Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 12743-12755 [05-4792]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 49 (Tuesday, March 15, 2005)]
[Notices]
[Pages 12743-12755]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-4792]


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

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 February 18, 2005, through March 3, 2005. 
The last biweekly notice was published on March 1, 2005 (70 FR 9986).

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

[[Page 12744]]

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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
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 pdr@nrc.gov.

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

    Date of amendment request: February 2, 2005.
    Description of amendment request: The amendment would revise Tables 
3.1.1 and 4.1.1 of the Technical Specifications (TSs) to incorporate 
the isolation trip setting and the instrumentation surveillance 
requirements of the reactor water cleanup (RWCU) system high energy 
line break (HELB) detection and isolation equipment.
    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's analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The equipment modification, which is the subject of the 
proposed amendment, had been installed by the licensee in 1998 using 
the provisions of 10 CFR 50.59, and the licensee had been performing 
the surveillance requirement as is now proposed for this amendment. The 
purpose of the modification was to

[[Page 12745]]

ensure that the RWCU system can be isolated on an HELB downstream of 
the RWCU system isolation valves. The proposed addition of the RWCU 
HELB detection/isolation equipment setpoints and surveillance 
requirements to the TSs satisfies the 10 CFR 50.36 requirements for 
limiting conditions for operation (LCOs) and surveillance requirements 
(SRs) that should be included in the TSs. Thus, the proposed amendment 
would not alter the physical design or operational procedures 
associated with any plant structure, system, or component (i.e., the 
RWCU system will be isolated by existing equipment, in case there is an 
HELB, in the same way as before the amendment). Consequently, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The second standard requires that operation of the unit 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 does not lead to any changes in the 
physical design, safety limits, or safety analysis assumptions 
associated with the operation of the plant. The proposed amendment 
would only add requirements to the TSs for the operability and 
surveillance testing of the RWCU system HELB detection/isolation 
equipment. Accordingly, the proposed amendment does not introduce any 
new accident initiators, nor does it reduce or adversely affect the 
capabilities of any plant structure or system in the performance of 
their safety function. Therefore, the proposed amendment does not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. The proposed amendment will not affect 
any margin of safety as defined in the Oyster Creek Nuclear Generating 
Station Final Safety Analysis Report. The amendment only adds LCOs and 
SRs to assure that the RWCU system HELB detection/isolation equipment 
is operable under the plant operating conditions when an RWCU system 
HELB is possible. The amendment does not change the RWCU system 
isolation time as compared to original plant design. Therefore, the 
proposed amendment does not involve a significant reduction in a margin 
of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Richard J. Laufer.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendment request: January 27, 2005.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) testing frequency for the 
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' 
The proposed change would revise the test frequency of SR 3.1.4.2, 
control rod scram time testing, from ``120 days cumulative operation in 
MODE 1'' to ``200 days cumulative operation in MODE 1.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in licensing amendment applications in the Federal Register on August 
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the 
model NSHC determination in its application dated January 27, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: February 14, 2005.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Surveillance Requirement (SR) 3.3.7.1 to 
extend the frequency of the channel functional test to once every 31 
days to once every 92 days.
    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 LAR [license amendment request] extends the current 
31 day surveillance frequency to a 92 day surveillance frequency. 
The proposed LAR does not alter the method of operating or 
configuration for any structure, system, or component. Extension of 
the surveillance interval will not affect any accident analysis or 
the plant safety system response to the accident. The extension of 
the surveillance interval will not affect the ability of ES 
[engineered safeguards] to actuate Engineered Safeguards Protective 
System (ESPS) equipment. Therefore, the proposed LAR does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:

[[Page 12746]]

    The proposed change does not necessitate a change in parameters 
governing plant operation. Consequently, the proposed LAR does not 
alter the nature of events postulated in the UFSAR [Updated Final 
Safety Analysis Report] nor does the LAR introduce any unique 
precursor mechanisms. Therefore, the proposed amendment will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Involve a significant reduction in the margin of Safety
    The proposed change does not adversely affect any plant safety 
limits, setpoints, or design parameters. The changes will not 
adversely affect the fuel, fuel cladding, RCS [reactor coolant 
system], or containment integrity. The proposed change to the 
frequency for SR [surveillance requirement] 3.3.7.1 will not impact 
the operation of the ESPS Digital Automatic Actuation Logic Channels 
nor the actuation of ESPS equipment. Additionally, the channel 
functional testing of the ESPS Digital Channels will continue to be 
performed within an acceptable timeframe following implementation of 
the proposed change. As such, 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: Anne W. Cottingham, Winston and Strawn LPP, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: December 22, 2004.
    Description of amendment request: The requested change would delete 
Technical Specification (TS) 6.9.1.5, ``Occupational Radiation Exposure 
Report,'' and 6.9.1.6, ``Monthly Operating Reports,'' as described in 
the Notice of Availability published in the Federal Register on June 
23, 2004 (69 FR 35067).
    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.

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

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

    Date of amendment request: January 27, 2005.
    Description of amendment request: The proposed License Amendment 
Request (LAR) would allow the licensee to utilize a probabilistic 
methodology to determine the contribution to main streamline break 
(MSLB) leakage rates for the once-through steam generator (OTSG) from 
the tube end crack (TEC) alternate repair criteria (ARC) described in 
Crystal River Unit 3 (CR-3) Improved Technical Specification (ITS) 
5.6.2.10.2.f. This LAR involves no change to the CR-3 ITS.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This LAR proposes to change the method to determine the 
projected MSLB leakage rates for TEC. Potential leakage from OTSG 
tubes, including leakage contribution from TEC, is bounded by the 
MSLB evaluation presented in the Final Safety Analysis Report 
(FSAR). The inspection required by the ARC will continue being 
performed as required by CR-3 ITS 5.6.2.10. This inspection provides 
continuous monitoring of tubes with TEC indications remaining in 
service, and ensures that degradation of new tubes containing TEC 
indications is detected. The proposed change in method to determine 
MSLB leakage rates for TEC does not change any accident initiators.
    Therefore, granting this LAR 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.
    This LAR proposes to change the method to determine the 
projected MSLB leakage rates for TEC. The change introduces no new 
failure modes or accident scenarios. The proposed change does not 
change the assumptions made in Topical Report BAW-2346P, Revision 0, 
which demonstrated structural and leakage integrity for all normal 
operating and accident conditions for CR-3. The design and 
operational characteristics of the OTSGs are not impacted by the use 
of a probabilistic methodology to determine MSLB leakage rates.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    This LAR proposes to change the method to determine the 
projected MSLB leakage rates for TEC. The resulting leakage 
estimates will be lower than the estimates from the old method. 
However, the estimates from the proposed method will be more 
realistic and do not impact the acceptance criteria. The methodology 
relies on the same accident analyses described in Topical Report 
BAW-2346P, Revision 0, and License Amendment Request 249, 
Revision 0, and utilizes the same leakage test data and leakage 
limit. The FSAR analyzed accident scenarios are not affected by the 
change and remain bounding. The limits established in CR-3 ITS 
3.4.12, and 5.6.2.10.2.f have not been changed. Therefore, the 
proposed change does not reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 12747]]

review, it appears that the three standards of 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: January 20, 2005.
    Description of amendment request: The submittal requests revision 
to several Technical Specifications (TSs) using eight TS Task Force 
(TSTF) generic changes. The eight TSTFs (nos. 5, 93, 95, 101, 258, 299, 
308, and 361) delete redundant safety limit violation notification 
requirements; extend the pressurizer heater surveillance frequency from 
92 days to 18 months; extend the completion time for reducing the Power 
Range High trip setpoint from 8 to 72 hours; change the auxiliary 
feedwater pump test frequency to be consistent with the inservice test 
program frequency; remove redundant requirements and add other 
requirements to Section 5.0, Administrative Controls; clarify the 
requirements regarding the frequency of testing for cumulative and 
projected dose contributions from radioactive effluents; and add a note 
to the residual heat removal requirements during Mode 6 low water level 
operations that allows one required residual heat removal (RHR) loop to 
be inoperable for up to 2 hours for surveillance testing provided the 
other RHR loop is operable and in operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes revise administrative requirements, 
actions, action times, surveillance requirements and surveillance 
frequencies. The revised requirements are not an initiator of any 
accident previously evaluated. As a result, the probability of any 
accident previously evaluated is not significantly increased by the 
proposed changes. The Technical Specifications continue to require 
the systems, structures, and components associated with the revised 
requirements to be operable. Therefore, any mitigation functions 
assumed in the accident analyses will continue to be performed. As a 
result, the consequences of any accident previously evaluated are 
not significantly increased. Therefore, the proposed amendments do 
not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed changes do not alter the design or physical 
configuration of the plant. No changes are being made to the plant 
that would introduce any new accident causal mechanisms. The 
proposed changes do not affect any other plant equipment. Therefore, 
operation of the facility in accordance with the proposed amendments 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes do not change the design or function of 
plant equipment. The proposed changes do not significantly reduce 
the level of assurance that any associated plant equipment will be 
available to perform its function. The proposed changes provide 
operating flexibility without significantly affecting plant 
operation. Therefore, operation of the facility in accordance with 
the proposed amendments would not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(copyright)) 
are satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Michael L. Marshall, Jr.

National Aeronautics and Space Administration, Docket No. 50-30, the 
Plum Brook Test Reactor, Sandusky, Ohio

    Date of amendment request: January 14, 2005.
    Description of amendment request: The proposed amendment will 
clarify the license requirements for confirmation of Final Status 
Survey results prior to backfilling or covering of excavated areas. The 
amendment will allow performance of the Final Status Survey for an area 
that has been excavated and allow backfilling of the area without the 
performance of confirmatory surveys by the NRC. Backfilling without the 
performance of a confirmatory survey would be allowed only when the NRC 
Staff has determined that there is appropriate safety or technical 
justification for backfilling and that, based on the NRC Staff review 
of the completed Final Status Survey for the affected area, there is 
reasonable assurance that the Licensee's surveys have demonstrated that 
the affected area satisfies the unrestricted release criteria.
    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:

    The proposed amendment to License TR-3 is necessary to assure 
that, in limited situations, areas excavated for the performance of 
dismantlement and remediation activities do not result in 
unnecessary industrial safety hazards. The proposed changes do not 
involve a significant hazard as shown in the following:
    A. The proposed amendment to License TR-3 does not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    The accident scenarios applicable to the decommissioning of the 
Plum Brook Reactor Facility are described in section 3.3 of the 
Decommissioning Plan for the Plum Brook Reactor. The Decommissioning 
Plan describes postulated events that could result in a release of 
radioactive materials from the site and analyzes the radiation dose 
consequences of these events and demonstrates that no adverse public 
health and safety impacts are expected from these events. 
Performance of Final Status Surveys is a continuation of 
decommissioning and is an activity that involves measurements and 
analysis of residual radioactivity in areas in which decommissioning 
has already been performed. It is a process used to confirm that 
radioactivity has been removed to achieve the acceptance criteria 
specified in 10 CFR 20, Subpart E. These surveys are subjected to 
NRC review, and in some instances NRC confirmatory surveys. These 
surveys are performed in areas where other decommissioning 
activities are already complete and where there is no credible event 
that could initiate the analyzed accidents. Backfilling an excavated 
area, or otherwise rendering it inaccessible, will have no impact on 
other decommissioning activities, or on postulated accidents from 
other decommissioning activities. Therefore, the proposed amendment 
will have no affect on the probability or consequences of accidents 
previously analyzed.
    B. The proposed amendment to License TR-3 will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Accidents previously analyzed in the Decommissioning Plan assess 
different

[[Page 12748]]

scenarios that could cause the dispersion of radioactive material to 
the environment. These scenarios arise from dismantlement activities 
associated with the decommissioning. Excavation of areas to support 
dismantlement and remediation is an activity that is described in 
the Decommissioning Plan. Excavated areas will be surveyed in 
accordance with the Final Status Survey Plan to verify that they 
have been radiologically remediated to the unrestricted release 
criteria. These areas may be backfilled without the performance of 
confirmatory surveys performed by the NRC or their Contractor. 
However, this will only be permitted after NRC review of the 
Licensee's survey results and a determination that there is 
reasonable assurance that no residual radioactivity in excess of the 
release criteria remains. Therefore, there will be adequate 
verification by the regulatory agency that there is reasonable 
assurance that there is no potential for the dispersion of 
radioactive material to the environment from the backfilled area. 
The methods and processes used for control of work activities and 
for control and monitoring of radioactivity will remain the same as 
those used prior to this amendment. Therefore, no new or different 
types of accidents are created by this proposed amendment.
    C. The proposed amendment to License TR-3 will not involve a 
significant reduction in a margin of safety.
    As discussed previously, the activities that will be performed 
at the facility are as previously described and evaluated in the 
accident analyses presented in the Decommissioning Plan. The 
radiological criteria to be used in applying for termination of the 
NRC Licenses will remain the same as originally proposed and are 
consistent with the criteria of 10 CFR 20 Subpart E. The results of 
Final Status Surveys performed by the Licensee will remain subject 
to review by the U.S. NRC for adequate implementation of the Final 
Status Survey Plan. Therefore, the margins of safety applicable to 
assessing the long term dose to members of the public from exposure 
to the facility after termination of the license remain unchanged. 
In addition, since this amendment does not impact any previously 
reviewed accident analyses as previously discussed, no margins of 
safety are affected by this proposed amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for the Licensee: J. William Sikora, Esquire, 21000 
Brookpark Road, Mail Stop 500-118, Cleveland, Ohio 44135.
    NRC Section Chief: Patrick M. Madden.

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

    Date of amendment request: October 15, 2004.
    Description of amendment request: The proposed amendment revises 
Technical Specification 5.5.6, ``Reactor Coolant Pump Flywheel 
Inspection Program,'' to extend the allowable inspection interval to 20 
years.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
availability of a model safety evaluation and model no significant 
hazards consideration (NSHC) determination for referencing in license 
amendment applications in the Federal Register on October 22, 2003 (68 
FR 60422). The licensee affirmed the applicability of the model NSHC 
determination in its application dated October 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 to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    The proposed change does not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, or configuration of the facility, or the manner in which 
the plant is operated and maintained; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves 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 Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: February 1, 2005.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.7.17, ``Spent Fuel Pool 
Storage,'' Technical Specification 4.3, ``Fuel

[[Page 12749]]

Storage'' and the corresponding TS bases, using revised spent fuel pool 
(SFP) criticality analysis methodology which takes credit for soluble 
boron in the spent fuel pool.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment proposes to revise the plant licensing 
basis by: (1) Replacing the spent fuel pool criticality analyses; 
and (2) revising the spent fuel storage Technical Specifications 
3.7.17, ``Spent Fuel Pool Storage'' and 4.3, ``Fuel Storage'' 
utilizing the proposed analyses. The proposed Technical 
Specification revisions allow spent fuel to be stored in different 
configurations.
    The proposed changes relate to prevention of criticality 
accidents in the spent fuel pool. Since the current spent fuel pool 
criticality analyses and Technical Specifications ensure that a 
criticality accident does not occur, criticality accidents have not 
been previously evaluated. Likewise the proposed spent fuel pool 
criticality analyses and Technical Specifications ensure that a 
criticality accident does not occur. Thus the proposed changes do 
not involve a significant increase in the consequences of an 
accident previously evaluated.
    Events that could cause a criticality accident were evaluated 
and analyses demonstrated that the current Technical Specification 
required soluble boron is more than adequate to assure that a 
criticality accident does not occur. Thus the proposed changes do 
not involve a significant increase in the probability of an accident 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This license amendment proposes to revise the plant licensing 
basis by: (1) Replacing the spent fuel pool criticality analyses; 
and (2) revising the spent fuel storage Technical Specifications 
3.7.17, ``Spent Fuel Pool Storage'' and 4.3, ``Fuel Storage'' 
utilizing the proposed analyses. The proposed Technical 
Specification revisions allow spent fuel to be stored in different 
configurations.
    The proposed licensing basis changes do not involve a change in 
system operation, or procedures involved with the fuel storage 
system. It does revise the allowable storage configurations. The 
proposed changes provide a conservative basis for evaluating spent 
fuel pool criticality and storage of fuel assemblies in a safe 
configuration which meets criticality evaluation acceptance 
criteria. There are no new failure modes or mechanisms created 
through use of the proposed analyses or proposed Technical 
Specifications. Use of these licensing basis changes for storage of 
fuel assemblies does not involve any modification in the operational 
limits of plant systems. There are no new accident precursors 
generated with use of these licensing basis changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This license amendment proposes to revise the plant licensing 
basis by: (1) Replacing the spent fuel pool criticality analyses; 
and (2) revising the spent fuel storage Technical Specifications 
3.7.17, ``Spent Fuel Pool Storage'' and 4.3, ``Fuel Storage'' 
utilizing the proposed analyses. The proposed Technical 
Specification revisions allow spent fuel to be stored in different 
configurations.
    The proposed licensing basis change will result in a 
conservative calculation of the required spent fuel pool soluble 
boron concentration for the proposed fuel storage configurations. 
The current Technical Specification required spent fuel pool boron 
concentration significantly exceeds the proposed criticality 
analyses required boron concentration. The proposed analyses 
demonstrate that the criticality analysis acceptance criteria for 
the proposed fuel storage configurations are met. The proposed 
analyses utilize industry accepted analysis codes which have been 
benchmarked for the spent fuel pool criticality analyses proposed 
for the Prairie Island Nuclear Generating Plant. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

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

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

    Date of amendment request: October 1, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) to maintain 
hydrogen recombiners and hydrogen and oxygen monitors. A notice of 
availability for this TS improvement using the consolidated line item 
improvement process was published in the Federal Register on September 
25, 2003 (68 FR 55416). 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 in 1979. Requirements 
related to combustible gas control were imposed by order for many 
facilities and were added to, or included in, the TSs for nuclear power 
reactors currently licensed to operate. The revised Title 10 of the 
Code of Federal Regulations (10 CFR) Section 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 Nuclear Regulatory Commission (NRC) staff issued a notice of 
availability of a model no significant hazards consideration (NSHC) 
determination for referencing 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 October 1, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

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

[[Page 12750]]

    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen 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 and oxygen 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 NRC 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. Also, as part of the 
rulemaking to revise 10 CFR 50.44, the NRC found that Category 2, as 
defined in RG 1.97, is an appropriate categorization for the oxygen 
monitors, because the monitors are required to verify the status of 
the inert containment.
    The regulatory requirements for the hydrogen and oxygen 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, 
[classification of the oxygen monitors as Category 2,] and removal 
of the hydrogen and oxygen monitors from TSs will not prevent an 
accident management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TSs, 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 and oxygen monitor requirements, 
including removal of these requirements from TSs, will not result in 
any failure mode not previously analyzed.
    The hydrogen recombiner and hydrogen and oxygen monitor 
equipment was intended to mitigate a design-basis hydrogen release. 
The hydrogen recombiner and hydrogen and oxygen 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 and oxygen monitor requirements, 
including removal of these requirements from TSs, 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 NRC 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.
    Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TSs will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: Darrell Roberts.

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

    Date of amendment request: February 4, 2005.
    Description of amendment request: The proposed change will relocate 
Technical Specification (TS) requirements regarding the Traversing In-
core Probe (TIP) System to the Hope Creek Updated Final Safety Analysis 
Report (UFSAR). Additionally, the associated TS Bases would be deleted. 
Formatting changes would also be made, as required, in order to 
incorporate these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will relocate the requirements from the 
Technical Specifications (TS) to the Hope Creek Updated Final Safety 
Analysis Report (UFSAR) a licensee-controlled document. The 
relocated requirements will be retained in licensee-controlled 
documents, which will be maintained under the requirements of the 
provisions of 10 CFR 50.59. Since any changes to licensee-controlled 
documents are required to be evaluated per 10 CFR 50.59, no increase 
in the probability or consequences of an accident previously 
evaluated is allowed.
    In addition, the proposed change will not affect any equipment 
important to safety, in structure or operation. This change will not 
alter operation of process variables[,] Structures, systems or 
components as described in the UFSAR or licensing basis. Therefore, 
the proposed change does not involve a significant increase in the 
probability or radiological 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 change the design function or 
operation of any plant equipment. No new failure mechanisms, 
malfunctions, or accident initiators are being introduced by the 
proposed changes. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not reduce the margin of safety since 
they have no impact on safety analysis assumptions. Any future 
changes to the TIP system requirements will be evaluated under 
10CFR50.59. 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

[[Page 12751]]

    NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: October 6, 2004.
    Brief description of amendments: The proposed amendments will (1) 
Revise Technical Specification (TS) 3.8.1, ``AC Sources-Operating,'' to 
allow surveillance requirement (SR) testing of the onsite standby 
diesel generators (DGs) during power operation, by removing specific 
surveillance test MODE restrictions, (2) incorporate changes based on 
industry approval TSs Task Force (TSTF) standard TS change traveler, 
TSTF-283, Revision 3, (3) add a new note to TS 3.8.1 Limiting Condition 
for Operation (LCO) that permits one DG to be connected in parallel 
with offsite power in order to conduct the required surveillance 
testing, and (4) delete the expired TS LCO 3.8.1, Required Action A.3 
one time 21 day completion time allowance for Startup Transformer XST2 
preventive maintenance.
    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 by focusing on the three standards set forth in 10 CFR 
50.92. The licensee's 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 design of plant equipment is not being modified by the 
proposed changes. In addition, the DGs and their associated 
emergency loads are accident mitigating features. As such, testing 
of the diesel generators (DGs) themselves is not associated with any 
potential accident-initiating mechanism. Therefore, there will be no 
significant impact on any accident probabilities by the approval of 
the requested changes.
    The changes include an increase in the online time that a DG 
under test will be paralleled to the grid (for SRs 3.8.1.10 and 
3.8.1.14) or unavailable due to testing (per SR 3.8.1.13). However, 
the overall time that the DG is paralleled in all modes (outage /
non-outage) should remain unchanged. As such, the ability of the 
tested DG to respond to a design basis accident [(DBA)] could be 
adversely impacted by the proposed changes. However, the impacts are 
not considered significant based, in part, on the ability of the 
remaining DG to mitigate a DBA or provide safe shutdown. With regard 
to SR 3.8.1.10 and SR 3.8.1.14, experience shows that testing per 
these SRs typically does not perturb the electrical distribution 
system and share[s] the same electrical configuration alignment as 
the current monthly surveillance. In addition, operating experience 
and qualitative evaluation of the probability of the DG or bus loads 
being adversely affected concurrent with or due to a