[Federal Register Volume 69, Number 148 (Tuesday, August 3, 2004)]
[Notices]
[Pages 46582-46596]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-17346]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 9, 2004 through July 22, 2004. The last 
biweekly notice was published on July 20, 2004 (69 FR 43457).

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 
01F21, 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

[[Page 46583]]

consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
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 
[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 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 
[email protected].

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of amendment request: May 21, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications 5.6.6, ``Reactor Coolant System (RCS) 
Pressure and Temperature Limits Report (PTLR),'' by adding a reference 
to the use of previous Nuclear Regulatory Commission approved Code 
Cases N-640 and N-588 as acceptable methods for determining reactor 
pressure vessel (RPV) pressure temperature (P-T) limits.
    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 TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The use of Code Cases N-588 and N-640 has been approved for 
Braidwood and Byron Stations. The use of P-T limits based on these 
Code Cases will continue to ensure that

[[Page 46584]]

the RPV integrity is maintained under all conditions.
    Thus there is no increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed TS 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 the use or installation of 
new equipment. No equipment will be operated in a new or different 
manner. No new or different system interactions are created and no 
new processes are introduced. The proposed change will not introduce 
any new failure mechanisms, malfunctions, or accident initiators not 
already considered in the design and licensing bases.
    Based on this evaluation, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed TS change does not involve a significant 
reduction in a margin of safety?
    The P-T limits provide assurance that RPV integrity is 
maintained. The use of Code Cases N-588 and N-640 has been 
previously approved by the NRC for Braidwood and Byron Stations and 
will continue to ensure that RPV integrity is maintained.
    Thus, there is no 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC-Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, 
Pennsylvania

    Date of amendment request: June 28, 2004.
    Description of amendment request: The proposed amendment would 
revise the BVPS-1 Technical Specification (TS) 4.4.5.4.a.8 to modify 
the definition of steam generator (SG) tube inspection to exclude the 
portion of the tube within the tube sheet below the W* distance. The W* 
distance is defined as the distance from the top of the tube sheet to 
the bottom of the W* length (7.0 in. on the hot leg side) including the 
distance from the top of the tube sheet to the bottom of the WEXTEX 
(Westinghouse explosive tube expansion) Transition (approximately 0.25 
in.) plus uncertainties (0.12 in.). The proposed amendment would also 
revise the SG tube repair criteria of TS 4.4.5.4.a.6 to indicate that 
service-induced degradation within the W* distance or less than 8.0 in. 
below the top of the tube sheet shall be repaired upon detection. The 
proposed amendment would also add TS 4.4.5.2.e to require a 100% 
rotating pancake coil probe inspection of the hot leg tube sheet W* 
distance, add new W* terminology definitions in TS 4.4.5.4.a.11, and 
add a new reporting criteria for W* inspection information to TS 
4.4.5.5.d.1 and TS 4.4.5.5.e. This proposed amendment would be 
effective for only one operating cycle, as the licensee plans to 
replace SGs during the 2006 refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change modifies the [BVPS-1] TSs to incorporate 
steam generator (SG) tube inspection scope based on WCAP-14797, 
Revision 2 [``Generic W* Tube Plugging Criteria for 51 Series Steam 
Generator Tubesheet Region WEXTEX Expansions,'' dated March 2003 
(proprietary)]. Of the various accidents evaluated in the [BVPS-1] 
Updated Final Safety Analysis Report (UFSAR), the proposed changes 
only affect the steam generator tube rupture (SGTR) event evaluation 
and the postulated steam line break (SLB) accident evaluation. Loss-
of-coolant accident (LOCA) conditions cause a compressive axial load 
to act on the tube. Therefore, since the LOCA tends to force the 
tube into the tubesheet rather than pull it out, it is not a factor 
in this amendment request. Another faulted load consideration is a 
safe shutdown earthquake (SSE); however, the seismic analysis of 
Series 51 steam generators has shown that axial loading of the tubes 
is negligible during an SSE.
    For the SGTR event, the required structural margins of the steam 
generator tubes will be maintained by the presence of the tubesheet. 
Tube rupture is precluded for cracks in the Westinghouse explosive 
tube expansion (WEXTEX) region due to the constraint provided by the 
tubesheet. Therefore, Regulatory Guide (RG) 1.121, ``Bases for 
Plugging Degraded PWR [pressurized-water reactor] Steam Generator 
Tubes,'' margins against burst are maintained for both normal and 
postulated accident conditions.
    The W* length supplies the necessary resistive force to preclude 
pullout loads under both normal operating and accident conditions. 
The contact pressure results from the WEXTEX expansion process, 
thermal expansion mismatch between the tube and tubesheet and from 
the differential pressure between the primary and secondary side. 
The proposed changes do not affect the other systems, structures, 
components or operational features. Therefore, the proposed change 
results in no significant increase in the probability of the 
occurrence of an SGTR or SLB accident.
    The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary 
leakage flow through a postulated broken tube is not affected by the 
proposed change since the tubesheet enhances the tube integrity in 
the region of the WEXTEX expansion by precluding tube deformation 
beyond its initial expanded outside diameter. The resistance to both 
tube rupture and collapse is strengthened by the tubesheet in that 
region. At normal operating pressures, leakage from primary water 
stress corrosion cracking (PWSCC) below the W* length is limited by 
both the tube-to-tubesheet crevice and the limited crack opening 
permitted by the tubesheet constraint. Consequently, negligible 
normal operating leakage is expected from cracks within the 
tubesheet region.
    SLB leakage is limited by leakage flow restrictions resulting 
from the crack and tube-to-tubesheet contact pressures that provide 
a restricted leakage path above the indications and also limit the 
degree of crack face opening compared to free span indications. The 
total leakage, that is, the combined leakage for all such tubes 
meet[s] the industry performance criterion, plus the combined 
leakage developed by any other alternate repair criteria, will be 
maintained below the maximum allowable SLB leak rate limit, such 
that off-site doses are maintained less than 10 CFR 100 guideline 
values and the limits evaluated in the [BVPS-1] UFSAR.
    Therefore, based on the above evaluation, the proposed changes 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed changes do not introduce any changes or 
mechanisms that create the possibility of a new or different kind of 
accident. Tube bundle integrity is expected to be maintained for all 
plant conditions upon implementation of the W* methodology.
    The proposed changes do not introduce any new equipment or any 
change to existing equipment. No new effects on existing equipment 
are created nor are any new malfunctions introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed changes maintain the required structural 
margins of the steam generator tubes for both normal and accident 
conditions. NRC [Nuclear Regulatory Commission] Regulatory Guide 
(RG) 1.121 is used as the basis in the development of the W* 
methodology for determining that steam generator tube integrity 
considerations are

[[Page 46585]]

maintained within acceptable limits. RG 1.121 describes a method 
acceptable to the NRC staff for meeting General Design Criteria 14, 
15, 31, and 32 by reducing the probability and consequences of an 
SGTR. RG 1.121 concludes that by determining the limiting safe 
conditions of tube wall degradation beyond which tubes with 
unacceptable cracking, as established by inservice inspection, 
should be removed from service or repaired, the probability and 
consequences of a[n] SGTR are reduced. This RG uses safety factors 
on loads for tube burst that are consistent with the requirements of 
Section III of the American Society for Mechanical Engineers (ASME) 
[Boiler and Pressure Vessel] Code.
    For primarily axially oriented cracking located within the 
tubesheet, tube burst is precluded due to the presence of the 
tubesheet. WCAP-14797, Revision 2, defines a length, W*, of 
degradation free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure induced forces (with 
applicable safety factors applied). Application of the W* criteria 
will preclude unacceptable primary-to-secondary leakage during all 
plant conditions. The methodology for determining leakage provides 
for large margins between calculated and actual leakage values in 
the W* criteria.
    Plugging of steam generator tubes reduces the reactor coolant 
flow margin for core cooling. Implementation of W* methodology at 
[BVPS-1] will result in maintaining the margin of flow that may have 
otherwise been reduced by tube plugging.
    Based on the above, it is concluded that the proposed changes do 
not result in a significant reduction [in a margin of safety] as 
defined in the [UFSAR] or [B]ases of the plant [TSs].

    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 O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: July 8, 2004.
    Description of amendment request: The amendment request proposes to 
delete one-time use footnotes that have expired or have already been 
used from the Crystal River Unit 3 (CR-3) Improved Technical 
Specifications (ITS). Specifically, obsolete notes will be removed from 
ITS 3.8.1, ``AC Sources--Operating (Emergency Diesel Generator),'' ITS 
3.7.9, ``Nuclear Services Seawater System,'' and ITS 3.7.18, ``Control 
Complex Cooling System.'' This change is administrative in nature and 
does not alter any operating license requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below and states that the amendment 
request:
    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Each footnote was added to ITS through the license amendment 
process. The activities supported by the footnotes were performed 
and, therefore, the footnotes have no further utility. Deleting the 
footnotes is administrative in nature and does not affect plant 
conditions that could impact accident probability or consequences. 
Therefore, granting this LAR [license amendment request] does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    The proposed license amendment deletes footnotes that were used 
on a one-time basis for several specifications. The proposed LAR 
will not result in changes to the design, physical configuration of 
the plant or the assumptions made in the safety analysis. Therefore, 
the proposed change will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The deletion of the footnotes from the ITS does not affect 
properties of plant components or their operation. Therefore, 
granting this LAR 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602
    NRC Acting Section Chief: Michael L. Marshall, Jr.

Indiana Michigan Power Company, Docket Nos. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: June 25, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) and the bases to reduce the 
temperature at which shutdown and control rod drop tests are performed 
from greater than or equal to 541 degrees Fahrenheit to greater than or 
equal to 500 degrees Fahrenheit. Additionally, the proposed amendment 
would make format changes to improve the TS appearance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The probability of occurrence of an accident previously evaluated 
is not altered by the proposed amendment. The proposed change does not 
impact the integrity of the reactor coolant system pressure boundary 
and, therefore, does not increase the potential for the occurrence of a 
loss-of-coolant accident. The change does not make any physical changes 
to the facility design, material or construction standards, and the 
proposed change is not an initiator or contributor to any currently 
evaluated accident. The format changes are intended to improve 
appearance, and do not alter any requirements. Thus, neither the 
probability nor the consequences of a previously analyzed accident are 
significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The rod drop test is routinely performed during each refueling 
outage. Decreasing the test temperature will not create the possibility 
of a new or different accident. The proposed test conditions remain 
bounded by the analysis of record since the rod drop time assumed in 
the accident analysis will not be changed. The format changes are 
intended to improve appearance, and do not alter any requirements. 
Since no new failure modes are associated with the proposed changes, 
the proposed amendment 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 Technical Specification change does not involve a significant 
reduction in margin because the acceptance

[[Page 46586]]

criterion for the rod drop time will not change. The proposed change 
will reduce the minimum rod drop test temperature from greater than or 
equal to 541 degrees Fahrenheit to greater than or equal to 500 degrees 
Fahrenheit. This will slightly increase the measured test rod drop 
time. The measured test rod drop time, however, will be within the 
current Technical Specification limit of 2.4 seconds. The format 
changes are intended to improve appearance, and do not alter any 
requirements. Therefore, the margin of safety is not impacted by the 
proposed amendment.

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

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

    Date of amendment request: July 15, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Section 3.8.1, AC Sources--
Operating, Condition B, to extend the allowed outage time for one 
Diesel Generator (DG) inoperable from 7 days to 14 days and TS Section 
3.8.3, Diesel Fuel Oil, Lube Oil, and Starting Air, Limiting Condition 
for Operation, to allow the use of temporary fuel oil storage tanks to 
supply the required fuel oil storage inventory.
    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 Standby AC Power System (Diesel Generators) provides onsite 
electrical power to vital systems should offsite electrical power be 
interrupted. It is not an initiator to any accident previously 
evaluated. Therefore, the extended period of operation with one diesel 
generator inoperable and the seven day required fuel oil supply being 
provided in part by temporary storage tanks will not increase the 
probability of an accident previously evaluated.
    The Standby AC Power System acts to mitigate the consequences of 
design basis accidents that assume a loss of offsite power. For that 
purpose, redundant diesel generators are provided to protect against a 
single failure. During the Technical Specification seven day allowed 
outage time, an operating unit is allowed by the Technical 
Specifications to remove one diesel generator from service, thereby 
losing this single failure protection. During the requested fourteen 
day allowed outage time for fuel oil storage tank cleaning and coating 
maintenance activities, the inoperable diesel generator will be 
maintained available to start and load, with a minimum of five (5) 
hours of fuel available in the day tank. Manual actions contained in 
approved procedures to provide fuel from temporary storage tanks to 
either the operable diesel generator or the inoperable but available 
diesel generator will be implemented. A risk evaluation determined that 
the probability of failure to implement the contingency actions is 
sufficiently low that it does not adversely impact the availability of 
the Standby AC Power System.
    The vulnerability to external events, seismic, high winds and fire, 
was also evaluated and judged to be not significant due to the low 
probability of these events during the period of time this proposed 
amendment will be in effect, and the defense in depth strategies being 
put in place during the tank maintenance activities.
    In the event that fuel stored in the temporary tanks is not 
available to support full load operation of the diesel generator beyond 
four (4) days, replenishment of fuel oil from offsite can be 
accomplished in approximately 24 hours through the use of existing 
purchase orders for fuel oil and diesel fuel analysis. Therefore, 
during the period of the extended allowed outage time and the use of 
temporary fuel oil storage tanks, there is no significant increase in 
the 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.
    Operation with one diesel generator inoperable but available for an 
extended period or with part of the required diesel fuel stored in 
temporary tanks does not involve any new mode of plant operation or 
different function for plant equipment. Operation in this configuration 
does introduce proceduralized manual actions to supply fuel to either 
diesel generator from the permanent storage tank or the temporary tank. 
These actions can be accomplished within the five hours of full load 
diesel operation from fuel stored in the day tank. A risk evaluation 
determined that the probability of failure to implement the contingency 
actions is sufficiently low that it does not adversely impact the 
availability of the Standby AC Power System. There are no new accident 
precursors generated due to this temporary extension of allowed outage 
time or the use of a temporary fuel oil storage system.
    3. Do the proposed changes involve a significant reduction in the 
margin of safety?
    Response: No.
    A single failure of the operable fuel oil transfer pump could 
prevent DG operation beyond five hours. Proceduralized manual actions 
to supply fuel to either diesel generator from the permanent storage 
tank or the temporary tank will be implemented to mitigate this single 
failure vulnerability. These actions can be accomplished within the 
five (5) hours of full load diesel operation from fuel stored in the 
day tank. A risk evaluation determined that the probability of failure 
to implement the contingency actions is sufficiently low that it does 
not adversely impact the availability of the Standby AC Power System. 
Therefore, during the extended allowed outage time and the use of a 
temporary fuel oil storage system, the Standby AC Power System 
maintains the ability to provide a source of on-site AC power adequate 
for maintaining the safe shutdown of the reactor following abnormal 
operational transients and postulated accidents.
    IEEE [Institute of Electrical and Electronics Engineers] Design 
Standard 308-1970, ``IEEE Criteria for Class 1E Electric Systems for 
Nuclear Power Generating Station,'' Section 5.2.4, ``Standby Power 
Supply,'' Paragraph 6), ``Energy Storage,'' contains the requirement 
for stored energy capacity to be the longer of (a) seven days or (b) 
time required to replenish the energy from sources away from the 
generating unit's site following the limiting design basis event. 
Cooper Nuclear Station's Updated Safety Analysis Report documents that 
the Standby AC Power System conforms to the applicable sections of IEEE 
308-1970.
    The Diesel Generator Diesel Oil Storage and Transfer System will be 
configured to ensure a minimum fuel oil inventory to support greater 
than four (4) days of full load diesel generator operation is 
maintained in the operable permanent storage tank. Existing cross-

[[Page 46587]]

tie capabilities in the fuel storage and transfer system piping, in 
conjunction with proceduralized manual actions, ensure the four day 
fuel supply is available to either diesel generator. The remaining 
three (3) day fuel supply will be stored in temporary non-Class I tanks 
and would potentially be vulnerable to external events. The 
vulnerability to external events, seismic, high winds and fire, was 
evaluated and judged to be not significant due to the low probability 
of these events during the period of time this proposed amendment will 
be in effect, and the defense in depth strategies being put in place 
during the tank maintenance activities.
    In the event that fuel stored in the temporary tanks is not 
available to support full load operation of the diesel generator beyond 
four (4) days, replenishment of fuel oil from offsite can be 
accomplished in approximately 24 hours through the use of existing 
purchase orders for fuel oil and diesel fuel analysis.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of amendment request: June 8, 2004.
    Description of amendment request: The Humboldt Bay Power Plant, 
Unit 3, is a decommissioning nuclear power plant that was permanently 
shutdown in July 1976. The plant is currently in a safe storage 
(SAFSTOR) condition to ensure that necessary plant systems will be 
operated and maintained as needed to preserve safe conditions within 
the facility to prevent deterioration until active decommissioning can 
commence. All spent fuel is stored in the spent fuel pool. Pacific Gas 
and Electric Company (PG&E) has proposed a license amendment to clarify 
the technical specifications applicability to current plant conditions 
and practices. Specifically, the requested changes clarify that:
    (1) Fuel fragments within the spent fuel pool totaling less than 
one fuel assembly and damaged fuel assembly UD-6N do not have to be 
stored in containers made of neutron absorbing material. Furthermore, 
that one additional assembly can be removed from a neutron absorbing 
container to perform fuel handling activities.
    (2) The control station for Humboldt Bay Units 1 and 2 is 
considered to be anywhere on the +27 foot operating deck.
    (3) References to certain technical specification section 
designators that contain typographical errors have been corrected.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed changes provide either clarification to reflect 
plant conditions or correct typographical errors. Existing accident 
analysis assumptions bound the proposed addition of not storing fuel 
fragments, which may be considered as less than or equal to a fuel 
assembly, in a container made with neutron absorbing material. The 
proposed changes involve no changes to plant systems or accident 
analysis, and as such, do not affect initiators of analyzed events 
or assumed mitigation of accidents. Therefore, the proposed changes 
do not increase the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different type of accident from any accident previously 
evaluated?
    No. The proposed changes provide either clarification to reflect 
plant conditions or correct typographical errors. Existing accident 
analysis assumptions bound the proposed addition of not storing fuel 
fragments, which may be considered as less than or equal to a fuel 
assembly, in a container made with neutron absorbing material. The 
proposed changes do not involve a physical alteration to the plant, 
add any new equipment, or require existing equipment to be operated 
in a manner different from the present design. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed changes provide either clarification to reflect 
existing plant conditions or correct typographical errors. Existing 
accident analysis assumptions bound the proposed addition of not 
storing fuel fragments, which may be considered as less than or 
equal to a fuel assembly, in a container made with neutron absorbing 
material. They have no effect on plant equipment, operating 
practices or safety analysis assumptions. 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: Richard F. Locke, Esquire, Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Claudia Craig.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of amendment request: June 23, 2004.
    Description of amendment request: The Humboldt Bay Power Plant, 
Unit 3, is a decommissioning nuclear power plant that was permanently 
shutdown in July 1976. The plant is currently in a safe storage 
(SAFSTOR) condition to ensure that necessary plant systems will be 
operated and maintained as needed to preserve safe conditions within 
the facility to prevent deterioration until active decommissioning can 
commence. All spent fuel is stored in the spent fuel pool. Currently, 
the facility operating license only allows maintaining the facility in 
SAFESTOR. At the time the license condition for SAFSTOR was specified, 
Pacific Gas and Electric Company (PG&E), the licensee, had intended to 
maintain SAFSTOR until the Department of Energy (DOE) established a 
permanent repository for spent fuel. The licensee has recently 
reassessed its near-term options for the facility and in December of 
2003 applied for a license to store its spent fuel in an onsite dry 
cask independent spent fuel storage installation (ISFSI). Moving the 
spent fuel to an ISFSI would permit the licensee to begin significant 
decommissioning activities. Consequently, PG&E has submitted a license 
amendment request to permit the licensee to proceed with 
decontamination and decommissioning activities in accordance with 
applicable NRC requirements and the regulations for decommissioning 
reactors in 10 CFR 50.82.
    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:


[[Page 46588]]


    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change eliminates the restriction to remain in 
SAFSTOR status, and allows PG&E to take actions necessary to 
decommission and decontaminate the facility in accordance with NRC 
regulations. The proposed change involves no changes to plant 
systems or accident analysis, and as such, do not affect initiators 
of analyzed events or assumed mitigation of accidents. Therefore, 
the proposed changes do not increase the probability or consequences 
of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different type of accident from any accident previously 
evaluated?
    No. The proposed change eliminates the restriction to remain in 
SAFSTOR status, and allows PG&E to take actions necessary to 
decommission and decontaminate the facility in accordance with NRC 
regulations. The proposed change does not involve a physical 
alteration to the plant, add any new equipment, or require existing 
equipment to be operated in a manner different from the present 
design. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change eliminates the restriction to remain in 
SAFSTOR status, and allows PG&E to take actions necessary to 
decommission and decontaminate the facility in accordance with NRC 
regulations. The proposed change has no effect on plant equipment, 
operating practices or safety analysis assumptions. 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: Richard F. Locke, Esquire, Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Claudia Craig.

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

    Date of amendment request: March 31, 2004.
    Description of amendment request: The proposed change will allow 
operation in regions of the power/flow map currently restricted by the 
requirements of interim corrective actions (ICAs) and certain limiting 
conditions for operations (LCOs) of Technical Specification 3.4.1. The 
oscillation power range monitor (OPRM) will allow operations in the 
regions restricted by the administrative controls mentioned above by 
using inputs from the local power range monitoring (LPRM) system to 
monitor core conditions and generate a reactor protection system (RPS) 
trip when required to prevent a violation of the minimum critical power 
ratio (MCPR) safety limit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's 
analysis is presented below:

1. Does the Proposed Change Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated?

    The proposed change would allow operation in regions of the 
power/flow map currently restricted by administrative controls. The 
purpose of the administrative controls were to ensure adequate 
capability to detect and suppress conditions consistent with the 
onset of a thermal-hydraulic (T-H) event which is postulated to 
cause a violation of the MCPR safety limit. The mitigation of a T-H 
instability event will be ensured by the RPS trip signal generated 
by the OPRM prior to challenging the MCPR safety limit. Since 
automatic protective functions of the OPRM will be replacing 
administrative controls which require operator action, the 
probability or consequence of a T-H instability event is not 
significant. Therefore, the proposed change does not result in a 
significant increase in the probability or consequence 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?

    The proposed change would allow operation in regions of the 
power/flow map currently restricted by administrative controls. The 
OPRM system uses inputs from the LPRMs to monitor core conditions 
and generate a RPS trip when required. Quality requirements for 
software design, testing, implementation and module self-testing of 
the OPRM system provide assurance that no new equipment malfunctions 
due to software errors are created. The design of the OPRM system 
also ensures that neither operation nor malfunction of the OPRM 
system will adversely impact the operation of other systems, and no 
accident or equipment malfunction of these other systems could cause 
the OPRM system to malfunction or cause a different kind of 
accident. Therefore, operation with the OPRM system does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.

3. Does the Proposed Change Involve a Significant Reduction in a Margin 
of Safety?

    The proposed change would allow operation in regions of the 
power/flow map currently restricted by administrative controls. The 
margin of safety for the unmitigated T-H instability event will not 
be significantly reduced due to the capability of the OPRM to 
automatically detect and suppress conditions which might result in 
an MCPR safety limit violation. The automatic functions of the OPRM 
will be replacing administrative controls which rely on operator 
action to prevent an unmitigated T-H instability event. The OPRM 
will maintain the margin of safety while significantly reducing the 
burden on the control room operators. Therefore, operation with the 
OPRM system does not involve a significant reduction in a margin of 
safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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: June 29, 2004.
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TS) to implement the following 
miscellaneous changes: (1) Revise the reporting period of TS 2.2.5 from 
30 days to 60 days for the safety limit violations Licensee Event 
Report, (2) revise the frequency of Surveillance Requirement (SR) 
3.4.3.1.2 of TS 3.4.3.1, ``Pressurizer Heatup and Cooldown Limits,'' to 
reflect pressurizer spray cyclic limits being governed by the 
temperature differentials between the spray nozzle and the spray line, 
(3) revise TS 5.5.2.11.f.1 of TS 5.5.2.11, ``Steam Generator (SG) Tube 
Surveillance Program,'' to correct typographical errors, (4) remove TS 
5.5.2.14, ``Configuration Risk Management Program (CRMP),'' in 
accordance with Federal Register Notice Vol. 64, No. 137 (July 19, 
1999), and (5) revise TS 5.7.1.5, ``Core Operating Limits Report 
(COLR),'' to delete revision numbers and dates from the referenced 
documents in this section consistent with the NRC-approved industry 
Technical Specifications Task Force (TSTF) Standard Technical 
Specifications Traveler number TSTF-

[[Page 46589]]

363, ``Revise Topical Report References in ITS (Improved Technical 
Specifications) 5.6.5 COLR,'' and incorporate editorial corrections.
    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.
    Southern California Edison (SCE) proposes to modify the San 
Onofre Units 2 and 3 Technical Specifications (TS) to accomplish 
several improvements by providing consistency with current Code of 
Federal Regulations (CFR) Licensee Event Report (LER) reporting 
requirements, clarifying a pressurizer heatup/cooldown Surveillance 
Requirement, TS editorial corrections, removing TS redundancy to the 
Maintenance Rule in accordance with Federal Register Notice Vol. 64, 
No. 137 (July 19, 1999), and eliminating need for TS amendment 
requests for cited Core Operating Limits Report (COLR) reference 
revisions consistent with the NRC approved Industry Technical 
Specifications Task Force (TSTF) Standard Technical Specifications 
Traveler number TSTF-363, ``Revise Topical Report References in ITS 
(Improved Technical Specifications) 5.6.5 COLR.'' These proposed 
changes do not involve any change in the design or operation of the 
plant. 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.
    Modifying the Technical Specifications to provide consistency 
with current CFR LER reporting requirements, clarify a pressurizer 
heatup/cooldown Surveillance Requirement, incorporate editorial 
corrections, remove TS redundancy to the Maintenance Rule in 
accordance with Federal Register Notice Vol. 64, No. 137 (July 19, 
1999), and to eliminate need for TS amendment requests for cited 
COLR reference revisions consistent with the NRC approved Industry 
Technical Specifications Task Force (TSTF) Standard Technical 
Specifications Traveler number TSTF-363, ``Revise Topical Report 
References in ITS (Improved Technical Specifications) 5.6.5 COLR'' 
does not involve any change in the design or operation of the plant. 
Therefore, a possibility of a new or different kind of accident from 
any accident previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Evaluation of these proposed modifications to the Technical 
Specifications to provide consistency with current CFR LER reporting 
requirements, clarify a pressurizer heatup/cooldown Surveillance 
Requirement, incorporate editorial corrections, remove TS redundancy 
to the Maintenance Rule in accordance with Federal Register Notice 
Vol. 64, No. 137 (July 19, 1999), and to eliminate need for TS 
amendment requests for cited COLR reference revisions consistent 
with the NRC approved Industry Technical Specifications Task Force 
(TSTF) Standard Technical Specifications Traveler number TSTF-363, 
``Revise Topical Report References in ITS (Improved Technical 
Specifications) 5.6.5 COLR'' does not involve any change in the 
design or operation of the plant and therefore does not create any 
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 Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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: June 30, 2004.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 5.5.2.15, ``Containment Leakage 
Rate Testing Program.'' Specifically, the licensee proposes a one-time 
extension of the ten-year period of the performance-based leakage rate 
testing program for Type A tests as prescribed by Nuclear Energy 
Institute 94-01, Revision 0, ``Industry Guideline for Implementing 
Performance-Based Option of 10 CFR Part 50, Appendix J.'' The ten-year 
interval between integrated leakage rate tests is to be extended to 15 
years from the previous integrated leakage rate tests. Under the 
current TS requirements, which include an allowance of a 15-month 
extension, the next Type A test would be performed during the Cycle 14 
refueling outages currently planned for November 2005 (Unit 2) and June 
2006 (Unit 3). The requested change reflects a one-time deferral of the 
next Type A containment integrated leak rate test to no later than 
March 30, 2010 (Unit 2) and September 9, 2010 (Unit 3). This proposed 
change is based on and has been evaluated using the ``risk informed'' 
guidance in Regulatory Guide 1.174, ``An Approach for Using 
Probabilistic Risk Assessment in Risk-informed Decisions on Plant-
Specific Changes to the Licensing Basis.''
    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 revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10 CFR 
50, Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 10 years, based on past performance, would be 
extended on a one time basis to 15 years from the last Type A test. 
The proposed extension to Type A testing does not involve a 
significant increase in the consequences of an accident since 
research documented in NUREG-1493, ``Performance-Based Containment 
System Leakage Testing Requirements,'' September 1995, has found 
that, generically, very few potential containment leakage paths are 
not identified by Type B and C tests. The NUREG concluded that 
reducing the Type A testing frequency to one per twenty years was 
found to lead to an imperceptible increase in risk. A high degree of 
assurance is provided through testing and inspection that the 
containment will not degrade in a manner detectable only by Type A 
testing. The last Type A tests show leakage to be below acceptance 
criteria, indicating a leak tight containment. Inspections required 
by the American Society of Mechanical Engineers (ASME) Code Section 
XI (Subsections IWE and IWL) and maintenance rule monitoring (10 CFR 
50.65, ``Requirements for Monitoring the Effectiveness of 
Maintenance at Nuclear Power Plants) are performed in order to 
identify indications of containment degradation that could affect 
that leak tightness. Type B and C testing required by Technical 
Specifications will identify any containment opening such as valves 
that would otherwise be detected by the Type A tests. These factors 
show that a Type A test extension will not represent a significant 
increase in the consequences of an accident.
    Therefore, the proposed changes do 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 revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10 CFR 
50, Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 10 years, based on past performance, would be 
extended on a one time basis to 15 years from the last Type A test. 
The proposed extension to Type A testing cannot create the 
possibility of a new or different type of accident since there are 
no physical changes being made to the plant and there are no changes 
to the operation of the plant that could introduce a new failure 
mode creating

[[Page 46590]]

an accident or affecting the mitigation of an accident. Therefore, 
the proposed changes do 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 revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10 CFR 
50, Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 10 years, based on past performance, would be 
extended on a one time basis to 15 years from the last Type A test. 
The proposed extension to Type A testing will not significantly 
reduce the margin of safety. The NUREG 1493, ``Performance-Based 
Containment System Leakage Testing Requirements,'' September 1995, 
generic study of the effects of extending containment leakage 
testing found that a 20 year extension in Type A leakage testing 
resulted in an imperceptible increase in risk to the public. NUREG 
1493 found that, generically, the design containment leakage rate 
contributes about 0.1 percent to the individual risk and that the 
decrease in Type A testing frequency would have a minimal affect on 
this risk since 95% of the potential leakage paths are detected by 
Type C testing. Regular inspections required by the American Society 
of Mechanical Engineers (ASME) Code Section XI (Subsections IWE and 
IWL) and maintenance rule monitoring (10 CFR 50.65, ``Requirements 
for Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants) will further reduce the risk of a containment leakage path 
going undetected.
    Therefore the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve 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: Stephen Dembek.

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 amendment request: June 28, 2004.
    Description of amendment request: The proposed amendments would 
revise existing Technical Specifications (TSs) 3.4.13, ``RCS [Reactor 
Coolant System] Operational Leakage,'' TS 5.59, ``Steam Generator [SG] 
Tube Surveillance Program,'' and TS 5.610, ``Steam Generator Tube 
Inspector Report.'' It would also add a new TS 3.4.17, ``Steam 
Generator Tube Integrity.'' These changes would facilitate the 
implementation of industry initiative NEI [Nuclear Energy Institute] 
97-06, ``Steam Generator Program Guidelines,'' which would allow for a 
comprehensive, performance-based approach to managing SG performance at 
Farley Nuclear Plant, Units 1 and 2.
    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 change requires a Steam Generator Program that 
includes performance criteria that will provide reasonable assurance 
that the steam generator (SG) tubing will retain integrity over the 
full range of operating conditions (including startup, operation in 
the power range, hot standby, cooldown and all anticipated 
transients included in the design specification). The SG performance 
criteria are based on tube structural integrity, accident induced 
leakage, and operational LEAKAGE.
    The structural integrity performance criterion is:
    ``All inservice SG tubes shall retain structural integrity over 
the full range of normal operating conditions (including startup, 
operation in the power range, hot standby and cooldown and all 
anticipated transients included in the design specification) and 
design basis accidents. This includes retaining a safety factor of 
3.0 against burst under normal steady state full power operation 
primary to secondary pressure differential and a safety factor of 
1.4 against burst applied to the design basis accident primary to 
secondary pressure differentials. Apart from the above requirements, 
additional loading conditions associated with the design basis 
accidents, or combination of accidents in accordance with the design 
and licensing basis, shall also be evaluated to determine if the 
associated loads contribute significantly to burst or collapse. In 
the assessment of tube integrity, those loads that do significantly 
affect burst or collapse shall be determined and assessed in 
combination with the loads due to pressure with a safety factor of 
1.2 on the combined primary loads and 1.0 on axial secondary 
loads.''
    The accident induced leakage performance criterion is:
    ``The primary to secondary accident induced leakage rate for all 
design basis accidents, other than a SG tube rupture, shall not 
exceed the leakage rate assumed in the accident analysis in terms of 
total leakage rate for all SGs and leakage rate for an individual 
SG. For FNP Units 1 and 2, leakage is not to exceed 1 gpm [gallons 
per minute] total for all three SGs. Exceptions to the 1 gpm limit 
can be applied if approved by the NRC in conjunction with approved 
alternate repair criteria.''
    The operational LEAKAGE performance criterion is:
    The RCS operational primary to secondary LEAKAGE through any one 
SG shall be limited to 150 gpd [gallons per day].
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents analyzed as part of the plant licensing basis. In 
the analysis of a SGTR event, a bounding primary to secondary 
LEAKAGE rate equal to the operational LEAKAGE rate limits in the 
licensing basis plus the LEAKAGE rate associated with a double-ended 
rupture of a single tube is assumed.
    For other design basis accidents such as main steam line break 
(MSLB), rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). For FNP Units 1 and 2, these analyses 
assume that primary to secondary LEAKAGE for all SGs is 1 gpm. The 
accident induced leakage criterion introduced by the proposed 
changes accounts for tubes that may leak during design basis 
accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed in this change to the TS 
identify the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining reactor coolant pressure 
boundary integrity throughout each operating cycle and in the 
unlikely event of a design basis accident. The performance criteria 
are only a part of the Steam Generator Program required by the 
proposed change to the TS. The program, defined by NEI 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates 
a balance of prevention, inspection, evaluation, plugging, and 
leakage monitoring.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the TS for operational leakage and 
for DOSE EQUIVALENT I-131 in primary coolant to ensure the plant is 
operated within its analyzed condition. The analysis of the limiting 
design basis accident assumes that primary to secondary leak rate 
after the accident is 1 gpm with no more than 500 gpd in any one SG, 
and that the reactor coolant activity levels of DOSE EQUIVALENT I-
131 are at the technical specification values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TS and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TS.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of a MSLB,

[[Page 46591]]

rod ejection, or a reactor coolant pump locked rotor event.
    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 performance based requirements are an improvement 
over the requirements imposed by the current TS.
    Implementation of the proposed Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the Steam Generator Program will be 
an enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the Steam Generator Program to manage SG 
tube inspection, assessment and plugging. The requirements 
established by the Steam Generator Program are consistent with those 
in the applicable design codes and standards and are an improvement 
over the requirements in the current TS.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: Stephanie M. Coffin, Acting.

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

    Date of amendment request: April 26, 2004.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification Section 5.5.12, ``Primary 
Containment Leakage Rate Testing Program'' to reflect a one-time 
deferral of the Type A Containment Integrated Leak Rate Test (ILRT). 
This change would extend the 10 year interval between ILRTs to 15 years 
from the previous ILRT.
    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 Technical Specification change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed revision to Technical Specification 5.5.12 
(``Primary Containment Leakage Rate Testing Program'') involves a 
one-time extension to the current interval for Type A containment 
testing. The current test interval of ten (10) years would be 
extended on a one-time basis to no longer than fifteen (15) years 
from the last Type A test. The proposed Technical Specification 
change does not involve a physical change to the plant or a change 
in the manner which the plant is operated or controlled. The reactor 
containment is designed to provide an essentially leak tight barrier 
against the uncontrolled release of radioactivity to the environment 
for postulated accidents. As such the reactor containment itself and 
the testing requirements invoked to periodically demonstrate the 
integrity of the reactor containment exist to ensure the plant's 
ability to mitigate the consequences of an accident, and do not 
involve the prevention or identification of any precursors of an 
accident. Therefore, the proposed Technical Specification change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The proposed change involves only the extension of the interval 
between Type A containment leakage tests. Type B and C containment 
leakage tests will continue to be performed at the frequency 
currently required by plant Technical Specifications. Industry 
experience has shown, as documented in NUREG-1493, that Type B and C 
containment leakage tests have identified a very large percentage of 
containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is very 
small. HNP [Hatch Nuclear Plant ] Unit 2 ILRT test history supports 
this conclusion. NUREG-1493 concluded, in part, that reducing the 
frequency of Type A containment leak tests to once per twenty (20) 
years leads to an imperceptible increase in risk. The integrity of 
the reactor containment is subject to two types of failure 
mechanisms which can be categorized as (1) activity based and (2) 
time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as design change control and procedural requirements 
for system restoration ensure that containment integrity is not 
degraded by plant modifications or maintenance activities. The 
design and construction requirements of the reactor containment 
itself combined with the containment inspections performed in 
accordance with ASME [American Society of Mechanical Engineers] 
Section XI, the Maintenance Rule and the containment coatings 
program serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is detectable only by 
Type A testing. Therefore, the proposed Technical Specification 
change does not involve a significant increase in the consequences 
of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed revision to the Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
testing. The reactor containment and the testing requirements 
invoked to periodically demonstrate the integrity of the reactor 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident and do not involve the prevention or 
identification of any precursors of an accident. The proposed 
Technical Specification change does not involve a physical change to 
the plant or the manner in which the plant is operated or 
controlled. Therefore, the proposed Technical Specification change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The proposed revision to Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
testing. The proposed Technical Specification change does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The specific requirements 
and conditions of the Primary Containment

[[Page 46592]]

Leakage Rate Testing Program, as defined in Technical 
Specifications, exist to ensure that the degree of reactor 
containment structural integrity and leak-tightness that is 
considered in the plant safety analysis is maintained. The overall 
containment leakage rate limit specified by Technical Specifications 
is maintained. The proposed change involves only the extension of 
the interval between Type A containment leakage tests. Type B and C 
containment leakage tests will continue to be performed at the 
frequency currently required by plant Technical Specifications.
    HNP Unit 2 and industry experience strongly supports the 
conclusion that Type B and C testing detects a large percentage of 
containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is small. The 
containment inspections performed in accordance with ASME Section 
XI, the Maintenance Rule and the Coatings Program serve to provide a 
high degree of assurance that the containment will not degrade in a 
manner that is detectable only by Type A testing. Additionally, the 
on-line containment monitoring capability that is inherent to 
inerted BWR containments allows for the detection of gross 
containment leakage that may develop during power operation. The 
combination of these factors ensures that the margin of safety that 
is inherent in plant safety analysis is maintained. Therefore, the 
proposed Technical Specification 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Stephanie M. Coffin, Acting.

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

    Date of amendment request: June 22, 2004.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS), Appendix A in order to change 
the frequency of the logic system functional test, for the 4 kV 
emergency busses' loss of power instrumentation, from once every 18 
months to once every 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This is a proposed change to the surveillance requirement (SR) 
for the logic system functional test (LSFT) of the loss of power 
(LOP) instrumentation for Plant Hatch Units 1 and 2 (SR 3.3.8.1.4). 
The LOP instrumentation functions to monitor the voltage on the 4 kV 
emergency busses and, if necessary, to disconnect these busses from 
the offsite power source and re-connect them to on-site power. This 
would, of course, be necessary if a bus experienced a loss of, or a 
degraded, voltage. This ensures an adequate response to a loss of 
coolant accident (LOCA) if that accident were to occur 
simultaneously with a loss of off-site power (LOSP). The probability 
of occurrence of a previously evaluated event, such as a LOCA/LOSP, 
will not increase since the LOP instrumentation is not being 
physically altered as a result of this change in such a manner which 
may increase the likelihood of failure. In fact, it is not being 
physically altered at all as a result of this submittal.
    Additionally, no other safety related equipment or components 
designed to prevent the occurrence of a previously evaluated event 
are being physically altered or otherwise affected as a result of 
this TS change request.
    The consequences of a previously evaluated event will not 
increase as a result of revising the surveillance frequency for the 
LOP instrumentation. Review of surveillance histories demonstrates 
adequate performance for the LOP relays in ultimately connecting the 
emergency power sources to the distribution bus, justifying the 
revision in the surveillance frequency. Therefore, the LOP 
instrumentation can be reasonably expected to perform its function 
in a LOCA/LOSP event, even with the revised frequency for the LSFT.
    For the above reasons, the change in the LSFT frequency does not 
involve a significant increase in the probability or consequences of 
a previously evaluated event.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The LOP instrumentation is not being physically altered. 
Furthermore, its operation and maintenance will remain within the 
design bases. The only proposed change is the frequency of the logic 
system functional test. Since no new modes of operation are being 
introduced, a new or different kind of accident from any previously 
evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The function of the LOP instrumentation is to ensure that the 
emergency power distribution busses receive adequate power from 
either the off-site or on-site sources. The LOP relays will initiate 
a transfer of the emergency 4 kV busses to the on-site diesel 
generators on a loss of coolant accident with a concurrent loss of 
off-site power. The diesel logic will then sequence the cooling 
water pumps and other safety related equipment onto their respective 
emergency bus. This sequencing of loads is tested by a different 
surveillance requirement which is not affected by this TS change 
request and has already been revised to a frequency of once per 24 
months. This proposed TS revision only changes the frequency of 
performance of the LSFT for the LOP instrumentation. A review of 
surveillance histories shows that these relays perform adequately in 
the re-connection of the emergency busses to the on-site power 
source. Some problems have been noted in the history review with the 
loss of off-site power annunciation. However, the annunciator does 
not affect the safety function of providing power to the 
distribution bus.
    For the above reasons, the margin of safety is not reduced by 
this proposed Technical Specifications change.

    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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Stephanie M. Coffin, Acting.

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

    Date of amendment request: July 8, 2004 (TS-448)
    Description of amendment request: The proposed amendment requests 
the modification of Technical Specification Section 5.5.12 ``Primary 
Containment Leakage Rate Testing Program'' to allow a one-time 5-year 
extension to the 10-year frequency of the performance-based leakage 
rate testing program for Type A tests. The proposed changes are 
submitted on a risk-informed basis as described in Regulatory Guide 
1.174, An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis. 
The risk-informed analysis supporting the proposed changes indicates 
that the increase in risk from extending the integrated leak rate test 
interval from 10 to 15 years is insignificant.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 46593]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    TVA has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
Amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed revision to TS adds a one-time extension to the 
current interval for Type A testing. The current test interval of 10 
years, based on past performance, would be extended on a one-time 
basis to 15 years from the last Type A test. The proposed extension 
to Type A testing cannot increase the probability of an accident 
previously evaluated since the containment Type A testing extension 
is not a modification and the test extension is not of a type that 
could lead to equipment failure or accident initiation.
    The proposed extension to Type A testing does not involve a 
significant increase in the consequences of an accident since 
research documented in NUREG-1493 has found that, generically, very 
few potential containment leakage paths are not identified by Type B 
and C tests. The NUREG concluded that reducing the Type A (ILRT) 
testing frequency to once per 20 years was found to lead to an 
imperceptible increase in risk. These generic conclusions were 
confirmed by a plant specific risk assessment.
    Testing and the containment inspection programs in place at BFN 
provide a high degree of assurance that the containment will not 
degrade in a manner detectable only by Type A testing. The last four 
Type A tests show leakage to be below acceptance criteria, 
indicating a very leak tight containment. Type B and C testing 
required by TS will identify any containment opening such as valves 
that would otherwise be detected by the Type A tests. Inspections, 
including those required by the American Society of Mechanical 
Engineers code are also performed in order to identify indications 
of containment degradation that could affect that leak tightness.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The change does not create the possibility of a new or 
different kind of accident from any accident previously analyzed. 
The proposed revision to TS adds a one-time extension to the current 
interval for Type A testing. The current test interval of 10 years, 
based on past performance, would be extended on a one-time basis to 
15 years from the last Type A test. The proposed extension to Type A 
testing cannot create the possibility of a new or different type of 
accident since there are no physical changes being made to the plant 
and there are no changes to the operation of the plant that could 
introduce a new failure mode creating an accident or affecting the 
mitigation of an accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. BFN Units 2 and 3 are General Electric BWR/4 plants with 
Mark I primary containments. The Mark I primary containment consists 
of a drywell, which encloses the reactor vessel; reactor coolant 
recirculation system and branch lines of the Reactor Coolant System; 
a toroidal-shaped pressure suppression chamber containing a large 
volume of water; and a vent system connecting the drywell to the 
water space of the suppression chamber. The primary containment is 
penetrated by personnel access hatches, piping, and electrical 
penetrations.
    The integrity of the primary containment penetrations and 
isolation valves is verified through Type B and Type C local leak 
rate tests and the overall leak-tight integrity of the primary 
containment is verified by a Type A integrated leak rate test as 
required by 10 CFR 50, Appendix J, ``Primary Reactor Containment 
Leakage Testing for Water-Cooled Power Reactors.'' These tests are 
performed to verify the essentially leak-tight characteristics of 
the primary containment at the design basis accident pressure. The 
proposed change for a one-time extension of the Type A tests does 
not affect the method for Type A, B, or C testing, or the test 
acceptance criteria. In addition, based on previous Type A testing 
results, TVA does not expect additional degradation during the 
extended period between Type A tests, which would result in a 
significant reduction in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: July 8, 2004.
    Description of amendment request: The proposed amendment will 
revise the Technical Specification (TS) to remove the term ``inter-
rack'' and associated wording from Surveillance Requirements 3.8.4.6 
and 3.8.4.10 for the 125 Volt (V) Direct Current (DC) Electrical Power 
Subsystems of the Emergency Diesel Generators (DGs).
    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 TS change eliminates an inaccurate term and 
associated wording, but the actual TS amendment does not result in 
any change to the actual surveillance field test for the associated 
batteries. The proposed wording will only clarify the surveillances. 
Prior field tests were adequate to verify proper battery connection 
integrity since it tested the inside (inter-tier) jumper cable 
connections as if they were interchangeable with inter-rack. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed TS change does not alter the configuration of 
the plant's 125 V DC Electrical Power Subsystems of the Emergency 
DGs. The change does not directly affect plant operation. The change 
will not result in the installation of any new equipment or system 
or the modification of any existing equipment or systems. No new 
operations procedures, conditions, or modes will be created by this 
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 
margin of safety?
    No. The battery connection continuity check for the 125 V DC 
Electrical Power Subsystems of the Emergency DGs will continue to be 
monitored by the same process as previously performed. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Acting Section Chief: Michael L. Marshall, Jr.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the

[[Page 46594]]

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].
    AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek 
Nuclear Generating Station (OCNGS), Ocean County, New Jersey, Docket 
No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin 
County, Pennsylvania
    Date of application for amendments: March 8, 2004.
    Brief description of amendment: The amendments deleted the License 
Condition entitled ``Long Range Planning Program'' from the OCNGS and 
TMI-1 operating licenses. In addition, for TMI-1, the amendment 
relocated a requirement (regarding surveillance of the depth of water 
in the spent fuel pool) from the Long Range Planning Program to the 
Technical Specifications.
    Date of Issuance: July 13, 2004.
    Effective date: These license amendments are effective as of their 
date of issuance, and shall be implemented within 30 days of issuance.
    Amendment Nos.: 244 and 250
    Facility Operating License Nos. DPR-16 and DPR-50: Amendments 
revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19563 and 19564). The Commission's related evaluation of this amendment 
is contained in a Safety Evaluation dated July 13, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 6, 2003, as supplemented 
February 13 and June 16, 2004.
    Brief description of amendment: The amendment revised the reactor 
building tendon surveillance criteria to incorporate a reference to 
Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a. 
The amendment also includes an administrative change to provide 
consistency between Technical Specification Definition 1.22 (MEMBERS OF 
THE PUBLIC) and the definition contained in 10 CFR 20.1003, and a 
change to correct a typographical error in a reference title.
    Date of issuance: July 13, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 251.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68655) and March 16, 2004 (69 FR 12363). The February 13, 2004, 
supplemental letter provided clarifying information and expanded the 
scope of the application as originally noticed. Therefore, the original 
proposed no significant hazards consideration determination was changed 
and republished. The June 16, 2004, supplement provided clarifying 
information, did not expand the scope of the application and did not 
change the NRC staff's proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 13, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 24, 2003.
    Brief description of amendment: The amendment deleted requirements 
from the Technical Specifications (TSs) 3.7.A.7.c and 4.7.A.7.c 
associated with hydrogen analyzers. The associated TS Bases are also 
deleted.
    Date of issuance: July 22, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 206.
    Facility Operating License No. DPR-35: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19568).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 22, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 14, 2003, as supplemented by 
letters dated January 22, and May 6, 2004.
    Description of amendment request: This license amendment modifies 
Technical Specification (TS) Table 3.3.6.1-1, ``Primary Containment and 
Drywell Isolation Instrumentation,'' Item 1.f, to increase the 
analytical limit for detected temperature and the resulting TS 
Allowable Value related to the setpoint for the Main Steam Line Turbine 
Building Temperature--High system isolation function. Additionally, it 
authorizes the use of the GOTHIC 7.0 computer program to perform 
analyses of main steamline leaks in the turbine building for Perry 
Nuclear Power Plant to replace the currently approved COMPARE computer 
program for performing the analyses listed above.
    Date of issuance: July 9, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 130.

[[Page 46595]]

    Facility Operating License No. NPF-58: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: (69 FR 696) January 6, 
2004.
    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 July 9, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 27, 2004.
    Brief description of amendment: The amendment deletes Technical 
Specification Section 5.6.2.6, ``Post-Accident Sampling,'' requirements 
to maintain a Post-Accident Sampling System.
    Date of issuance: July 6, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 213.
    Facility Operating License No. DPR-72: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19571).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 6, 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, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: October 23, 2002, as 
supplemented by letters dated August 28, 2003, December 11, 2003, 
February 3, 2004, and March 25, 2004.
    Brief description of amendments: These amendments revised Technical 
Specification Section 5.6, ``Design Features--Fuel Storage,'' for St. 
Lucie Units 1 and 2 to include the design of a new cask pit spent fuel 
storage rack for each unit, and increase each unit's spent fuel storage 
capacity by combining the cask pit rack and existing spent fuel pool 
storage rack capacities. The cask pit racks will be used to store spent 
fuel to allow refueling outage fuel offloads and nonoutage fuel 
shuffles and, for Unit 1, to store new fuel prior to loading it into 
the reactor.
    Date of Issuance: July 9, 2004.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 192 and 135.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 28, 2003 (68 FR 
4244), as corrected March 31, 2003 (68 FR 15487). The August 28, 2003, 
December 11, 2003, February 3, 2004, and March 25, 2004, 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 an Environmental Assessment dated July 2, 2004 and in a Safety 
Evaluation dated July 9, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 29, 2004, as supplement by 
letter dated April 8, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification 3.4.9 Pressure Temperature (P/T) limit curve Figures 
3.4.9-1, 3.4.9-2, and 3.4.9-3.
    Date of issuance: July 14, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 204.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12371). The April 8, 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 amendment is contained 
in a Safety Evaluation dated July 14, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 30, 2004, as supplemented by 
letter dated June 17, 2004.
    Brief description of amendment request: The proposed amendment 
would revise the Cooper Nuclear Station (CNS) Technical Specifications 
(TSs), by adding a temporary note to allow a one-time extension of a 
limited number of TS Surveillance Requirements (SRs). The temporary 
note states that the next required performance of the SRs may be 
delayed until the current cycle refueling outage, but no later than 
February 2, 2005, and it expires upon startup from the refueling 
outage. With the exception of one SR, the period of additional time 
requested occurs during the next planned refueling outage.
    Date of issuance: July 14, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 205.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 2004 (69 
FR 7023). The June 17, 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 amendment is contained 
in a Safety Evaluation dated July 14, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 24, 2002, and its 
supplements dated November 21, 2003, and March 9, 2004.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Section 3.4.11, ``Pressurizer Power Operated Relief 
Valves (PORVs),'' to credit the automatic actuation of the pressurizer 
PORVs for mitigating the plant transient of inadvertent actuation of 
the safety injection (SI) system. The amendments also modify the 
wording in Criteria A, B, and E of TS 3.4.11 to reflect the new 
requirement of ensuring automatic function of PORVs and adds two new 
surveillance requirements. The licensee withdrew the changes to TS 
3.4.10, ``Pressurizer Safety Valves,'' in its letter dated March 9, 
2004.
    Date of issuance: July 2, 2004.
    Effective date: July 2, 2004, and shall be implemented within 30 
days from the date of issuance.

[[Page 46596]]

    Amendment Nos.: Unit 1--171; Unit 2--172.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 24, 2002 (67 
FR 78522)
    The November 21, 2003, and March 9, 2004, supplemental letters 
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 amendments is contained 
in a Safety Evaluation dated July 2, 2004.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: September 19, 2003.
    Brief description of amendment: This amendment revised Surveillance 
Requirement 4.2.4.2 to specifically identify the Power Distribution 
Monitoring System being used in determining the Quadrant Power Tilt 
Ratio with one inoperable Power Range Channel.
    Date of issuance: July 6, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 168.
    Renewed Facility Operating License No. NPF-12: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 2004 (69 FR 
16623).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 6, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendment: March 5, 2004.
    Brief description of amendment: The amendment revises the reactor 
coolant pump flywheel inspection interval from 10 years to 20 years.
    Date of issuance: July 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 293 and 283.
    Facility Operating License No. DPR-77 and DPR-79: Amendment revises 
the technical specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19577).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 8, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 8, 2004.
    Brief description of amendment: The amendment revises TS 5.5.7, 
``Reactor Coolant Pump Flywheel Inspection Program,'' to increase the 
inspection interval from 10 years to 20 years.
    Date of issuance: July 12, 2004.
    Effective date: July 12, 2004, and shall be implemented within 90 
days from the date of issuance.
    Amendment No.: 163.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 11, 2004 (69 FR 
26193).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 12, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 30, 2003, as supplemented by 
letters dated December 18, 2003, and April 13, 2004.
    Brief description of amendment: The amendment revises several 
surveillance requirements (SRs) in Technical Specification (TS) 3.8.1 
on alternating current sources for plant operation. The revised SRs 
have notes deleted or modified to allow the SRs to be performed, or 
partially performed, in reactor modes that previously were not allowed 
by the TSs. The proposed changes to SRs 3.8.4.7 and 3.8.4.8 for direct 
current sources were withdrawn by letter dated April 13, 2004.
    Date of issuance: July 12, 2004.
    Effective date: July 12, 2004, and shall be implemented within 90 
days of the date of issuance including the incorporation of the changes 
to the TS Bases for TS 3.8.1 as described in the licensee's letters 
dated April 30 and December 18, 2003, and April 13, 2004.
    Amendment No.: 154.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34673).
    The December 18, 2003, and April 13, 2004, supplemental letters 
provided additional clarifying information, did not expand the scope of 
the application as 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 July 12, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 26th day of July 2004.

    For the Nuclear Regulatory Commission.
James E. Lyons,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 04-17346 Filed 8-2-04; 8:45 am]
BILLING CODE 7590-01-P