[Federal Register Volume 69, Number 225 (Tuesday, November 23, 2004)]
[Notices]
[Pages 68180-68193]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-25664]



[[Page 68180]]

-----------------------------------------------------------------------

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 October 29, 2004, through November 12, 2004. 
The last biweekly notice was published on November 9, 2004 (69 FR 
64984).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. (Note: 
Public access to ADAMS has been temporarily suspended so that security 
reviews of publicly available documents may be performed and 
potentially sensitive information removed. Please check the NRC Web 
site for updates on the resumption of ADAMS access.) The filing of 
requests for a hearing and petitions for leave to intervene is 
discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ 
reading-rm/doc-collections/cfr/. (Note: Public access to ADAMS has been 
temporarily suspended so that security reviews of publicly available 
documents may be performed and potentially sensitive information 
removed. Please check the NRC Web site for updates on the resumption of 
ADAMS access.) 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

[[Page 68181]]

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, <a href="/cdn-cgi/l/email-protection#e7afa2a6b5aea9a0a3a8a4aca2b3a7a9b5a4c9a0a8b1"><span class="__cf_email__" data-cfemail="94dcd1d5c6dddad3d0dbd7dfd1c0d4dac6d7bad3dbc2">[email protected]</span></a>; 
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 
<a href="/cdn-cgi/l/email-protection#e0afa7a3ad81898ca3858e948592a08e9283ce878f96"><span class="__cf_email__" data-cfemail="94dbd3d7d9f5fdf8d7f1fae0f1e6d4fae6f7baf3fbe2">[email protected]</span></a>. 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. (Note: Public access to ADAMS has been 
temporarily suspended so that security reviews of publicly available 
documents may be performed and potentially sensitive information 
removed. Please check the NRC Web site for updates on the resumption of 
ADAMS access.) 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 
<a href="/cdn-cgi/l/email-protection#f7879385b7998594d9909881"><span class="__cf_email__" data-cfemail="b6c6d2c4f6d8c4d598d1d9c0">[email protected]</span></a>.

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

    Date of amendment request: October 20, 2004.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) Table 4.1-1 functional testing 
surveillance interval from monthly to semi-annually for the following 
reactor protection system instrument channels: Table 4.1-1, Item No. 4, 
``Power Range Channel,'' Item No. 7, ``Reactor Coolant Temperature 
Channel,'' Item No. 8, ``High Reactor Coolant Pressure Channel,'' Item 
No. 9, ``Low Reactor Coolant Pressure Channel,'' Item No. 10,'' Flux-
Reactor Coolant Flow Comparator,'' Item No. 11, ``Reactor Coolant 
Pressure-Temperature Comparator,'' Item No. 12, ``Pump Flux 
Comparator,'' Item No. 13, ``High Reactor Building Pressure Channel,'' 
Item No. 45, ``Loss of Feedwater Reactor Trip,'' and Item No. 46, 
``Turbine Trip/Reactor Trip.'' The TS Section 4.1 Bases would be 
revised to reflect the proposed change from monthly to semi-annually 
and to specify that one channel is being tested every 46 days on a 
continual sequential rotation, which is consistent with the 
calculations of BAW-10167A, Supplement 1, and associated Nuclear 
Regulatory Commission Safety Evaluation Report that indicate that the 
reactor protection system retains a high level of reliability for this 
test interval. The proposed change would also revise TS Table 4.1-1 
functional testing surveillance interval from monthly to quarterly for 
the following reactor protection system reactor trip devices: Table 
4.1-1, Item No. 1, ``Protection Channel Coincidence Logic,'' and Item 
No. 2, ``Control Rod Drive Trip Breaker and Regulating Rod Power 
SCRs.'' The TS Section 4.1 Bases would be revised to reflect the 
proposed change from monthly to quarterly testing and to specify that 
one channel is being tested every 23 days on a continual sequential 
rotation, which is consistent with the calculations of BAW-10167A, 
Supplement 3, February 1998, and the NRC SER for BAW-10167A, Supplement 
3, dated January 7, 1998, that indicate that the reactor trip system 
retains a high level of reliability for this test interval.
    Basis for proposed valuated 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 reactor protection system monitors parameters related to 
safe operation and trips the reactor to protect the reactor core 
against fuel cladding damage. It also assists in protecting against 
reactor coolant system damage caused by high system pressure by 
limiting energy input to the system through reactor trip action. 
Therefore, this change has no impact on the probability of an 
accident previously evaluated. The results of the reliability 
analyses conducted in accordance

[[Page 68182]]

with NRC [Nuclear Regulatory Commission] approved methodology and 
criteria show that the test interval extension of the reactor 
protection system instrument channels and reactor trip devices is 
not a significant contributor to trip system unavailability or the 
risk of core damage. The reactor protection system instrument 
channel and reactor trip device functional test surveillance program 
will continue to ensure that the reactor protection system is 
capable of performing its intended safety function during a design 
basis accident.
    Therefore, this change has no effect on the consequences of an 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change involves the reactor protection system 
instrument channel and reactor trip device surveillance test 
interval, which is not, in and of itself, considered to be an 
accident initiator. Postulated failure of the reactor protection 
system instrument channel or reactor trip device to function is an 
analyzed condition and does not constitute a new or different kind 
of accident. The proposed change does not create any new failure 
modes not bounded by previously analyzed accidents.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The results of the reliability analysis conducted in accordance 
with NRC approved methodology and criteria show that the test 
interval extension of the reactor protection system instrument 
channels and reactor trip devices is not a significant contributor 
to trip system unavailability or the risk of core damage. The 
Technical Specifications will continue to require the reactor 
protection system trip setpoints to remain within the assumptions of 
the accident analysis and that adequate reliability of the reactor 
protection system trip devices is maintained, thus preserving 
existing margins of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: August 20, 2004.
    Description of amendment request: The proposed amendment would 
revise the Allowable Values for the following Reactor Protection System 
(RPS) instrumentation functions: Intermediate Range Neutron Flux, 
Reactor Coolant Flow--Low, Steam Generator Water Level--Low Coincident 
with Steam Flow/Feedwater Flow Mismatch, and Intermediate Range Neutron 
Flux (P-6) Interlock. Additionally, these changes revise the Allowable 
Value for the Engineered Safety Feature Actuation System 
Instrumentation function for High Steam Flow in Two Steam Lines 
Coincident with Steam Line Pressure--Low. Also the proposed amendment 
would delete an unnecessary footnote associated with the applicability 
for the Automatic Trip Logic RPS instrumentation function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposal to revise the Allowable Values for the affected 
reactor protection and engineered safety feature actuation functions 
was developed in accordance with the current setpoint methodology 
for HBRSEP [H. B. Robinson Steam Electric Plant], Unit No. 2, thus 
ensuring that the probability and consequences of previously 
evaluated accidents are not significantly increased. The proposed 
deletion of the unnecessary footnote associated with the Automatic 
Trip Logic reactor protection instrumentation function does not 
change the requirements for operability of this function. Therefore, 
the proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated, 
because the factors that are used to determine the probability and 
consequences of accidents are not being affected.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    The proposed changes will continue to ensure that the 
operability of the previously described functions will be 
appropriately maintained. No physical changes to the HBRSEP, Unit 
No. 2, systems, structures, or components are being implemented. 
There are no new or different accident initiators or sequences being 
created by the proposed Technical Specifications changes. Therefore, 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed changes, as previously described, ensure that the 
margin of safety for the applicable fission product barriers that 
are protected by these functions will continue to be maintained. 
This conclusion is based on the use of a valid setpoint methodology 
for determining the Allowable Values for the reactor protection and 
engineered safety feature actuation functions. Therefore, these 
changes do not involve a significant reduction in the margin of 
safety.
    Based on the preceding discussion, the requested changes do not 
involve a significant hazards consideration.

    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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina; Docket 
Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1and 2, 
Mecklenburg County, North Carolina; Docket Nos. 50-269, 50-270, and 50-
287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: September 28, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation 
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 28, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:


[[Page 68183]]



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

    The proposed change eliminates the TS reporting requirements to 
provide a monthly operating report of shutdown experience and 
operating statistics if the equivalent data is submitted using an 
industry electronic database. It also eliminates the TS reporting 
requirement for an annual occupational radiation exposure report, 
which provides information beyond that specified in NRC regulations. 
The proposed change involves no changes to plant systems or accident 
analyses. As such, the change is administrative in nature and does 
not affect initiators of analyzed events or assumed mitigation of 
accidents or transients. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any 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 previously 
evaluated.

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

    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

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

    Date of amendment request: October 5, 2004.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TSs) 3/4.3.1, ``Reactor Trip System 
Instrumentation,'' and 3/4.3.2, ``Engineered Safety Feature Actuation 
System Instrumentation,'' to modify steam generator (SG) level 
allowable value setpoints. The proposed changes address recent generic 
issues involving new SG level uncertainty considerations and margins 
associated with Westinghouse-designed SGs.
    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 SG water level-low-low setpoint and allowable value have 
been revised to address Westinghouse Nuclear Safety Advisory Letter 
NSAL-03-9 and other considerations on steam generator water level 
uncertainties. The revised setpoint and allowable value calculations 
continues to follow the setpoint methodology previously approved for 
BVPS Unit No. 1 and No. 2 while addressing newly identified level 
uncertainty considerations. The proposed changes to the SG water 
level-low-low Allowable Value for BVPS Unit No. 1 and No. 2 and to 
the SG water level-high-high Allowable Value for BVPS Unit No. 2 
continue [to] maintain the validity of the safety analysis limits 
used in the safety analyses that credit the actuations based on SG 
water level.
    The proposed changes do not alter the causes for any accident 
described in the Updated Final Safety Analysis Report (UFSAR) that 
credit the SG water level setpoint actuations. Therefore, they do 
not involve a significant increase in the probability of an accident 
previously evaluated.
    The proposed changes do not alter the accident analyses that 
credit the SG water level-low-low setpoint actuation or the 
associated accident acceptance criteria. Therefore, they do not 
involve a significant increase in the 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 SG water level-low-low setpoint and allowable value have 
been revised to address Westinghouse Nuclear Safety Advisory letter 
NSAL-03-9 and other considerations on steam generator water level 
uncertainties. Implementation of the proposed setpoint changes have 
no significant effect on either the configuration of the plant, or 
the manner in which the plant is operated. The proposed changes to 
the SG water level-low-low allowable value for BVPS Unit No. 1 and 
No. 2 and to the SG water level-high-high allowable value for BVPS 
Unit No. 2 continue to maintain the validity of the safety analysis 
limits used in the safety analyses that credit the actuations based 
on SG water level.
    Therefore, since the plant configuration is not adversely 
changed and the proposed changes do not alter the accident analyses 
that credit actuation based on SG water level, the proposed change 
does not create the possibility of a new or different [kind of] 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The Reactor Trip System and Engineered Safety Feature 
Actuation System setpoint analysis methodology and acceptance 
criteria provide the margin of safety. The SG water level-low-low 
and SG water level-high-high actuation setpoint and allowable value 
have been calculated using the same methodology as previously 
approved for the BVPS Unit No. 1 and No. 2 while addressing newly 
identified considerations needed to protect the limits used in the 
safety analyses. The applicable safety analyses have been performed 
and show acceptable results. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: October 25, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 2.1.1.2 for the dual recirculation loop 
and single recirculation loop Safety Limit Minimum Critical Power Ratio 
(SLMCPR) values to reflect results of a cycle specific calculation.
    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 probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. 
Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant

[[Page 68184]]

operation. Therefore, no individual precursors of an accident are 
affected.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established, consistent with NRC 
approved methods, to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
change conservatively establishes the safety limit for the minimum 
critical power ratio (SLMCPR) for Cooper Nuclear Station Cycle 23 
such that the fuel is protected during normal operation and during 
any plant transients or anticipated operational occurrences.
    The proposed change revises the SLMCPR to protect the fuel 
during normal operation as well as during any transients or 
anticipated operational occurrences. Operational limits Minimum 
Critical Power Ratio (MCPR) are established based on the proposed 
SLMCPR to ensure that the SLMCPR is not violated during all modes of 
operation. This will ensure that the fuel design safety criteria 
(i.e., that at least 99.9% of the fuel rods do not experience 
transition boiling during normal operation and anticipated 
operational occurrences) is met. Since the operability of plant 
systems designed to mitigate any consequences of accidents has not 
changed, the consequences of an accident previously evaluated are 
not expected to increase.
    Based on the above NPPD [Nebraska Public Power District] 
concludes that the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration or changes in allowable 
modes of operation. The proposed change does not involve any 
modifications of the plant configuration or allowable modes of 
operation. The proposed change to the SLMCPR assures that safety 
criteria are maintained for Cycle 23.
    Based on the above NPPD concludes that the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The value of the proposed SLMCPR provides a margin of safety by 
ensuring that no more than 0.1% of the rods are expected to be in 
boiling transition if the MCPR limit is not violated. The proposed 
change will ensure the appropriate level of fuel protection is 
maintained. Additionally, operational limits are established based 
on the proposed SLMCPR to ensure that the SLMCPR is not violated 
during all modes of operation. This will ensure that the fuel design 
safety criteria (i.e., that at least 99.9% of the fuel rods do not 
experience transition boiling during normal operation as well as 
anticipated operational occurrences) are met.
    Based on the above NPPD concludes that 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: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Acting Section Chief: Michael K. Webb.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: April 15, 2004.
    Description of amendment request: The proposed change would modify 
the Salem Updated Final Safety Analysis Report (UFSAR) with respect to 
fire protection requirements for the 4160 Volt Switchgear Rooms, 460 
Volt Switchgear Rooms, and the Lower Electrical Penetration Area Rooms. 
Specifically, the amendment would reduce the UFSAR description of the 
Carbon Dioxide Tank volume from being able to provide two full 
discharges to an affected room to one full and one partial discharge to 
an affected room. Additionally, the assumed ability of the Carbon 
Dioxide system would be reduced from an ability to produce a 
CO<INF>2</INF> concentration of 50% for 30 minutes to an ability to 
produce a CO<INF>2</INF> concentration of 27.6% for a length of time 
sufficient to suppress a fire and allow the PSEG Nuclear Fire 
Department to respond.
    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 likelihood of a fire event is not increased since the 
proposed change does not alter the fire hazards contained in the 
plant. The ability to achieve and maintain safe shutdown in the 
event of a fire is not impacted by the reduction of CO<INF>2</INF> 
concentration, since the Fire Brigade will respond in ample time and 
extinguish a fire using alternate means. In addition, the proposed 
changes to the UFSAR would not change any response to a fire event. 
Also, the probability of occurrence or the consequences for an 
accident or malfunction of equipment is not increased by the 
proposed changes since the response to a fire event would not change 
and the fire brigade would continue to respond rapidly to any fires 
or fire alarms. Further, the proposed changes do not alter the way 
any structure, system, or component (SSC) functions, do not modify 
the manner in which the plant is operated, and do not significantly 
alter equipment out-of-service time. Changing the CO<INF>2</INF> 
concentration requirement in the 4160 Volt Switchgear Rooms, 460 
Volt Switchgear Rooms and Lower Penetration Area Rooms at Salem 
Units 1 and 2 does not change the probability or consequences of any 
accident and dose consequences are unaffected. No changes to the 
design of structures, systems, or components (SSC) are made and 
there are no effects on accident mitigation.
    Therefore, the proposed changes do not involve 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?
    Response: No.
    The possibility of a new or different kind of accident from any 
accident or malfunction in the Salem Updated Final Safety Analysis 
Report (UFSAR) is not created. The design basis event applicable to 
the proposed change is a fire in the 4160 Volt Switchgear Rooms, 460 
Volt Switchgear Rooms and Lower Penetration Area Rooms at Salem 
Units 1 and 2. Therefore a different accident is not created. In 
addition, the proposed changes cannot initiate an accident. Further, 
the proposed changes to the UFSAR do not change the design function 
or operation of any SSCs.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The reduction in CO<INF>2</INF> concentration provides ample 
response time for the onsite dedicated fire brigade to respond to a 
fire event and a 20% safety factor in CO<INF>2</INF> concentration 
remains. The proposed changes do not affect the ability to safely 
shutdown and maintain the shutdown conditions of either unit 
following a fire in the affected areas. The proposed changes do not 
rely on compensatory measures or actions deviating from the 
licensing or design basis. In addition, the proposed changes do not 
change the margin of safety since no SSCs are changed. The results 
of accident analysis remain unchanged by the proposed changes to the 
UFSAR.
    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

[[Page 68185]]

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.

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

    Date of application request: September 17, 2004.
    Description of amendment request: The amendment is to support the 
replacement of the steam generators (SGs) at Callaway during the 
refueling outage in the Fall of 2005. The amendment would (1) change 
the affected technical specifications (TSs) such as the reactor core 
safety limits (TS 2.1.1), reactor trip system (RTS) and engineered 
safety feature actuation system (ESFAS) instrumentation (TSs 3.3.1 and 
3.3.2), reactor coolant system (RCS) limits (TS 3.4.1), RCS loops (TSs 
3.4.5, 3.4.6, and 3.4.7), RCS operational leakage (TS 3.4.13), SG tube 
integrity (new TS 3.4.17), main steam safety valves (TS 3.7.1), SG 
surveillance program (TS 5.5.9), containment integrated leakage rate 
testing (ILRT) program (TS 5.5.16), and SG inspection report (TS 
5.6.10); (2) revise the affected transient analyses such as excessive 
increase in secondary steam flow event, loss of normal feedwater event, 
transient mass and energy releases, radiological consequences of 
associated events, and containment pressure/temperature responses; and 
(3) revise nuclear steam and supply system (NSSS) design parameters and 
transients, and fatigue usage factors and stresses for the replacement 
SGs. The amendment involves the following areas of change to the 
license: nuclear steam supply system evaluations for the replacement 
steam generators, trip time delay (TTD) elimination for certain RTS and 
ESFAS functions, the SG surveillance program in Technical Specification 
Task Force (TSTF) No. 449 (TSTF-449), and the post-modification 
containment ILRT exception.
    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 for the above areas of review, which is presented below 
(with the terms defined in the plant Technical Specifications 
capitalized):

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Nuclear Steam Supply System Evaluations for Replacement Steam 
Generators

    As discussed in the NSSS Licensing Report (Appendix A to this 
amendment application), all acceptance criteria continue to be met. 
All major NSSS components (e.g., Reactor Vessel, Pressurizer, RCPs 
[(reactor coolant pumps)], Steam Generators, etc.) have been 
assessed with respect to bounding conditions expected for 
replacement steam generator (RSG) conditions. In all cases operation 
has been found to be acceptable. Major systems and subsystems (e.g., 
safety injection, RHR [residual heat removal], etc.) have been 
reviewed and acceptable performance has been verified for their 
normal operation and, as applicable, for their safety-related 
functions. All reactor trip and ESFAS actuation setpoints have been 
assessed, and the proposed setpoint modifications will assure 
adequate protection is afforded for all design basis events.
    The reactor core safety limits have been revised based on the 
RSG project parameters. All of the acceptance criteria for the 
accident analyses (e.g., DNBR [departure from nucleate boiling 
ratio] limits, fuel centerline temperatures, etc.) continue to be 
met with the revised safety limit lines. Therefore, the revised core 
safety limit line changes are acceptable. The proposed changes to 
the reactor core safety limits will not initiate any accidents; 
therefore, they do not increase the probability of an accident 
previously evaluated in the FSAR [Callaway Final Safety Analysis 
Report]. The comprehensive analytical efforts performed to support 
the proposed RSG conditions include a reanalysis or evaluation of 
all accident analyses that are impacted by the revised reactor core 
safety limits.
    The changes in various SG-related RTS and ESFAS Allowable Values 
have resulted from the analyses performed to support plant operation 
at the proposed RSG conditions. Setpoint uncertainty calculations 
confirm the acceptability of these revised Allowable Values. The 
affected RTS and ESFAS Allowable Values have been modified to 
reflect the results of updated setpoint calculations based on plant-
specific uncertainties, calibration practices, calibration 
equipment, and installed hardware and procedures. The Allowable 
Values were calculated using the same Westinghouse setpoint 
methodology used for the current trip setpoints, but improved in a 
conservative fashion to include refinements that better reflect 
plant calibration practices and equipment performance. These 
refinements include the incorporation of a sensor reference accuracy 
term to address repeatability effects when performing a single pass 
calibration (i.e., one up and one down pass at several points 
verifies linearity and hysteresis, but not repeatability). In 
addition, sensor and rack error terms for calibration accuracy and 
drift are grouped in the Channel Statistical Allowance equation with 
their dependent measurement and test equipment (M&TE) terms, then 
combined with the other independent error terms using the square 
root sum of the squares (SRSS) methodology. This improved setpoint 
methodology has been previously review[ed] and approved by the NRC. 
The proposed RTS and ESFAS Allowable Value changes will not initiate 
any accidents; therefore, they do not increase the probability of an 
accident previously evaluated in the FSAR. The comprehensive 
analytical effort performed to support the proposed RSG conditions 
included a reanalysis or evaluation of all accident analyses that 
are impacted by the revised RTS and ESFAS Allowable Values. All 
systems will function as designed.
    The decrease in the Maximum Allowable Power for 3 OPERABLE MSSVs 
[main steam safety valves] per SG from < 49% of Rated Thermal Power 
to < 45% of Rated Thermal Power resulted from the analyses and 
evaluations performed to support plant operation at the proposed RSG 
conditions. The accident analysis acceptance criteria continue to be 
met with these changes. These proposed plant system changes do not 
increase the probability of an accident previously evaluated in the 
FSAR. The comprehensive analytical effort performed to support the 
proposed RSG conditions has included a review and evaluation of all 
components and systems (including interface systems and control 
systems) that could be affected by this change. All systems will 
function as designed. The change in the manner in which the Reactor 
Coolant Flow--Low Allowable Value is defined (while retaining the 
same numerical value), the change in the manner in which RCS average 
temperature is defined and the reduced upper limit for nominal T-avg 
[average temperature] at full power conditions in the 
Overtemperature [Delta]T [delta temperature] and Overpower [Delta]T 
setpoint equations, and the changes to the pressurizer pressure and 
RCS average temperature limits in the DNB LCO [departure from 
nucleate boiling limiting condition for operation] [TS] 3.4.1 have 
also been evaluated. None of these proposed changes will initiate 
any accidents; therefore, the probability of an accident has not 
been increased.
    The potential dose consequences have been analyzed with respect 
to the above changes collectively. The dose increases are less than 
minimal (i.e., <10% of the margin between the regulatory limits and 
the currently reported doses). The applicable dose acceptance 
criteria continue to be met.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Trip Time Delay Elimination

    This design change will eliminate only the Trip Time Delay 
portion of the SG Water Level Low-Low trip functions and return that 
portion of the design to condition that existed prior to Callaway 
Amendment 43 dated April 14, 1989. The coincidence logic in the 
Solid State Protection System will be unaffected. In all other 
regards, the design of the RTS and ESFAS instrumentation will be 
unaffected. These protection systems will continue to function in a 
manner consistent with the plant design basis. All design, material, 
and construction standards that were applicable prior to this 
amendment request are maintained.

[[Page 68186]]

    The probability and consequences of accidents previously 
evaluated in the FSAR are not adversely affected because the removal 
of the trip time delay circuitry assures a faster response by the 
affected trip functions, consistent with the safety analysis 
acceptance criteria and the original plant licensing basis.
    The proposed change will not affect the probability of any event 
initiators. There will be no degradation in the performance of, or 
an increase in the number of challenges imposed on, safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance.
    The proposed change will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in 
the FSAR.
    Therefore, the proposed TTD elimination does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

TSTF-449 Generic Licensing Change Package

    This 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.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that are analyzed as part of a plant's licensing 
basis. In the analysis cases for the SGTR event at Callaway Plant, a 
primary to secondary LEAKAGE rate of 1 gallon per minute (gpm) to 
the unaffected SGs is assumed, in excess of the RCS Operational 
LEAKAGE rate limit in TS 3.4.13, and the LEAKAGE rate associated 
with a double-ended rupture of a single tube in the ruptured SG is 
also assumed. For other design basis accidents such as main steam 
line break (MSLB), rod ejection, and reactor coolant pump locked 
rotor, the SG tubes are assumed to retain their structural integrity 
(i.e., they are assumed not to rupture). These additional analyses 
for Callaway Plant assume, as an initial condition, that primary to 
secondary LEAKAGE for all SGs is 1 gpm. The accident induced leakage 
criterion introduced by the proposed change to TS 5.5.9 accounts for 
tubes that may leak during design basis accidents. The accident 
induced leakage criterion limits this leakage to no more than the 1 
gpm value assumed in the accident analyses.
    The SG performance criteria added to TS 5.5.9 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 TS 
5.5.9. The program, defined by NEI [Nuclear Energy Institute] 97-06, 
Steam Generator Program Guidelines, includes a framework that 
incorporates a balance of prevention, inspection, evaluation, 
repair, 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 TS 3.4.13 for RCS Operational 
leakage and in TS 3.4.16 for DOSE EQUIVALENT I-131 in the primary 
coolant to ensure the plant is operated within its analyzed 
condition. The radiological consequence analyses at Callaway Plant 
assume that the primary to secondary LEAKAGE rate is 1 gpm (more 
conservative than the limit in TS 3.4.13), and that the reactor 
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
3.4.16 limits.
    The proposed TSTF-449 changes reflect the design of the 
replacement SGs, but do not affect their method of operation or 
primary or secondary coolant chemistry controls. The proposed 
changes update the TS and enhance the requirements for SG 
inspections. The proposed changes do not adversely impact the 
conclusions of any previously evaluated design basis accident and 
are an improvement over the existing TS.
    Therefore, this proposed change to implement TSTF-449 does not 
affect the consequences of a SGTR accident and the probability of 
such an accident is reduced. In addition, this proposed change does 
not affect the consequences of an MSLB, rod ejection, reactor 
coolant pump locked rotor, or any other accident event involving the 
potential release of radioactive fluids from the secondary side of 
Callaway Plant. [Therefore, this proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.]

Post-Modification ILRT Exception

    This proposed change would provide Callaway Plant with an 
exception from performing a post-modification containment integrated 
leak rate test following the replacement of the steam generators 
during Refuel [Outage] 14.
    Integrated leak rate tests are performed to assure the leak-
tightness of the primary containment boundary system, and as such 
they are not accident initiators. Therefore, not performing an 
integrated leak rate test will not affect the probability of an 
accident previously evaluated. The intent of post-modification 
integrated leak rate testing requirements is to assure the leak-
tight integrity of the area affected by the modification. For the 
Callaway Plant steam generator replacement modification, this intent 
will be satisfied by performing the American Society of Mechanical 
Engineers code required inspections and tests. Since the leak-
tightness integrity of the primary containment boundary affected by 
the steam generator replacement will be assured, there is no change 
in the containment boundary's ability to confine radioactive 
materials during an accident. Therefore, adding a Technical 
Specification exception from the steam generator replacement post-
modification integrated leak rate testing requirements does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

Nuclear Steam Supply System Evaluations for Replacement Steam 
Generators

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    This amendment does not alter the safe performance of the plant 
protection systems to trip the reactor when necessary or actuate ESF 
[engineered safety feature] systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.

Trip Time Delay Elimination

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    This amendment does not alter the safe performance of the plant 
protection systems to trip the reactor when necessary or actuate ESF 
systems.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

TSTF-449 Generic Licensing Change Package

    The proposed performance based requirements are an improvement 
over the requirements imposed by the existing 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.
    This proposed change does not impact the method of SG operation 
or primary or secondary coolant chemistry controls. In addition, 
this proposed change does not impact any other plant system or 
component. The change enhances SG inspection requirements.
    Therefore, this proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

Post-Modification ILRT Exception

    The proposed change would provide Callaway Plant with an 
exception from performing a post-modification containment

[[Page 68187]]

integrated leak rate test following the replacement of the steam 
generators during Refuel 14. Providing an exception from performing 
a test does not involve a physical change to the plant nor does it 
change the operation of the plant. Thus it cannot introduce a new 
failure mode. Therefore adding a Technical Specification requirement 
that provides an exception from the steam generator replacement 
post-modification integrated leak rate testing requirement does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.

Nuclear Steam Supply System Evaluations for Replacement Steam 
Generators

    The analyses and evaluations supporting the proposed RSG 
conditions reflect the reactor core safety limits. All acceptance 
criteria continue to be met.
    The analyses supporting the proposed RSG conditions reflect the 
proposed RTS and ESFAS Allowable Values. Setpoint calculations 
demonstrate that margin exists between these Allowable Values and 
the corresponding safety analysis limits used in the RSG analyses. 
The calculations are based on plant instrumentation and calibration/
functional test methods and include allowances for the RSG 
conditions. All analyses and evaluations supporting the proposed RSG 
core safety limits, decrease in maximum allowable power level for 3 
operable MSSVs per SG, the change in the manner in which the Reactor 
Coolant Flow--Low Allowable Value is defined (while retaining the 
same numerical value), the change in the manner in which RCS average 
temperature is defined and the reduced upper limit for nominal T-avg 
at full power conditions in the Overtemperature [Delta]T and 
Overpower [Delta]T setpoint equations, and the changes to the 
pressurizer pressure and RCS average temperature limits in the DNB 
LCO [TS] 3.4.1 are acceptable. All acceptance criteria continue to 
be met. Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

Trip Time Delay Elimination

    This proposed change does not eliminate any RTS or ESFAS 
surveillances or alter the frequency of those surveillances as 
required by the TS. The SG Water Level Low--Low safety analysis 
limit of 0% span assumed in the analyses supporting the approval of 
the TTD design in Callaway Amendment 43 dated April 14, 1989 is also 
used in the RSG analyses discussed above. None of the acceptance 
criteria for any accident analysis is changed for TTD elimination.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. The radiological dose consequence 
acceptance criteria will continue to be met.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

TSTF-449 Generic Licensing Change Package

    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. This proposed 
change to implement TSTF-449 does not, of itself, 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, repair 
(only under NRC-approved methods, none of which currently apply to 
the RSGs), 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 existing TS.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by this proposed change.

Post-Modification ILRT Exception

    The proposed change would provide Callaway Plant with an 
exception from performing a post-modification containment integrated 
leak rate test following the replacement of the steam generators 
during Refuel 14. The intent of post-modification integrated leak 
rate testing requirements is to assure the leak-tight integrity of 
the area affected by the modification. This intent will be satisfied 
by performing American Society of Mechanical Engineers code required 
inspections and tests. The acceptance criterion for American Society 
of Mechanical Engineers code system pressure testing for the base 
metal and welds is no leakage. In addition, the test pressure for 
the system pressure test will be several times that required during 
an integrated leak rate test. Since the leak-tight integrity of the 
primary containment boundary affected by the steam generator 
replacement will be assured, there is no change in the primary 
containment boundary's ability to confine radioactive materials 
during an accident. Therefore, adding a Technical Specification 
requirement that provides an exception from the steam generator 
replacement post-modification integrated leak rate testing 
requirements 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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert A. Gramm.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: September 15, 2004.
    Description of amendment request: The proposed changes will change 
the Administrative Controls Section of the Technical Specifications 
(TS) in order to incorporate title changes, change the location where 
the plant-specific titles and TS titles are correlated, and relocate 
the unit staff requirements to the Quality Assurance Program. These 
proposed changes will support the implementation of proposed Virginia 
Electric and Power Company Topical Report DOM-QA-1, ``Nuclear Facility 
Quality Assurance Program Description,'' currently under U.S. Nuclear 
Regulatory Commission (NRC) staff review.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of North Anna Units 1 and 2 in accordance with the 
proposed license amendments would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change is administrative in nature and does not 
affect plant systems, structures or components (SSCs) or plant 
operation during normal or accident conditions. The proposed change 
only affects the designated titles of personnel, the location of the 
TS title and plant-specific title correlation, and the location of 
the unit staff qualification requirements. Therefore, this change 
has no bearing on the probability of an accident. Management 
organizational structure and safety and operational reviews have not 
changed and there is no change in the method of plant operation, 
operation review, or system design review. As such, this change does 
not alter the conclusions of the existing safety analyses and 
therefore does not alter the consequences of an accident previously 
evaluated.
    2. Operation in accordance with the proposed license amendments 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed administrative change continues to ensure that 
adequate management oversight exists at the plant in accordance with 
the existing Technical

[[Page 68188]]

Specifications. The proposed change only affects the designated 
titles of personnel, the location of the TS title and plant-specific 
title correlation, and the location of the unit staff qualification 
requirements. This change does not impact plant SSCs or plant 
operation. Management organizational structure and safety and 
operational reviews have not changed and there is no change in the 
method of plant operation, operation review, or system design 
review. There are no new or different accident scenarios, accident 
initiators, nor failure mechanisms that will be introduced due to 
this change. Therefore, the proposed change does not create the 
possibility of an accident of a different type than evaluated 
previously.
    3. Operation in accordance with the proposed license amendments 
would not involve a significant reduction in a margin of safety.
    The proposed change only affects the designated titles of 
personnel, the location of the TS title and plant-specific title 
correlation, and the location of the unit staff qualification 
requirements. This change does not impact plant design, plant 
operation or any safety margin. Therefore, the proposed change does 
not significantly reduce a margin of safety.
    This evaluation concludes that the proposed amendments to the 
North Anna Units 1 and 2 Technical Specifications do not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident, do not create the possibility of a 
new or different kind of accident and 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: Mary Jane Ross-Lee (Acting).

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

    Date of amendment request: October 7, 2004.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) 5.3, ``Unit Staff Qualifications,'' to 
reinstate the qualification requirements for the shift manager and 
control room supervisor positions that were inadvertently eliminated 
through Amendment No. 150. Also, TS 5.3 would be revised to reference 
this amendment application for the use of the National Academy for 
Nuclear Training guideline, ACAD 00-003, Revision 1, ``Guidelines for 
Initial Training and Qualification of Licensed Operators.'' Various 
other TSs would be revised to make corrections that were identified by 
the NRC staff in its letter dated January 28, 2004, and additional 
reviews performed by the licensee.
    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.

Unit Staff Qualifications

    The proposed change is an administrative change to reinstate the 
qualification requirements for specific control room positions that 
were inadvertently eliminated through the issuance of Amendment No. 
150 and utilize Revision 1 to ACAD 00-003, ``Guidelines for Initial 
Training and Qualification of Licensed Operators.'' The proposed 
change does not directly impact accidents previously evaluated. 
WCNOC's [Wolf Creek Nuclear Operating Corporation's] licensed 
operator training program is accredited by the National Academy for 
Nuclear Training and is based on a systems approach to training 
consistent with the requirements of 10 CFR 55. Although licensed 
operator qualifications and training may have an indirect impact on 
accidents previously evaluated, the NRC considered this impact 
during the rulemaking process, and by promulgation of the revised 10 
CFR 55 rule, concluded that this impact remains acceptable as long 
as the licensed operator training program is certified to be 
accredited and is based on a systems approach to training.

Corrections

    The proposed change involves corrections to the Technical 
Specifications that are either associated with the issuance of the 
Improved Technical Specifications (Amendment No. 123) or subsequent 
amendments. The changes are considered administrative changes and do 
not modify, add, delete, or relocate any technical requirements of 
the Technical Specifications. As such, administrative changes do not 
effect initiators of analyzed events or assumed mitigation of 
accident or transient events.
    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.

Unit Staff Qualifications

    The proposed change is an administrative change to reinstate the 
current requirements of specific control room positions and allow 
the use of Revision 1 of ACAD 00-003 for initial training and 
qualification of licensed operators. WCNOC's licensed operator 
training program is accredited by the National Academy for Nuclear 
Training and is based on a systems approach to training consistent 
with the requirements of 10 CFR 55. Although licensed operator 
qualifications and training may have an indirect impact on accidents 
previously evaluated, the NRC considered this impact during the 
rulemaking process, and by promulgation of the revised 10 CFR 55 
rule, concluded that this impact remains acceptable as long as the 
licensed operator training program is certified to be accredited and 
is based on a systems approach to training.

Corrections

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods of governing normal plant operation. The 
proposed change will not impose any new or eliminate any old 
requirements.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.

Unit Staff Qualifications

    The proposed change is an administrative change to reinstate the 
current requirements of specific control room positions and allow 
the use of Revision 1 of ACAD 00-003 for initial training and 
qualification of licensed operators. As noted previously, WCNOC's 
licensed operator training program is accredited and is based on a 
systems approach to training consistent with the requirements of 10 
CFR 55. Licensed operator qualifications and training can have an 
indirect impact on the margin of safety. However, the NRC considered 
this impact during the rulemaking process, and by promulgation of 
the revised 10 CFR 55 rule, determined that this impact remains 
acceptable when licensees maintain a licensed operator training 
program that is accredited and based on a systems approach to 
training.

Corrections

    The proposed change will not reduce a margin of safety because 
they have no effect on any safety analysis assumptions. The change 
is administrative in nature.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.

[[Page 68189]]

    NRC Section Chief: Robert Gramm.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: October 22, 2004.
    Description of amendment request: The proposed amendment would 
revise the allowed outage times of Technical Specification 3.3.3.6, 
``Accident Monitoring Instrumentation,'' to be consistent with the 
completion times in the related specification in NUREG-1431, Revision 
3, ``Standard Technical Specifications Westinghouse Plants.''
    Date of publication of individual notice in Federal Register: 
November 2, 2004 (69 FR 63560).
    Expiration date of individual notice: December 2, 2004 (public 
comments) and January 3, 2005 (hearing requests).

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to <a href="/cdn-cgi/l/email-protection#7c0c180e3c120e1f521b130a"><span class="__cf_email__" data-cfemail="05756177456b77662b626a73">[email protected]</span></a>. (Note: Public access 
to ADAMS has been temporarily suspended so that security reviews of 
publicly available documents may be performed and potentially sensitive 
information removed. Please check the NRC Web site for updates on the 
resumption of ADAMS access.)

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

    Date of application for amendment: December 23, 2003, as 
supplemented by letter dated June 16, 2004.
    Brief description of amendment: The amendment revised Section 4.5.D 
of the Technical Specifications to specify testing the main steam 
isolation valves at a pressure lower than Pa, the calculated peak 
containment internal pressure related to the design-basis loss-of-
coolant accident.
    Date of Issuance: November 2, 2004.
    Effective date: November 2, 2004 and shall be implemented within 30 
days of issuance
    Amendment No.: 250.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: February 17, 2004 (69 
FR 7518).
    The June 16, 2004, letter provided clarifying information within 
the scope of the original application and did not change the staff's 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of this amendment is contained in a 
Safety Evaluation dated November 2, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: January 15, 2004, as 
supplemented by letter dated March 15, 2004.
    Brief description of amendments: The amendments revise the 
Technical Specifications associated with the control rod drive trip 
devices. The amendments are needed to support implementation of the 
reactor trip breaker replacement.
    Date of Issuance: November 2, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 341, 343, 342.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19566). The supplement dated March 15, 2004, provided clarifying 
information that did not change the scope of the January 15, 2004, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 22, 2004, as supplemented 
July 23 and October 11, 2004.
    Brief description of amendments: The amendments modified Technical 
Specification (TS) requirements to adopt the provisions of Industry/TS 
Task

[[Page 68190]]

Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode 
Restraints.''
    Date of issuance: November 4, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 263 and 144.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53108). The supplemental letters dated July 23 and October 11, 2004, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the NRC staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 4, 2004.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410, 
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New 
York

    Date of application for amendments: January 8, 2004 (2 letters), as 
supplemented by letter dated June 17, 2004.
    Brief description of amendments: The amendments approve 
implementation of the Boiling Water Reactor Vessel and Internals 
Project Reactor Pressure Vessel Integrated Surveillance Program as the 
basis for demonstrating the units' compliance with the requirements of 
appendix H to Title 10 of the Code of Federal Regulations. 
Specifically, the amendments approved the wording proposed by the 
licensee to update the units' Updated Safety Analysis Reports. In 
addition, the Unit 1 amendment also revised the Technical 
Specifications to delete any reference to plant-specific surveillance 
requirements.
    Date of issuance: November 8, 2004.
    Effective date: As of the date of issuance. Integrated Surveillance 
Program shall be implemented within 90 days of issuance. The units' 
Final Safety Analysis Report (Updated) shall be updated in accordance 
with 10 CFR 50.71(e).
    Amendment Nos.: 184 and 114.
    Facility Operating License Nos. DPR-63 and NPF-69: Amendments 
revise the Technical Specifications (for Unit 1), the operating license 
(for Unit 2), and approve revision of licensing basis for both units.
    Date of initial notice in Federal Register: February 17, 2004 (69 
FR 7524). The June 17, 2004, letter provided clarifying information 
within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in two Safety Evaluations, both dated November 8, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 23, 2003, as 
supplemented June 21, 2004.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Limiting Condition for Operation (LCO) Tables 3.2.1 
and 3.2.4 to (1) eliminate the reactor head cooling containment 
isolation function from the TSs, (2) correct and clarify the 
description of the number of instrument channels per trip system as 
defined in the TSs, and (3) revise an existing LCO for radiation 
monitors used to isolate reactor building ventilation and initiate the 
standby gas treatment system.
    Date of issuance: November 2, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 140.
    Facility Operating License No. DPR-22. Amendment revised the TSs.
    Date of initial notice in Federal Register: March 30, 2004 (69 FR 
16621).
    The supplemental letter 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 November 2, 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: March 18, 2004, and its 
supplements dated August 18 and 20, and September 17, 2004.
    Brief description of amendments: The amendments authorize revisions 
to the Final Safety Analysis Report (FSAR) Update to incorporate the 
NRC approval of a permanently revised steam generator voltage-based 
repair criteria probability of detection (POD) method. The revised POD 
method is referred to as the probability of prior cycle detection 
method. In addition, a reporting requirement is added to the DCPP 
Technical Specifications as TS 5.6.10.i.
    Date of issuance: October 28, 2004.
    Effective date: October 28, 2004, and shall be implemented within 
30 days of the date of issuance. The implementation of the amendment 
includes the incorporation into the FSAR Update the changes discussed 
above, as described in the licensee's application dated March 18, 2004, 
and its supplements dated August 18 and 20, and September 17, 2004, and 
evaluated in the staff's Safety Evaluation attached to the amendments.
    Amendment Nos.: Unit 1-177; Unit 2-179.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the FSAR Update and the Technical Specifications.
    Date of initial notice in Federal Register: June 22, 2004 (69 FR 
34704).
    The August 18 and 20, and September 17, 2004, supplemental letters 
provided additional clarifying information, 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 October 28, 2004.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of application for amendments: December 22, 2003, as 
supplemented by letters dated June 18, July 15, and September 8, 2004.
    Brief description of amendments: The amendment added TS 3.3.1.3, 
``Oscillation Power Range Monitor (OPRM) Instrumentation,'' and changed 
TS 3.4.1, ``Recirculation Loops Operating,'' and TS 5.6.5, ``Core 
Operating Limits Report,'' to remove specifications and information 
related to current stability specifications which will no longer be 
needed with the operation of the OPRM system.
    Date of issuance: November 9, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 217 and 192.

[[Page 68191]]

    Facility Operating License Nos. NPF-14 and NPF-22: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2745). The supplements dated June 18, July 15, and September 8, 2004, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 9, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 31, 2003, as supplemented 
by letter dated July 30, 2004.
    Brief description of amendment: The amendment revised the reactor 
pressure vessel pressure-temperature limits and extends their validity 
to 32 effective full power years.
    Date of issuance: November 1, 2004.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 157.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 2004 (69 FR 
32076). The July 30, 2004 letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 1, 2004.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Access and 
Management System's (ADAMS) Public Electronic Reading Room on the 
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. 
If you do not have access to ADAMS or if there are problems in 
accessing the documents located in ADAMS, contact the NRC Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#e8988c9aa8869a8bc68f879e"><span class="__cf_email__" data-cfemail="49392d3b09273b2a672e263f">[email protected]</span></a>. (Note: Public access to ADAMS has been 
temporarily suspended so that security reviews of publicly available 
documents may be performed and potentially sensitive information 
removed. Please check the NRC Web site for updates on the resumption of 
ADAMS access.)
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to

[[Page 68192]]

intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.309, which is 
available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1-800-397-4209, 301-415-4737, or by e-mail to <a href="/cdn-cgi/l/email-protection#d5a5b1a795bba7b6fbb2baa3"><span class="__cf_email__" data-cfemail="bbcbdfc9fbd5c9d895dcd4cd">[email protected]</span></a>. (Note: 
Public access to ADAMS has been temporarily suspended so that security 
reviews of publicly available documents may be performed and 
potentially sensitive information removed. Please check the NRC Web 
site for updates on the resumption of ADAMS access.) If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or a presiding officer designated by the Commission or 
by the Chief Administrative Judge of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the Chief Administrative Judge of the Atomic Safety and 
Licensing Board will issue a notice of a hearing or an appropriate 
order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, <a href="/cdn-cgi/l/email-protection" class="__cf_email__" data-cfemail="630b0602110a0d04070c00080617230d11004d040c15">[email protected]</a>; 
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 
<a href="/cdn-cgi/l/email-protection#cc838b8f81ada5a08fa9a2b8a9be8ca2beafe2aba3ba"><span class="__cf_email__" data-cfemail="fcb3bbbfb19d9590bf999288998ebc928e9fd29b938a">[email protected]</span></a>. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer or the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, 
Unit 2, Matagorda County, Texas

    Date of amendment request: September 30, 2004.
    Description of amendment request: The amendment changes Technical 
Specification 4.4.4.2 to expand the range of conditions under which 
quarterly testing of block valves for the pressurizer power operated 
relief valves would be unnecessary.
    Date of issuance: October 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 153.

[[Page 68193]]

    Facility Operating License No. NPF-80: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. October 6, 2004 (69 FR 59969). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided an opportunity to request a hearing by December 6, 2004, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated October 21, 2004.
    Attorney for licensee: Mr. John E. Matthews, Morgan, Lewis & 
Bokius, LLP, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Michael K. Webb, Acting.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: November 2, 2004.
    Description of amendment request: The amendment revises Technical 
Specification Limiting Condition for Operation 3.4.3, ``Primary Coolant 
System (PCS) Pressure and Temperature (P/T) Limits'' to add 
restrictions to the cooldown rate limits. This amendment supports plant 
restart following repairs of two reactor vessel closure head control 
rod drive nozzle penetrations at the Palisades Nuclear Power Plant.
    Date of issuance: November 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
immediately.
    Amendment No.: 218.
    Facility Operating License No. DPR-20: Amendment revises the 
Technical Specification.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC):
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated November 8, 
2004.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: November 4, 2004.
    Description of amendment request: The proposed amendment extended 
the implementation period for License Amendment 294 to May 15, 2005.
    Date of issuance: November 9, 2004.
    Effective date: As of date of issuance, to be implemented by May 
15, 2005.
    Amendment No.: 297.
    Facility Operating License No. DPR-77: Amendment revises the 
implementation date for License Amendment No. 294.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC):
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated November 9, 
2004.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

    Dated at Rockville, Maryland, this 15th day of November, 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-25664 Filed 11-22-04; 8:45 am]
BILLING CODE 7590-01-P


</pre><script data-cfasync="false" src="/cdn-cgi/scripts/5c5dd728/cloudflare-static/email-decode.min.js"></script></body>
</html>