The January 12 and March 11, 2004, 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 staff's related evaluation of the 
amendment is contained in a Safety Evaluation dated May 7, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 30, 2004.
    Brief description of amendment: The amendment relocates the 
requirements for hydrogen monitors from the Technical Specifications to 
the Technical Requirements Manual.
    Date of issuance: May 13, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 174.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9862).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 13, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 25, 2003, as supplemented on 
December 5, 2003
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) 2.1.4, ``Reactor Coolant System (RCS) Leakage 
Limits,'' by (1) adding a requirement for no RCS pressure boundary 
leakage, (2) combining the existing RCS leakage limits into a format 
similar to the Improved Standard TS (ISTS), and (3) replacing the 
existing basis associated with this TS with a basis similar in format 
and content to the ISTS.
    Date of issuance: May 7, 2004.
    Effective date: As of the date of issuance, to be implemented 
within 90 days from issuance.
    Amendment No.: 226.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 19, 2003 (68 FR 
49818).
    The December 5, 2003, supplemental letter provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 7, 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: December 30, 2003, and its 
supplement dated March 11, 2004.
    Brief description of amendments: The amendments eliminate the 
requirements in the technical specifications associated with hydrogen 
recombiners and hydrogen monitors.
    Date of issuance: May 4, 2004.
    Effective date: May 4, 2004, and shall be implemented within 60 
days from the date of issuance.
    Amendment Nos.: Unit 1--168; Unit 2--169.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9864).
    The March 11, 2004, supplemental letter provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 4, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 14th May 2004.
    For the Nuclear Regulatory Commission.
Eric J. Leeds,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 04-11507 Filed 5-24-04; 8:45 am]
BILLING CODE 7590-01-P