[Federal Register Volume 70, Number 11 (Tuesday, January 18, 2005)]
[Notices]
[Pages 2886-2907]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-779]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 23, 2004, through January 5, 2005. 
The last biweekly notice was published on January 4, 2005 (70 FR 398).

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

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

[[Page 2887]]

opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: July 15, 2004, supplemented by letter 
dated August 23, 2004.
    Description of amendment request: The amendment would revise 
Operating License DPR-65 to address the resolution of a non-
conservative Technical Specification (TS) associated with control room 
isolation radiation monitoring instrumentation. Specifically, the 
amendment would revise the TS to require two operable channels of 
control room isolation radiation monitoring instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves requirements to maintain two 
operable channels in order to add a level of detection capability 
and greater assurance that the safety function for control room 
isolation is met. In addition, the proposed change will not alter 
the setpoint value for the radiation monitors nor will it affect the 
method for control room air filtration during the emergency mode of 
operation. Therefore, the proposed change from one operable channel 
to two operable channels for the control room isolation radiation 
monitoring instrumentation will not increase the probability of 
consequences of any previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change involves radiation monitoring channels 
designed to send a signal to isolate the control room when high 
radiation levels are detected to limit the radiological dose to the 
control room operators in the event of an accident. In addition, the 
proposed change will not have an impact on the setpoint value to 
change the radiation level at which control room isolation is 
assumed to occur. Again, the proposed change will not introduce 
failure modes, accident initiators, or malfunctions. Therefore, the 
proposed change from one operable channel to two operable channels 
for the control room isolation radiation monitoring instrumentation, 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Increasing the number of radiation monitoring channels for the 
control room isolation radiation monitoring instrumentation will not 
reduce a margin of safety. The proposed change to add requirements 
to the TS for a redundant radiation monitoring channel will increase 
the reliability of the system to perform its intended function. In 
addition, the proposed change will add appropriate compensatory 
actions for conditions when both channels are not available. 
Therefore, given that the proposed change will continue to meet the 
current design basis, any reduction in a margin of safety would not 
be significant.

    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.929(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

[[Page 2888]]

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New London 
County, Connecticut

    Date of amendment request: September 8, 2002.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications to support the implementation of 
the proposed Dominion Nuclear Facility Quality Assurance Program 
(Topical Report DOM-QA-1). Implementation of this Topical Report would 
create a common quality assurance program for all sites owned by 
Dominion Nuclear Connecticut, Inc. Review of this proposed amendment 
was requested to be done in concert with review of the Topical Report. 
The Topical Report is available in the Agencywide Document Access and 
Management System under accession number ML042470015.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequence of an accident previously analyzed. 
The changes involve the transfer of requirements from the 
administrative section of the Technical Specifications to the 
Consolidated Quality Assurance Program and other licensee controlled 
documents. Therefore, the proposed changes are administrative in 
nature, and have no effect on a design basis accident, and will not 
increase the probability or consequences of any previously analyzed 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The implementation of the proposed changes does not create the 
possibility of an accident of a different type than was previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR). The 
transfer of requirements concerning facility staff qualifications 
from the administrative section of the Technical Specifications to 
the Consolidated Quality Assurance Program and other licensee 
controlled documents can not initiate a new or different kind of 
accident.
    These changes do not alter the nature of events postulated in 
the UFSAR nor do they introduce any unique precursor mechanisms. 
Therefore, the proposed changes are administrative in nature and do 
not create the possibility of a new or different kind of accident 
from those previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The implementation of the proposed changes does not reduce the 
margin of safety. The proposed changes to transfer certain 
requirements from the administration section of the Technical 
Specifications to the Consolidated Quality Assurance Program and 
other licensee controlled documents have no effect on design bases 
radiological events. It is thus concluded that the proposed changes 
are administrative in nature and the margin of safety will not be 
reduced by the implementation of the changes.

    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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: December 6, 2004.
    Description of amendment request: The proposed amendment would make 
administrative changes to the Technical Specifications (TSs) including 
correction of references and deleting obsolete or redundant TS 
requirements and surveillances.
    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 operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes are administrative or editorial in nature 
and do not involve any physical changes to the plant. The changes do 
not revise the methods of plant operation which could increase the 
probability or consequences of accidents. No new modes of operation 
are introduced by the proposed changes such that a previously 
evaluated accident is more likely to occur or more adverse 
consequences would result.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    These changes are administrative or editorial in nature and do 
not affect the operation of any systems or equipment, nor do they 
involve any potential initiating events that would create any new or 
different kind of accident. There are no changes to the design 
assumptions, conditions, configuration of the facility, or manner in 
which the plant is operated and maintained. The changes do not 
affect assumptions contained in plant safety analyses or the 
physical design and/or modes of plant operation. Consequently, no 
new failure mode is introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    There are no changes being made to the Technical Specification 
(TS) safety limits or safety system settings. The operating limits 
and functional capabilities of systems, structures and components 
are unchanged as a result of these administrative and editorial 
changes. These changes do not affect any equipment involved in 
potential initiating events or plant response to accidents. There is 
no change to the basis for any TS that is related to the 
establishment, or maintenance of, a nuclear safety margin.
    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: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: Allen G. Howe.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: December 7, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to: (1) Delete the 
surveillance requirement (SR) associated with testing of the standby 
liquid control (SLC) pump discharge pressure relief valves; and (2) 
remove details from the SR for testing of the recirculation pump 
discharge valves.
    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 operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance

[[Page 2889]]

with the proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment removes details of SLC pressure relief 
valve and recirculation pump discharge valve testing requirements 
from the TS. Following implementation of the proposed change, the VY 
TS will still require operability testing of the subject components 
by reference to the VY IST [Inservice Testing] Program. Details of 
SLC pressure relief valve and recirculation pump discharge valve 
testing requirements will still be contained in the VY IST Program. 
The SLC pressure relief valve and recirculation pump discharge valve 
setpoint values related to the safety functions of those systems 
will continue to be contained in the VY UFSAR [Updated Final Safety 
Analysis Report]. Changes to the VY UFSAR are evaluated per the 
requirements of 10 CFR 50.59. These controls are adequate to ensure 
the required inservice testing is performed to verify the components 
are operable and capable of performing their respective safety 
functions. The proposed amendment introduces no new equipment or 
changes to how equipment is operated. Neither the SLC pressure 
relief valves nor the recirculation pump discharge valves are 
initiators of any analyzed accidents. Therefore, operation of VY in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed amendment removes details of SLC pressure relief 
valve and recirculation pump discharge valve testing requirements 
from the TS. The proposed amendment does not change the design or 
function of any component or system. No new modes of failure or 
initiating events are being introduced. Therefore, operation of VY 
in accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed amendment removes details of SLC pressure relief 
valve and recirculation pump discharge valve testing requirements 
from the TS. The proposed amendment does not change the design or 
function of any component or system. The proposed amendment does not 
involve any safety limits or limiting safety system settings.
    Since the proposed controls are adequate to ensure the required 
inservice testing is performed, there will still be high assurance 
that the components are operable and capable of performing their 
respective safety functions, and that the systems will respond as 
designed to mitigate the subject events. Therefore, operation of VY 
in accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037-1128.
    NRC Section Chief: Allen G. Howe.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: December 15, 2004.
    Description of amendment request: The proposed amendment would 
revise the limiting conditions for operation in Technical Specification 
(TS) 3.3 and the surveillance requirements in TS 4.3 associated with 
the control rod system. Specifically, the proposed changes would revise 
the TSs associated with: (1) Control rod operability; (2) control rod 
scram time testing; and (3) control rod accumulator operability.
    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 operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes do not significantly affect the design or 
fundamental operation and maintenance of the plant. Accident 
initiators or the frequency of analyzed accident events are not 
significantly affected as a result of the proposed changes; 
therefore, there will be no significant change to the probabilities 
of accidents previously evaluated.
    The proposed changes do not significantly alter assumptions or 
initial conditions relative to the mitigation of an accident 
previously evaluated. The proposed changes continue to ensure 
process variables, structures, systems, and components (SSCs) are 
maintained consistent with the safety analyses and licensing basis. 
The revised technical specifications continue to require that SSCs 
are properly maintained to ensure operability and performance of 
safety functions as assumed in the safety analyses. The design basis 
events analyzed in the safety analyses will not change significantly 
as a result of the proposed changes to the TS.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes do not involve any physical alteration of 
the plant (no new or different type of equipment being installed) 
and do not involve a change in the design, normal configuration or 
basic operation of the plant. The proposed changes do not introduce 
any new accident initiators. In some cases, the proposed changes 
impose different requirements; however, these new requirements are 
consistent with the assumptions in the safety analyses and current 
licensing basis. Where requirements are relocated to other licensee-
controlled documents, adequate controls exist to ensure their proper 
maintenance.
    The proposed changes do not involve significant changes in the 
fundamental methods governing normal plant operation and do not 
require unusual or uncommon operator actions. The proposed changes 
provide assurance that the plant will not be operated in a mode or 
condition that violates the essential assumptions or initial 
conditions in the safety analyses and that SSCs remain capable of 
performing their intended safety functions as assumed in the same 
analyses. Consequently, the response of the plant and the plant 
operator to postulated events will not be significantly different.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. The proposed changes do 
not significantly affect any of the assumptions, initial conditions 
or inputs to the safety analyses. Plant design is unaffected by 
these proposed changes and will continue to provide adequate 
defense-in-depth and diversity of safety functions as assumed in the 
safety analyses.
    There are no proposed changes to any of the Safety Limits or 
Limiting Safety System Setting requirements. The proposed changes 
maintain requirements consistent with safety analyses assumptions 
and the licensing basis. Fission product barriers will continue to 
meet their design capabilities without any significant impact to 
their ability to maintain parameters within acceptable limits. The 
safety functions are maintained within acceptable limits without any 
significant decrease in capability.
    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 2890]]

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. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: Allen G. Howe.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: December 20, 2004.
    Description of amendment request: The requested change will delete 
the requirements in 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 December 20, 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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating letter 
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.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not 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.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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 significance hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Michael A. Webb (Acting).

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

    Date of amendment request: December 20, 2004.
    Description of amendment request: The requested change will delete 
the requirements in Technical Specification (TS) 6.6.1, ``Occupational 
Radiation Exposure Report,'' and TS 6.6.4, ``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 December 20, 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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) 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.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not 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.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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 significance hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Michael A. Webb (Acting).

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

    Date of amendment request: October 14, 2004.
    Brief description of amendments: The proposed change will revise 
the surveillance requirement (SR) 3.6.6.8 frequency of every 10 years. 
Instead, the proposed change to SR 3.6.6.8 will require verification 
that spray nozzles are unobstructed following maintenance that could 
result in nozzle blockage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below and states that the amendment 
request:

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change modifies the [Surveillance Requirements] SR 
to verify that the [Reactor Building] RB spray nozzles are 
unobstructed after maintenance that could introduce material that 
could result in nozzle blockage. The spray nozzles are not assumed 
to be initiators of any previously analyzed

[[Page 2891]]

accident. Therefore, the change does not increase the probability of 
any accident previously evaluated. The spray nozzles are assumed in 
the accident analyses to mitigate design basis accidents. The 
revised SR to verify system OPERABILITY following maintenance is 
considered adequate to ensure OPERABILITY of the RB spray system. 
Since the system will still be able to perform its accident 
mitigation function, the consequences of accidents previously 
evaluated are not increased. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    The proposed change revises the SR to verify that the RB spray 
nozzles are unobstructed after maintenance that could result in 
nozzle blockage. The change does not introduce a new mode of plant 
operation and does not involve physical modification to the plant. 
The change will not introduce new accident initiators or impact the 
assumptions made in the safety analysis. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed change revises the frequency for performance of the 
SR to verify that the RB spray nozzles are unobstructed. The 
frequency is changed from every 10 years to following maintenance 
that could result in nozzle blockage. This requirement, along with 
foreign material exclusion programs and the remote physical location 
of the spray nozzles, provides assurance that the spray nozzles will 
remain unobstructed. As the spray nozzles are expected to remain 
unobstructed and able to perform their post-accident mitigation 
function, plant safety is not significantly affected. 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: 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, Jr.

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

    Date of amendment request: November 22, 2004.
    Description of amendment request: The requested change will delete 
the requirements in 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 November 22, 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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) 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.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not 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.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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 significance hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Michael K. Webb (Acting).

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

    Date of amendment request: April 29, 2004, as supplemented November 
23, 2004.
    Description of amendment request: The proposed amendment is a 
selective-scope application of an alternative source term (AST) for the 
fuel handling accident (FHA) in accordance with Title 10 of the Code of 
Federal Regulations (10 CFR) Section 50.67, ``Accident Source Term.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment involves implementation of the AST for 
the fuel handling accident at MNGP [Monticello Nuclear Generating 
Plant]. There are no physical design modifications to the plant 
associated with the proposed amendment. The revised calculations do 
not impact the initiators of an FHA in any way.
    The changes also do not impact the initiators for any other 
design basis accident (DBA) or events. Therefore, because DBA 
initiators are not being altered by adoption of the AST analyses, 
the probability of an accident previously evaluated is not affected.
    With respect to consequences, the only previously evaluated 
accident that could be affected is the FHA. The AST is an input to 
calculations used to evaluate the consequences of the accident, and 
does not, in and of itself, affect the plant response or the actual 
pathways to the environment utilized by the radiation/activity 
released by the fuel. It does however, better represent the physical 
characteristics of the release, so that appropriate mitigation 
techniques may be applied. For the FHA, the AST analyses demonstrate 
acceptable doses that are within regulatory limits after 24 hours of 
radiological decay, without credit for Secondary Containment 
integrity, selected ESF [engineered safety feature] filtration 
system operation (i.e., SBGT [standby gas treatment] System or 
Control Room EFT [emergency filtration] System) or Control Room 
isolation. Therefore, the consequences of an accident previously 
evaluated are not significantly increased.
    Based on the above conclusions, this proposed amendment does not 
involve a

[[Page 2892]]

significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a physical alteration of 
the plant. No new or different types of equipment will be installed 
and there are no physical modifications to existing equipment 
associated with the proposed changes. Also, no changes are proposed 
to the methods governing plant/system operation during handling of 
irradiated fuel, so no new initiators or precursors of a new or 
different kind of accident are created. New equipment or personnel 
failure modes that might initiate a new type of accident are not 
created as a result of the proposed amendment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed amendment is associated with the implementation of 
a new licensing basis for the MNGP FHA. Approval of this change from 
the original source term to an alternative source term derived in 
accordance with the guidance of RG 1.183 [``Alternative Radiological 
Source Terms for Evaluating Design Basis Accidents at Nuclear Power 
Reactors''] is being requested. The results of the FHA accident 
analysis, revised in support of the proposed license amendment, are 
subject to revised acceptance criteria. The AST FHA analysis has 
been performed using conservative methodologies, as specified in RG 
1.183. Safety margins have been evaluated and analytical 
conservatism has been utilized to ensure that the analyses 
adequately bound the postulated limiting event scenario. The dose 
consequences of the limiting FHA remain within the acceptance 
criteria presented in 10 CFR 50.67 and RG 1.183.
    The proposed changes continue to ensure that the doses at the 
Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) 
boundaries, as well as the Control Room, are within the 
corresponding regulatory limits. For the FHA, RG 1.183 
conservatively sets the EAB and LPZ limits below the 10 CFR 50.67 
limit, and sets the Control Room limit consistent with 10 CFR 50.67.
    Since the proposed amendment continues to ensure the doses at 
the EAB, LPZ and Control Room are within corresponding regulatory 
limits, the proposed license amendment does not involve a 
significant reduction in a margin of safety.
    Based on the above, NMC has determined that operation of the 
facility in accordance with the proposed change does not involve a 
significant hazards consideration as defined in 10 CFR 50.92(c), in 
that it: (1) Does not involve a significant increase in the 
probability or consequences of an accident previously evaluated; (2) 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated; and (3) does not 
involve a significant reduction in a margin of safety.

    The U. S. Nuclear Regulatory Commission (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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: June 30, 2004, as supplemented November 
5, 2004.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) to implement a 24-month fuel 
cycle.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration (NSHC), which is 
presented below:

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

a. Surveillance Testing Interval Extensions

    The proposed Technical Specification (TS) changes involve 
changes in the surveillance testing to facilitate a change in the 
operating cycle from 18 months to 24 months. The proposed TS changes 
do not physically impact the normal operation of the plant, nor do 
they impact any design or functional requirements of the associated 
systems. That is, the proposed TS changes neither impact the TS SRs 
[surveillance requirements] themselves nor the manner in which the 
surveillances are performed.
    In addition, the proposed TS changes do not introduce any 
accident initiators, since no accidents previously evaluated relate 
to the frequency of surveillance testing. Also, evaluations of the 
proposed TS changes demonstrate that the availability of equipment 
and systems required to prevent or mitigate the radiological 
consequences of an accident are not significantly affected because 
of other, more frequent testing that is performed, the availability 
of redundant systems and equipment, or the high reliability of the 
equipment. Since the impact on the systems is minimal NMC [Nuclear 
Management Company] has concluded that the overall impact on the 
plant safety analysis is negligible.
    A historical review of surveillance test results and associated 
maintenance records indicated that there was no evidence of any 
failure that would invalidate the above conclusions.
    Therefore, the proposed TS changes do not significantly increase 
the probability or consequences of an accident previously evaluated.

b. TS Trip Setting Changes

    Changes are proposed to the Monticello TS Trip Settings. The 
proposed changes are a result of application of the Monticello 
Instrument Setpoint Methodology using plant-specific drift values. 
Application of this methodology results in Trip Setpoints that more 
accurately reflect total instrumentation loop accuracy, as well as 
that of test equipment and calculated drift between surveillances. 
The proposed changes will not result in hardware changes. The 
instrumentation is not assumed to be initiators of any analyzed 
events, nor do they impact any design or functional requirements of 
the associated systems. Existing operating margins between plant 
conditions and actual plant setpoints are not significantly reduced 
due to the proposed changes. The role of the instrumentation is in 
mitigating and thereby, limiting the consequences of accidents.
    The Nominal Trip Setpoints were developed to ensure the design 
and safety analysis limits are satisfied. The methodology used for 
the development of the Trip Settings ensures: (1) The affected 
instrumentation remains capable of mitigating design basis events as 
described in the safety analysis; and, (2) the results and 
radiological consequences described in the safety analysis remain 
bounding. The proposed changes do not alter the plant's ability to 
detect and mitigate events.
    Therefore, these changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

c. Surveillance Testing Interval Reductions

    The proposed TS changes involve reductions in the surveillance 
testing intervals from once per operating cycle or refueling outage 
to once every three (3) months or once per quarter for the equipment 
associated with these TS SRs. The shorter intervals are based upon 
the plant-specific results of a review of the surveillance test 
history for this equipment. The implementing procedures for these 
SRs have been performed on a once per three (3) month or once per 
quarter interval for a number of years, and these changes more 
accurately reflect actual plant maintenance practices. The proposed, 
more restrictive TS changes do not physically impact the plant, nor 
do they impact any design or functional requirements of the 
associated systems. That is, the proposed TS changes neither degrade 
the performance of, nor increase the challenges to, any safety 
system assumed to function in the safety analysis. These proposed TS 
changes neither impact the TS SRs themselves nor the manner in which 
the surveillances are performed.
    The proposed TS changes do not introduce any accident 
initiators, since no accident previously evaluated relate to the 
frequency

[[Page 2893]]

of surveillance testing. The proposed TS intervals demonstrate that 
the equipment and systems required to prevent or mitigate the 
radiological consequences of an accident are continuing to meet the 
assumptions of the setpoint evaluation on a more frequent basis. 
Since the impacts on systems are minimal and the assumptions of the 
safety analyses are maintained, NMC has concluded that the overall 
impact on the plant safety analysis is negligible.
    Therefore, the proposed TS changes do not significantly increase 
the probability or consequences of any accident previously 
evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind or accident from any accident previously 
evaluated.

a. Surveillance Testing Interval Extensions

    The proposed TS changes involve changes in the surveillance 
testing intervals to facilitate a change in the operating cycle 
length. The proposed TS changes do not introduce any failure 
mechanisms of a different type than those previously evaluated. 
There are no physical changes being made to the facility. No new or 
different equipment is being installed. No installed equipment is 
being operated in a different manner. As a result no new failure 
modes are introduced. The SRs themselves, and the manner in which 
surveillance tests are performed, remain unchanged.
    A historical review of surveillance test results and associated 
maintenance records indicated that there was no evidence of any 
failure that would invalidate the above conclusions.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

b. TS Trip Setting Changes

    The proposed changes to the Trip Settings are a result of 
applying the Monticello Instrument Setpoint Methodology using plant-
specific drift values. The application of this methodology does not 
create the possibility of any new or different kinds of accidents 
from any accidents previously evaluated. This is based upon the fact 
that the method and manner of plant operations are unchanged.
    The use of the proposed Trip Setpoints does not impact the safe 
operation of the plant in that the safety analysis limits are 
maintained. The proposed changes in Trip Settings involve no system 
additions or physical modifications to plant systems. The Trip 
Settings are revised to ensure the affected instrumentation remains 
capable of mitigating accidents and transients. Plant equipment will 
not be operated in a manner different from previous operation. Since 
operational methods remain unchanged and the operating parameters 
were evaluated to maintain the plant within existing design basis 
criteria no different type of failure or accident is created.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

c. Surveillance Testing Interval Reductions

    The proposed TS changes involve reductions in the surveillance 
testing intervals from once per operating cycle or refueling outage 
to once every three (3) months or once per quarter for the equipment 
associated with these TS SRs. The shorter intervals are based upon 
the plant-specific results of a review of the surveillance test 
history for this equipment. The implementing procedures for these 
SRs have been performed on a once per three (3) month or once per 
quarter interval for a number of years and these changes more 
accurately reflect actual plant maintenance practices. The proposed 
more restrictive TS changes do not physically impact the plant, nor 
do they impact any design or functional requirements of the 
associated systems. That is, the proposed TS changes neither degrade 
the performance of, nor increase the challenges to, any safety 
system assumed to function in the safety analysis. These proposed TS 
changes neither impact the TS SRs themselves nor the manner in which 
the surveillances are performed.
    The proposed TS changes do not introduce any failure mechanism 
of a different type than those previously evaluated. The proposed 
changes make no physical changes to the plant. No new or different 
equipment is being installed. No installed equipment is being 
operated in a different manner.
    A historical review of surveillance test results and associated 
maintenance records indicate that there is no evidence of any 
failure that would invalidate the above conclusions.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.

a. Surveillance Testing Interval Extensions

    Although the proposed TS changes result in changes in the 
interval between surveillance tests, the impact, if any, on system 
availability is minimal based upon other, more frequent testing that 
is performed, the existence of redundant systems and equipment or 
overall system reliability. Evaluations show there is no evidence of 
any time-dependant failure that would impact system availability.
    The proposed changes do not significantly impact the condition 
or performance of structures, systems and components relied upon for 
accident mitigation. The proposed TS changes do not physically 
impact the plant, nor do they impact any design or functional 
requirements of the associated systems. The proposed changes do not 
significantly impact any safety analysis assumptions or results.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

b. TS Trip Setting Changes

    The proposed changes do not involve a reduction in a margin of 
safety. The proposed changes were developed using a Monticello 
Instrument Setpoint Methodology using plant-specific drift values. 
This methodology ensures no safety analysis limits are exceeded. The 
proposed TS changes do not physically impact the plant, nor do they 
impact any design or functional requirements of the associated 
systems.
    As such, these proposed changes do not involve a reduction in a 
margin of safety.

c. Surveillance Testing Interval Reductions

    The proposed TS changes result in a shorter interval between 
surveillance tests to ensure the assumptions of the safety analysis 
are maintained. The impact, if any, on system availability is 
minimal, as a result of the more frequent testing that is performed. 
The proposed changes do not significantly impact the condition or 
performance of structures, systems and components relied upon for 
accident mitigation. The proposed TS changes do not physically 
impact the plant, nor do they impact any design or functional 
requirements of the associated systems. The proposed changes do not 
significantly impact any safety analysis assumptions or results.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The U. S. Nuclear Regulatory Commission (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 NSHC.
    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.

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

    Date of amendment request: September 8, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the technical specifications (TS) to maintain 
containment hydrogen monitors. Licensees were generally required to 
implement upgrades as described in NUREG-0737, ``Clarification of TMI 
[Three Mile Island] Action Plan Requirements,'' and Regulatory Guide 
(RG) 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI Unit 2. 
Requirements related to combustible gas control were imposed by Order 
for many facilities and were added to or included in the TS for nuclear 
power reactors currently licensed to operate. The revised 10 CFR 50.44, 
``Standards for Combustible Gas Control System in

[[Page 2894]]

Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
relevant portions of the model NSHC determination (TS for Fort Calhoun 
do not include requirements for hydrogen recombiners) in its 
application dated September 8, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOPs), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    Based upon the reasoning presented above, the requested change does 
not involve a significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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 amendment requests: November 1, 2004.
    Description of amendment requests: The requested change will 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 November 1, 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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating letter 
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.

[[Page 2895]]

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 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.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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 a significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 22, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), part 50, Sec.  50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TSs would be 
eliminated, several notes or specific exceptions are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated September 22, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment requests: December 10, 2004.
    Description of amendment requests: The proposed amendment will 
delete the requirements from the Technical Specifications (TS) to 
maintain hydrogen recombiners and hydrogen monitors. Licensees were 
generally required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TS for nuclear power reactors currently licensed to operate.

[[Page 2896]]

The revised Sec.  50.44 of Title 10 of the Code of Federal Regulations 
(10 CFR), ``Standards for Combustible Gas Control System in Light-
Water-Cooled Power Reactors,'' eliminated the requirements for hydrogen 
recombiners and relaxed safety classifications and licensee commitments 
to certain design and qualification criteria for hydrogen and oxygen 
monitors.
    The proposed license amendment will revise TS 3.3.11, ``Post 
Accident Monitoring Instrumentation (PAMI),'' to delete the Note in 
Condition C. Also in TS 3.3.11, Condition D will be deleted. In TS 
Table 3.3.11-1, Item 10, ``Containment Hydrogen Monitors,'' is deleted. 
Other TS changes included in this application are limited to 
renumbering and formatting changes that resulted directly from the 
deletion of the above requirements related to hydrogen monitors. The 
changes to TS requirements result in changes to various TS Bases 
sections. The TS Bases changes will be submitted with a future update 
in accordance with TS 5.4.4, ``Technical Specifications (TS) Bases 
Control.''
    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 September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 10, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG 1.97 Category 1, is intended for key variables that most directly 
indicate the accomplishment of a safety function for design-basis 
accident events. The hydrogen monitors no longer meet the definition 
of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 
50.44 the Commission found that Category 3, as defined in RG 1.97, 
is an appropriate categorization for the hydrogen monitors because 
the monitors are required to diagnose the course of beyond design-
basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the SAMGs [severe accident 
management guidelines], the emergency plan (EP), the emergency 
operating procedures (EOP), and site survey monitoring that support 
modification of emergency plan protective action recommendations 
(PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.

    Therefore, this change does not involve a significant reduction in 
the margin of safety. Removal of hydrogen monitoring from TS will not 
result in a significant reduction in their functionality, reliability, 
and availability.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment requests: December 17, 2004.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' 
TS 3.8.4, ``DC Sources--Operating,'' TS 3.8.5, ``DC Sources--
Shutdown,'' TS 3.8.6, ``Battery Cell Parameters,'' TS 3.8.7, 
``Inverters--Operating,'' and TS 3.8.9, ``Distribution Systems--
Operating.'' This change will also add a new Battery Monitoring and 
Maintenance Program, section 5.5.2.16. The proposed change will provide 
operational flexibility to credit DC electrical subsystem design 
upgrades that are in progress. These upgrades will provide increased 
capacity batteries, additional battery chargers, and the

[[Page 2897]]

means to cross-connect DC subsystems while meeting all design battery 
loading requirements. With these modifications in place, it will be 
feasible to perform routine surveillance as well as battery 
replacements online.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to Technical Specifications (TS) 3.8.4 and 
3.8.6 would allow extension of the Completion Time (CT) for 
inoperable Direct Current (DC) distribution subsystems to manually 
cross-connect DC distribution buses of the same safety train of the 
operating unit for a period of 30 days. Currently the CT only allows 
for 2 hours to ascertain the source of the problem before a 
controlled shutdown is initiated. Loss of a DC subsystem is not an 
initiator of an event. However, complete loss of a Train A 
(subsystems A and C) or Train B (subsystems B and D) DC system would 
initiate a plant transient/plant trip.
    Operation of a DC Train in cross-connected configuration does 
not affect the quality of DC control and motive power to any system. 
Therefore, allowing the cross-connect of DC distribution systems 
does not significantly increase the probability of an accident 
previously evaluated in Chapter 15 of the Updated Final Safety 
Analysis Report (UFSAR).
    The above conclusion is supported by Probabilistic Risk Analysis 
(PRA) evaluation which encompasses all accidents, including UFSAR 
Chapter 15.
    Modification to the Frequency for Surveillance Requirement (SR) 
3.8.6.1 is consistent with the recommendations of TSTF 360 Rev. 1 
and IEEE 450-2002, and similarly does not impact safety 
considerations.
    Further changes are made of an editorial nature or provide 
clarification only. For example, discussions regarding electrical 
`Trains' and `Subsystems' will be in more conventional terminology. 
Limiting Condition for Operations (LCOs) affected by editorial 
changes include 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9.
    Enhancements from TSTF-360, Rev. 1 and IEEE 450-2002 have been 
incorporated into LCOs 3.8.4 and 3.8.6. TSTF-360, Rev. 1 was 
previously approved by the NRC, and IEEE 450-2002 includes industry-
generic recommendations.
    The changes being proposed do not affect assumptions contained 
in other safety analyses or the physical design of the plant other 
than the upgrades of the electrical systems described in this 
change, nor do they affect other Technical Specifications that 
preserve safety analysis assumptions.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change to Technical Specifications 3.8.4 will 
enable the cross-tie of subsystems. New equipment, swing battery 
chargers, distribution panels, and associated protective devices are 
added to increase overall DC system reliability. Both administrative 
and mechanical controls will be in place to ensure the design and 
operation of the DC distribution systems continue to perform to 
applicable design standards. During cross connecting of subsystem 
buses, two batteries would be paralleled for a short duration. An 
electrical fault during that duration could exceed the interrupting 
duties of the protective devices. This is standard industry practice 
during transfer of power sources and is considered to be an 
acceptable minimal risk. For example, the design of the 1E 4kV power 
system is based on this practice as well. Therefore, the addition of 
new equipment does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Enhancements from TSTF-360, Rev. 1 and IEEE 450-2002 have been 
incorporated into LCOs 3.8.4 and 3.8.6. TSTF-360, Rev. 1 is 
previously approved and IEEE 450-2002 includes industry-generic 
recommendations. Enhancements, including surveillance intervals or 
required completion times, will not create the possibility of a new 
or different kind of accident from any previously evaluated.
    LCOs 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9 are revised to 
incorporate editorial changes. Since these changes do not affect 
plant design but enhance clarity, these modifications do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    Therefore, operation of the facility in accordance with this 
proposed change will not create the possibility of new or different 
kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change does not alter the bases for assurance that 
safety-related activities are performed correctly or the basis for 
any Technical Specification that is related to the establishment of 
or maintenance of a safety margin. Specifically, battery sizing 
calculations continue to show that new upgraded capacity batteries 
will meet the most limiting load profile that includes margin for 
growth, with aging and temperature correction. Battery modified 
performance discharge testing will demonstrate on an on-going basis 
that battery capacity will be greater than or equal to 80% of 
original design requirements at all times during service life and 
that the service profiles will be met as is currently required by 
Surveillance Requirements 3.8.4.7 and 3.8.4.8. The addition of the 
DC cross-tie capability proposed for LCO 3.8.4 will ensure 
appropriate operations of the DC buses during maintenance activities 
such as battery testing or replacement. Enhancements from TSTF-360, 
Rev. 1 and IEEE 450-2002 have been incorporated into LCOs 3.8.4 and 
3.8.6. TSTF-360, Rev. 1 is previously approved and IEEE 450-2002 
includes industry-generic recommendations. Enhancements including 
surveillance intervals or required completion times will not involve 
a significant reduction in a margin of safety.
    Also, LCOs 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9 are 
revised to incorporate editorial changes. Since these changes do not 
affect plant design or operations but should enhance clarity, these 
modifications would not involve a significant reduction in margin of 
safety.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Robert A. Gramm.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: October 26, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of title 10 of the Code of Federal 
Regulations (10 CFR) part 50, Sec.  50.65(a)(4). Limiting Condition for 
Operation (LCO) 3.0.4 exceptions in individual TSs would be eliminated, 
several notes or specific exceptions are revised to reflect the related 
changes to LCO 3.0.4, and Surveillance Requirement 3.0.4 is revised to 
reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-

[[Page 2898]]

359. The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50475), on possible 
amendments concerning TSTF-359, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability 
of the following NSHC determination in its application dated October 
26, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: John A. Nakoski.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of amendment request: December 6, 2004 (TS 426).
    Description of amendment request: The proposed amendment would 
revise the current Unit 1 Diesel Generators (DG) Allowed Outage Time 
(AOT) in the Technical Specifications (TS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The DGs are designed as backup AC power sources in the event 
of loss of offsite power. The proposed DG TS AOT does not change the 
conditions, operating configurations, or minimum amount of operating 
equipment assumed in the safety analysis for accident mitigation. No 
changes are proposed in the manner in which the DGs provide plant 
protection or which create new modes of plant operation. In 
addition, a PSA [probabilistic safety assessment] evaluation 
concluded that the risk contribution of the DG TS AOT extension is 
non-risk significant. Therefore, the proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed amendment does not introduce new equipment, 
which could create a new or different kind of accident. No new 
external threats, release pathways, or equipment failure modes are 
created. Therefore, the implementation of the proposed amendment 
will not create a possibility for an accident of a new or different 
type than those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. BFN's emergency AC system is designed with sufficient 
redundancy such that a DG may be removed from service for 
maintenance or testing. The remaining DGs are capable of carrying 
sufficient electrical loads to satisfy the UFSAR [Updated Final 
Safety Analysis Report] requirements for accident mitigation or unit 
safe shutdown.

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

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant (BFN), Unit 1, Limestone County, Alabama

    Date of amendment request: December 6, 2004 (TS 428).
    Description of amendment request: The proposed amendment would 
revise the reactor vessel Pressure-Temperature (P-T) curves depicted in 
the Technical Specification (TS) Figure 3.4.9-1 and adds a new TS 
Figure 3.4.9-2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 2899]]

    No. The proposed changes deal exclusively with the reactor 
vessel P-T curves, which define the permissible regions for 
operation and testing. Failure of the reactor vessel is not 
considered as a design basis accident. Through the design 
conservatisms used to calculate the P-T curves, reactor vessel 
failure has a low probability of occurrence and is not considered in 
the safety analyses. The proposed changes adjust the reference 
temperature for the limiting material to account for irradiation 
effects and provide the same level of protection as previously 
evaluated and approved.
    The adjusted reference temperature calculations were performed 
in accordance with the requirements of 10 CFR 50 Appendix G using 
the guidance contained in Regulatory Guide 1.190, ``Calculational 
and Dosimetry Methods for Determining Pressure Vessel Neutron 
Fluence,'' to reflect use of the operating limits to no more than 16 
Effective Full Power Years (EFPY). These changes do not alter or 
prevent the operation of equipment required to mitigate any accident 
analyzed in the BFN Final Safety Analysis Report.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed changes to the reactor vessel P-T curves do not 
involve a modification to plant equipment. No new failure modes are 
introduced. There is no effect on the function of any plant system, 
and no new system interactions are introduced by this change. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed curves conform to the guidance contained in 
Regulatory Guide (RG) 1.190, ``Calculational and Dosimetry Methods 
for Determining Pressure Vessel Neutron Fluence,'' and maintain the 
safety margins specified in 10 CFR 50 Appendix G. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

Tennessee Valley Authority (TVA), Docket No. 50-328, Sequoyah Nuclear 
Plant, Unit 2, Hamilton County, Tennessee

    Date of amendment request: December 2, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.4.5, ``Steam Generators,'' 
including associated Bases 3/4.4.5 to change the inspection scope of 
steam generator tubing in the Westinghouse Electric Company explosive 
tube expansion region below the top of the tubesheet. Additionally, the 
proposed TS change removes the axial primary water stress corrosion 
cracking at dented tube support plate alternate repair criteria and the 
associated note for the exclusion made for Unit 2 Cycle 12 operation 
only and changes the current definition of plugging limit to exclude 
possible indications below the W* distance.
    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.
    Of the various accidents previously evaluated, the proposed 
changes only affect the steam generator tube rupture (SGTR) event 
evaluation and the postulated steam line break (SLB) accident 
evaluation. Loss-of-coolant accident (LOCA) conditions cause a 
compressive axial load to act on the tube. Therefore, since the LOCA 
tends to force the tube into the tubesheet rather than pull it out, 
it is not a factor in this amendment request. Another faulted load 
consideration is a safe shutdown earthquake (SSE); however, the 
seismic analysis of Westinghouse 51 series SGs has shown that axial 
loading of the tubes is negligible during an SSE.
    TVA's amendment request takes credit for how the tubesheet 
enhances the tube integrity in the Westinghouse Electric Company 
explosive tube expansion (WEXTEX) region by precluding tube 
deformation beyond its initial expanded outside diameter. For the 
SGTR and SLB events, the required structural margins of the SG tubes 
will be maintained due to the presence of the tubesheet. Tube 
rupture is precluded for axial cracks in the WEXTEX region due to 
the constraint provided by the tubesheet. Therefore, the normal 
operating 3[Delta]P margin and the postulated accident 1.43[Delta]P 
margin against burst are maintained.
    The W* length supplies the necessary resistive force to preclude 
pullout loads under both normal operating and accident conditions. 
The contact pressure results from the WEXTEX expansion process, 
thermal expansion mismatch between the tube and tubesheet, and from 
the differential pressure between the primary and secondary side. 
Therefore, the proposed change results in no significant increase in 
the probability or the occurrence of an SGTR or SLB accident.
    The proposed changes do not affect other systems, structures, 
components or operational features. Therefore, based on the above 
evaluation, the proposed changes do not involve a significant 
increase in the probability of an accident previously evaluated.
    The consequences of an SGTR event are primarily affected by the 
primary-to-secondary flow rate and the time duration of the primary-
to-secondary flow during the event. Primary-to-secondary flow rate 
through a postulated ruptured tube (i.e., complete severance of a 
single SG tube) is not affected by the proposed change since the 
flow rate is based on the inside diameter of a SG tube and the 
pressure differential. TVA's amendment request does not change 
either of these. The duration of primary-to-secondary leakage is 
based on the time required for an operator to determine that a SGTR 
has occurred, the time to identify and isolate the faulty SG, and 
ensure termination of radioactive release to the atmosphere from the 
faulty SG. TVA's amendment request does not affect the duration of 
the primary-to-secondary leakage because it does not change the 
control room indicators with which an operator would determine that 
an SGTR has occurred. The consequences of an SGTR are secondarily 
affected by primary-to-secondary leakage, which could occur due to 
axial cracks remaining in service in the WEXTEX region in a non-
faulted SG. During a SGTR, the primary-to-secondary differential 
pressure is less than or equal to the normal operating differential 
pressure; therefore, the primary-to-secondary leakage due to axial 
cracks in the WEXTEX region of a non-faulted SG during a SGTR would 
be less than or equal to the primary-to-secondary leakage 
experienced during normal operation. Primary-to-secondary leakage is 
considered in the calculation determining the consequences of a SGTR 
and the value is bounding.
    The postulated SLB has the greatest primary-to-secondary 
pressure differential, and therefore could experience the greatest 
primary-to-secondary leakage. TVA's amendment request requires the 
aggregate leakage, (i.e., the combined leakage for the tubes with 
service induced degradation inside the tubesheet) plus the combined 
leakage developed by other ARC [alternate repair criteria], to 
remain below the maximum allowable SLB primary-to-secondary leakage 
rate limit such that the doses are maintained to less than a 
fraction of the 10 CFR 100 limits and also less than the general 
design criteria (GDC)--19 limits.
    TVA's proposed change also removes the existing axial PWSCC 
[primary water stress corrosion cracking] at dented tube support 
plate ARC and removes the exclusion made for Unit 2 Cycle 12 
operation only from the TS. This ARC was not used on Unit 2 and was 
only intended through the Unit 2 Cycle 12 operation. Therefore, this 
change is inherently more conservative.

[[Page 2900]]

    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.
    TVA's amendment request does not introduce any physical changes 
to the Sequoyah Unit 2 SGs. TVA's amendment request takes credit for 
how the tubesheet enhances the SG tube integrity in the WEXTEX 
region by precluding tube deformation beyond its initial expanded 
outside. Removal of the existing PWSCC axial at dented tube support 
plate ARC incorporates the more conservative TS limit for SG tube 
plugging. A failure to meet SG tube integrity results in an SGTR. 
Because degradation detected within the WEXTEX region are required 
to be plugged, it is highly unlikely that a W* tube would fail as a 
result of a circumferential defect. Therefore a tube severance, 
which would strike neighboring tubes and create a multiple tube 
rupture, is not credible.
    The proposed change does not introduce any new equipment or any 
change to existing equipment. No new effects on existing equipment 
are created.
    Based on the above evaluation, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The amendment request maintains the structural margins of the SG 
tubes for both normal and accident conditions that are required by 
Regulatory Guide 1.121.
    For cracking located within the tubesheet, tube burst is 
precluded due to the presence of the tubesheet. WCAP-14797 defines a 
length, W*, of degradation free expanded tubing that provides the 
necessary resistance to tube pullout due to the pressure induced 
forces (with applicable safety factor applied). Application of the 
W* methodology will preclude unacceptable primary-to-secondary 
leakage during all plant conditions. The methodology for determining 
leakage provides for large margins between calculated and actual 
leakage values in the W* criteria. TVA's proposed change to remove 
PWSCC ARC from the TS does not compromise structural integrity or 
leakage integrity of SG tubes.
    Based on the above, it is concluded that the proposed changes do 
not result in a significant reduction of margin with respect to 
plant safety as defined in the safety analysis report or TSs.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: September 30, 2004.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to relocate the requirements for 
the emergency diesel generator start loss of power instrumentation and 
associated actions in the engineering safety features tables to a new 
limiting conditions for operation (LCO). In addition, an upper 
allowable value has been added to the voltage sensors for loss of 
voltage and degraded voltage consistent with Technical Specification 
Task Force (TSTF) Item TSTF-365 along with a lower allowable value 
limit for the degraded voltage diesel generator start and load shed 
timer. The auxiliary feedwater loss of power start setpoints and 
allowable values have been relocated to this new LCO.
    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 relocation and enhancement of the loss of power functions to 
a new LCO does not alter the intended functions of this feature or 
physically alter these systems. Changes to Avs [allowable values] 
have been evaluated in accordance with TVA [Tennessee Valley 
Authority] setpoint methodology and have been verified to acceptably 
protect the associated safety limits. Format changes provide a 
clearer representation of the requirements and provide more 
consistency with the standard TSs [Technical Specifications] in 
NUREG-1431. The EDG [emergency diesel generator] and AFW [auxiliary 
feedwater] start functions provided by this instrumentation are 
utilized for the mitigation of accident conditions and are not 
considered to be a potential source for accident generation. 
Additionally, these start functions are enhanced by the addition of 
an upper allowable value limit such that the accident mitigation 
functions are not challenged unnecessarily. This further assures the 
ability to mitigate accidents and maintain acceptable offsite dose 
limits. These changes continue to support or improve the required 
safety functions; therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes for the loss of power instrumentation will 
not alter plant processes, components, or operating practices. The 
function to start the EDGs and AFW pumps on a loss of voltage or 
degraded voltage to the shutdown boards will not be altered by the 
proposed change. Additionally, the EDGs and AFW system is not 
considered to be a source for the generation of postulated 
accidents. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter any plant settings or 
functions that are utilized to mitigate accident conditions. The 
enhanced allowable values for the voltage sensors help to prevent 
unnecessary actuation of mitigation systems to ensure their ability 
to respond to actual accident conditions. The parameters that ensure 
the required margin of safety will be maintained with the proposed 
changes or improved. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

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

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: December 2, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in

[[Page 2901]]

place for complying with the requirements of title 10 of the Code of 
Federal Regulations (10 CFR), part 50, Sec.  50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would 
be eliminated, several notes or specific exceptions are revised to 
reflect the related changes to LCO 3.0.4, and Surveillance Requirement 
(SR) 3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated December 2, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

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

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

    Date of amendment request: September 15, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of title 10 of the Code of Federal 
Regulations (10 CFR), part 50, Sec.  50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TSs would be 
eliminated, several notes or specific exceptions are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated September 15, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

[[Page 2902]]

consequences of an accident previously evaluated.

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

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

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

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

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

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

    Date of amendment request: November 8, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 5.9.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.9.4, ``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 November 8, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating letter 
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.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not 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.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

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; Docket Nos. 50-280 and 50-281, Surry Power Station, Units No. 
1 and 2, Surry County, VA

    Date of amendment request: September 8, 2004.
    Description of amendment request: The proposed amendments delete 
the requirements from the technical specifications (TS) to maintain 
hydrogen recombiners (North Anna Power Station only) and hydrogen 
monitors (North Anna and Surry Power Stations). Licensees were 
generally required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI, Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TS for nuclear power reactors currently licensed to operate. The 
revised title 10 of the Code of Federal Regulations (10 CFR), Sec.  
50.44, ``Standards for Combustible Gas Control System in Light-Water-
Cooled Power Reactors,'' eliminated the requirements for hydrogen 
recombiners and relaxed safety classifications and licensee commitments 
to certain design and qualification criteria for hydrogen and oxygen 
monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 8, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

[[Page 2903]]

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the amendment requests 
involve 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: John A. Nakoski.

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; Docket Nos. 5050-280 and 50-281, Surry Power Station, Unit 
No. 1 and No. 2, Surry County, Virginia

    Date of amendment request: December 21, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification requirements for the licensee to submit annual 
occupational radiation exposure reports and 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 December 21, 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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating letter 
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.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not 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.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.

[[Page 2904]]

    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 significance 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: John A. Nakoski.

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

    Date of amendment request: December 13, 2004.
    Description of amendment request: This amendment would revise 
Technical Specification Surveillance Requirement (SR) 3.8.1.7 (fast-
start test), SR 3.8.1.12 (safety injection actuation signal test), SR 
3.8.1.15 (hot restart test), and SR 3.8.1.20 (redundant unit test) to 
clarify what voltage and frequency limits are applicable during the 
transient and steady state portions of the diesel generator (DG) start 
testing performed by these SRs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not affect the DGs ability to supply 
the minimum voltage and frequency within 12 seconds or the steady 
state voltage and frequency. The DGs will continue to perform their 
intended safety function, in accordance with the safety analysis. 
The design of plant equipment is not being modified by the proposed 
change. In addition, the DGs and their associated emergency loads 
are accident mitigating features. As such, testing of the DGs 
themselves is not associated with any potential accident-initiating 
mechanism.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed changes do not increase the types or amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational [or] 
public radiation exposures. The proposed changes are consistent with 
the safety analysis assumptions and resultant consequences.
    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 accident from any accident previously evaluated.
    The proposed change revises surveillance requirements to clarify 
what voltage and frequency limits are applicable during the 
transient and steady state portions of the DG start testing. No 
changes are being made in equipment hardware, operational 
philosophy, testing frequency, system operation, or how the DGs are 
physically tested.
    The proposed changes do not result in a change in the manner in 
which the electrical distribution subsystems provide plant 
protection. The changes do not alter assumptions made in the safety 
analysis. The proposed changes are consistent with the safety 
analysis assumptions and current plant operating practice.
    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.
    The margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed change does not directly affect these barriers, 
nor do they involve any significantly adverse impact on the DGs 
which serve to support these barriers in the event of an accident 
concurrent with a loss of offsite power.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis.
    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.
    NRC Section Chief: Robert A. 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.

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

    Date of application for amendment: June 7, 2004.
    Brief description of amendment: The proposed amendment revised 
Technical Specification 3.9.4, ``Shutdown Cooling (SDC) and Coolant 
Circulation-High Water Level,'' to incorporate the use of an alternate 
cooling method to function as a path for decay heat removal when in 
MODE 6 with the refueling pool fully flooded. The spent fuel pool 
cooling system is the alternative cooling method intended to be used as 
a substitute for the SDC system during the refueling operations, 
including during fuel movement.
    Date of publication of individual notice in Federal Register: 
November 29, 2004 (69 FR 69417).
    Expiration date of individual notice: January 27, 2005.

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 amendment request: November 3, 2004.
    Brief description of amendment request: The proposed amendments 
would revise Technical Specification (TS) 3.7.17 and TS 4.3 for Cycles 
14-16 to allow installation and use of a temporary cask pit spent fuel 
storage

[[Page 2905]]

rack (cask pit rack) for Diablo Canyon Power Plant, Unit Nos. 1 and 2. 
The total spent fuel pool storage capacity for each unit would be 
increased from 1324 fuel assemblies to 1478 fuel assemblies for Cycles 
14-16.
    Date of publication of individual notice in Federal Register: 
December 21, 2004 (69 FR 76486).
    Expiration date of individual notice: February 22, 2005.

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 (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 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of amendment request: August 11, 2004.
    Brief description of amendment: The amendment revises Technical 
Specifications to eliminate operational requirements and certain design 
requirements that will no longer be applicable following the transfer 
of all of the spent fuel from the Haddam Neck Plant spent fuel pool 
into dry cask storage at the Haddam Neck Plant Independent Spent Fuel 
Storage Installation. The amendment relocates administrative 
requirements to the Connecticut Yankee Quality Assurance Program. The 
amendment also deletes the requirement for submittal of an annual 
Occupational Radiation Exposure Report.
    Date of issuance: December 20, 2004.
    Effective date: As of the date that all reactor fuel has been 
permanently removed from the spent fuel pool and stored in an 
Independent Spent Fuel Storage Installation. The license amendment 
shall be implemented within 60 days of its effective date.
    Amendment No.: 201.
    Facility Operating License No. DPR-61: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57978).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation Report, dated December 20, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 12, 2004.
    Brief description of amendment: This amendment approves an 
engineering evaluation performed in accordance with the Pilgrim Nuclear 
Power Station Technical Specifications (TS). TS 3.6.D.3 requires the 
licensee to perform an engineering evaluation when safety relief valve 
(SRV) discharge pipe temperatures exceed 212 [deg]F during normal 
reactor power operation for a period greater than 24 hours, and TS 
3.6.D.4 further requires that power operation may not continue beyond 
90 days from the initial discovery of discharge pipe temperatures in 
excess of 212 [deg]F, without prior NRC approval of the engineering 
evaluation. The Nuclear Regulatory Commission staff has reviewed the 
engineering evaluation and has determined that the licensee has 
adequately justified power operations beyond the end of the TS-required 
90-day period for plant shutdown, until the next cold shutdown of 72 
hours or more.
    Date of issuance: December 23, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 208.
    Facility Operating License No. DPR-35: Amendment does not revise 
the Technical Specifications.
    Date of initial notice in  Federal Register: October 20, 2004 (69 
FR 61695).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 23, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 8, 2003.
    Brief description of amendment: The proposed amendment would delete 
a portion of the Pilgrim Nuclear Power Station (Pilgrim) Technical 
Specification (TS) 4.6.A.2, ``Primary System Boundary--Thermal and 
Pressurization Limitations,'' and the associate TS Table 4.6-3, 
``Reactor Vessel Material Surveillance Program Withdrawal Schedule.'' 
The amendment would replace the existing Reactor Vessel Material 
Surveillance Program with the Boiling Water Reactor Vessel and Internal 
Project (BWRVIP) Integrated Surveillance Program (ISP) and Supplemental 
Surveillance Program (SSP). The BWRVIP ISP/SSP would be incorporated 
into the Pilgrim Updated Final Safety Analysis Report (UFSAR).
    Date of issuance: January 5, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 209.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications and updated the UFSAR.
    Date of initial notice in Federal Register: February 17, 2004 (69 
FR 7521).

[[Page 2906]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 5, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 11, 2003, as supplemented 
January 9, May 3, and July 19, 2004.
    Brief description of amendment: This amendment relocates the 
Technical Specification requirement to leak rate test the enclosure for 
decay heat removal system valves DH-11 and DH-12 to the Technical 
Requirements Manual.
    Date of issuance: December 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 263.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54750).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 27, 2004, as supplement by letter 
dated September 28, 2004.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to lower the reactor vessel water level at which 
the reactor water cleanup system isolates, secondary containment 
isolates, and the control room emergency filter system starts.
    Date of issuance: December 23, 2004.
    Effective date: As of the date of issuance and shall be implemented 
upon startup in Operating Cycle 23.
    Amendment No.: 209.
    Facility Operating License No. DPR-46: Amendment revised the TS.
    Date of initial notice in Federal Register: June 22, 2004 (69 FR 
34702).
    The supplement dated September 28, 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 as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 23, 2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: December 23, 2003.
    Brief description of amendments: The amendments modified TS 
requirements to adopt the provisions of Industry/TS Task Force (TSTF) 
change TSTF-359, ``Increased Flexibility in Mode Restraints.'' The 
availability of TSTF-359 for adoption by licensees was announced in the 
Federal Register on April 4, 2003 (68 FR 16579).
    Date of issuance: December 22, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 215, 220.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 16, 2004 (69 
FR 55844)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 22, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 4, 2003, as supplemented 
by letters dated December 24, 2003, and June 3, August 24, and October 
6 and 22, 2004.
    Brief description of amendments: The proposed amendments would 
revise Technical Specification 3.9.3, ``Containment Penetrations,'' by 
adding a note to the limiting condition for operation that permits the 
containment equipment hatch to be open during core alterations and 
movement of irradiated fuel in containment during refueling operations.
    Date of issuance: December 23, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 193/184.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54752). The supplemental letters dated December 24, 2003, and June 
3, August 24, October 6, and October 22, 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 December 23, 2004.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket No. 50-498, South Texas Project, 
Unit 1, Matagorda County, Texas

    Date of amendment request: September 30, 2004.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Surveillance Requirement 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: December 28, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 1--166.
    Facility Operating License No. NPF-76: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 2004 (69 FR 
62477).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment requests: September 22, 2003, and September 27, 
2004.
    Brief description of amendments: The amendments change Technical 
Specification (TS) Surveillance Requirement 4.7.1.6, ``Atmospheric 
Steam Relief Valves'' to provide consistency with TS 3.3.5.1,

[[Page 2907]]

``Atmospheric Steam Relief Valve Instrumentation,'' regarding 
atmospheric steam relief valve automatic controls. The amendments also 
correct typographical errors in TSs 3.7.1.6 and 3.2.4. The remaining 
proposed changes associated with the September 22, 2003, application 
were withdrawn as noted in the NRC staff's letter to the licensee dated 
October 19, 2004.
    Date of issuance: December 28, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: Unit 1--167; Unit 2--156.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64139) for the September 22, 2003, application and October 26, 2004 
(69 FR 62478) for the September 27, 2004, application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 10th day of January, 2005.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 05-779 Filed 1-14-05; 8:45 am]
BILLING CODE 7590-01-P