Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 398-408 [05-2]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 2 (Tuesday, January 4, 2005)]
[Notices]
[Pages 398-408]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-2]


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

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 10, 2004, through December 22, 
2004. The last biweekly notice was published on December 21, 2004 (69 
FR 76486).

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

[[Page 399]]

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, 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 basis 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,

[[Page 400]]

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.

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 amendment request: July 20, 2004.
    Description of amendment request: The proposed administrative 
amendment corrects references in Technical Specification (TS) 5.6.7 and 
in TS Table 3.3.10-1, and deletes reference to hydrogen analyzers which 
were removed from the TSs by Amendment Nos. 262 and 239, for Unit Nos. 
1 and 2, respectively, on March 2, 2004.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Amendment Nos. 262 and 239 were approved and issued by the 
Nuclear Regulatory Commission (NRC) on March 2, 2004. These 
amendments removed the requirements for the containment hydrogen 
recombiners and the hydrogen analyzers as equipment required to 
control hydrogen in the Containment. The amendments required the 
hydrogen analyzers to be retained as non-safety-related equipment to 
record hydrogen concentrations in beyond design-basis accidents. The 
request to remove hydrogen control from the design basis included a 
mark-up of proposed Technical Specification changes. However, 
related changes to Technical Specification Table 3.3.10-1, Technical 
Specification 5.6.7, and Technical Specification 3.8.1 were not 
included in the markup. Therefore, we are requesting an 
administrative change to correct this oversight.
    Since the justification for these changes has been approved in 
Calvert Cliffs Amendment Nos. 262 and 239, there is no technical or 
safety issue associated with this request.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed administrative amendment corrects references in a 
Technical Specification table and in a Technical Specification, and 
deletes reference to hydrogen analyzers. Since the justification for 
these changes has been approved in Calvert Cliffs Amendment Nos. 262 
and 239, there is no technical or safety issue associated with this 
request. This request does not involve a change in the operation of 
the plant, and no new accident initiation mechanism is created by 
the proposed change, nor does the change involve a physical 
alteration of the plant.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind] of accident from any accident 
previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    Amendment Nos. 262 and 239 were approved and issued by the 
Nuclear Regulatory Commission (NRC) on March 2, 2004. These 
amendments removed the requirements for the containment hydrogen 
recombiners and the hydrogen analyzers as equipment required to 
control hydrogen in the Containment. The amendments required the 
hydrogen analyzers to be retained as non-safety-related equipment to 
record hydrogen concentrations in beyond design-basis accidents. The 
request to remove hydrogen control from the design basis included a 
mark-up of proposed Technical Specification changes. However, 
related changes to Technical Specification Table 3.3.10-1, Technical 
Specification 5.6.7, and Technical Specification 3.8.1 were not 
included in the markup. Therefore, we are requesting an 
administrative change to correct this oversight.
    Because the hydrogen analyzers were removed from the Technical 
Specifications by Amendment Nos. 262 and 239, no margin of safety is 
impacted by the proposed administrative changes.
    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: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

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 amendment request: August 3, 2004.
    Description of amendment request: The proposed amendment would 
extend the surveillance requirement (SR) 3.3.3.1 test interval for 
reactor trip circuit breakers from 31 to 92 days and impose a staggered 
test interval consistent with SR 3.3.3.2.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The reactor trip circuit breakers (RTCB) are part of the Reactor 
Protective System (RPS). The RPS initiates a reactor trip to protect 
against violating the core specified acceptable fuel design limits 
and reactor coolant pressure boundary integrity during anticipated 
operational occurrences. By opening the RTCBs to trip the reactor, 
the RPS also assists the engineered safety features systems in 
mitigating accidents. All of the accident analyses that call for a 
reactor trip assume that the RTCBs operate and interrupt power to 
the control element drive mechanisms. The proposed testing interval 
will result in less wear on the RTCBs and, thereby, increase breaker 
reliability.
    The RTCBs are accident mitigators and do not affect the 
probability of an accident.
    Topical Report CE NPSD-951-A shows only one failure up to 1993 
in the plants studied. Calvert Cliffs' surveillance records show no 
failures from 1994 to 2003. This data demonstrates that the 
consequences of an accident will not be significantly

[[Page 401]]

increased by extending the surveillance interval and imposing a 
staggered test interval.
    Therefore, extending the surveillance interval and imposing a 
staggered test interval does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    There is no change in plant equipment or operation related to 
this license amendment request. The RTCBs are accident mitigators 
and extending the surveillance interval and imposing a staggered 
test interval does not adversely affect their operation.
    Therefore, the proposed amendment does not create the 
possibility of a new or different [kind] of accident from any 
accident previously evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The margin of safety in this case is the reliance on the RTCBs 
to open on a signal from the RPS. Extending the surveillance 
frequency and imposing a staggered test interval results in a test 
every six weeks as opposed to the current monthly test. The new 
interval will result in less wear on the RTCBs, thereby improving 
the margin of safety.
    Therefore, extending the surveillance interval and imposing a 
staggered test interval 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: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: December 17, 2004.
    Description of amendment request: The proposed change will revise 
the Technical Specification (TS) requirements for direct current (DC) 
sources. The current TS only includes Action Statements for an 
inoperable DC Power subsystem. The proposed change will add a new 
Action Statement to TS 3.8.4, ``DC Sources--Operating,'' to 
specifically address an inoperable battery charger.
    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 class 1E direct current (DC) electrical power system 
including the associated battery chargers are not initiators to any 
accident sequence analyzed in the Updated Safety Analysis Report 
(USAR). Operation in accordance with the proposed Technical 
Specification (TS) ensures that the DC system is capable of 
performing its function described in the USAR. While power to the 
non class 1E charger will be lost after a Design Basis Accident 
(DBA), the Division 1 and 2 batteries have the ability to supply all 
DBA loads and all other standby loads not automatically tripped on a 
LOCA [Loss of Coolant Accident] signal for 4 hours and have 
sufficient capacity to restore normal AC [alternating current] and 
DC power with the charger inoperable. The actions required to 
restore the power to the non-class 1E charger are included in the 
procedures for Station Blackout requiring the use of a non class 1E 
diesel generator. They allow the impacted DC battery and DC bus to 
be restored to perform its required function as described in the 
USAR.
    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 do not involve a physical change to the 
plant. No new equipment is being introduced, and installed equipment 
is not being operated in a new or different manner. There are no 
setpoints, at which protective or mitigative actions are initiated, 
affected by this change. These changes will not alter the manner in 
which equipment operation is initiated, nor will the function 
demands on credited equipment be changed. Any alterations in 
procedures will continue to assure that the plant remains within 
analyzed limits, and no change is being made to the procedures 
relied upon to respond to an off normal event as described in the 
USAR. As such, no new failures modes are being introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes are acceptable because the 
operability of the safety related DC systems are unaffected and 
there is no detrimental impact on any equipment design parameter. 
The plant will still be capable of operating within assumed 
conditions. Operations in accordance with the proposed TS ensures 
that the DC system is capable of performing its function as 
described in the USAR; therefore, the support of the DC system to 
the plant response to analyzed events will continue to provide the 
margins of safety assumed by the analysis. In addition, the DC 
system is within the scope of 10 CFR 50.65, ``Requirements for 
monitoring the effectiveness of maintenance at nuclear power 
plants,'' which will ensure the control of maintenance activities 
associated with the DC system. This provides sufficient management 
control of the requirements that assure the batteries are maintained 
in a highly reliable condition. The non-class 1E battery charger is 
the same model and has the same ratings as the installed Division 1 
and 2 class 1E battery chargers (i.e., same input loading and ampere 
current capability), and was purchased to Class 1E requirements. In 
addition, the backup battery charger can be powered from an onsite 
power source (Station Blackout (SBO) diesel generator) should it be 
required.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: August 31, 2004.
    Description of amendment request: The proposed amendment would 
modify the existing Technical Specification (TS) 3.4.1, ``Recirculation 
Loops Operating,'' associated with single recirculation loop operation 
by incorporating limits for the linear heat generation rate (LHGR) fuel 
thermal limit into the limiting condition of operation (LCO). 
Currently, TS 3.4.1 only contains thermal limits for the minimum 
critical power ratio and the average planar LHGR. Thermal limits 
associated with the two recirculation operations are contained in TS 
3.2.1, ``Average Planar Linear Heat Generation Rate (APLHGR),'' TS 
3.2.2, ``Minimum Critical Power Ratio (MCPR),'' and TS 3.2.3, ``Linear 
Heat Generation Rate (LHGR).'' The proposed TS change will reflect a 
consistency with the existing two recirculation loop LCOs by including 
the same three thermal limits into the single recirculation loop LCO.

[[Page 402]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR Section 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 LHGR is a measure of the heat generation rate of a fuel rod 
in a fuel assembly at any axial location. Limits on the LHGR are 
specified to ensure that fuel design limits are not exceeded 
anywhere in the core during normal operation, including anticipated 
operational occurrences (AOOs). Additionally, the LHGR limits 
provide assurance the fuel peak cladding temperature (PCT) during a 
Loss Of Coolant Accident (LOCA) will not exceed the requirements of 
10 CFR 50.46.
    The PNPP [Perry Nuclear Power Plant] Core Monitor previously 
automatically modified the ``composite'' LOCA/Thermal-Mechanical 
MAPLHGR [minimum average planar linear heat generation rate] limits 
for single recirculation loop operation. As a result, the LHGR limit 
was adjusted for single recirculation loop operation by application 
of the single recirculation loop operation MAPLHGR multiplier to the 
``composite'' MAPLHGR limits. The proposed TS change establishes a 
TS requirement for LHGR limits to be modified, as specified in the 
Core Operating Limits Report, during single recirculation loop 
operation. This TS requirement provides assurance that the fuel 
design limits will remain satisfied during the time the plant may be 
in single recirculation loop operation.
    There are no physical modifications being made to any plant 
system or component, including the fuel.
    The manual versus automatic adjustment of the LHGR limits when 
in single reactor loop operation is considered a change in the 
implementation of a core monitoring function. However, since the 
LHGR limits that will be applied to the core are consistent with the 
NRC-approved fuel design and LOCA methodologies in use at PNPP, this 
change in monitoring implementation is not considered significant.
    Therefore, since no significant changes are being made to the 
plant or its operation, the probability or the consequences of an 
accident have not increased over those previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no physical modifications being made to any plant 
system or component, including the fuel. The manual versus automatic 
adjustment of the LHGR limits when in single reactor loop operation 
is considered a change in the implementation of a core monitoring 
function. However, since the LHGR limits that will be applied to the 
core are consistent with the NRC-approved fuel design and LOCA 
methodologies in use at PNPP, this change in monitoring 
implementation is not considered significant. The proposed TS change 
provides assurance that the LHGR limits will be adjusted if the 
plant enters a condition of single recirculation loop operation, 
thereby ensuring the fuel design limits remain satisfied.
    Therefore, the proposed change does 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.
    There are no physical modifications being made to any plant 
system or component, including the fuel. The manual versus automatic 
adjustment of the LHGR limits when in single reactor loop operation 
is considered a change in the implementation of a core monitoring 
function. However, since the LHGR limits that will be applied to the 
core are consistent with the NRC-approved fuel design and LOCA 
methodologies in use at PNPP, this change in monitoring 
implementation is not considered significant. The proposed TS change 
provides assurance that the LHGR limits will be adjusted if the 
plant enters a condition of single recirculation loop operation, 
thereby ensuring the fuel design limits remain satisfied.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: October 25, 2004.
    Description of amendment request: The proposed amendment would 
revise the required channels per trip system for several instrument 
functions contained in technical specification tables 3.3.6.1-1 
(Primary Containment Isolation Instrumentation), 3.3.6.2-1 (Secondary 
Containment Isolation Instrumentation), and 3.3.7.1-1 (Control Room 
Emergency Filter System 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, 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.
    Revising the Required Channels Per Trip System to conform with 
the Cooper Nuclear Station (CNS) design basis resolves an 
inconsistency that will not result in any changes to instrumentation 
configuration, operating practices, or means of testing. Thus, these 
changes are administrative and have no associated effects on the 
probability or consequences of previously evaluated accidents.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes represent administrative changes to the 
Technical Specification controls over the affected instrumentation. 
Thus, the changes will not create new event initiators or alter 
plant response to postulated plant events.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed changes have no effect on the manner in which the 
affected instruments are configured, operated, or tested. Similarly, 
there is no relaxation in the application of Technical 
Specifications to inoperable channels. Thus these proposed changes 
will not result in 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: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Acting Section Chief: Michael K. Webb.

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: September 23, 2004.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, 
Starting Air, and Turbocharger Air Assist,'' to increase the required 
amount of stored diesel fuel to support use of low-sulfur fuel oil 
required by the California Air Resources Board (CARB).
    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

[[Page 403]]

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 revises the minimum amount of stored diesel 
fuel. The change is required to support the use of California Air 
Resources Board (CARB) fuel oil and ultra-low sulfur (ULS) fuel oil 
that is replacing the existing Environmental Protection Agency (EPA) 
red dyed fuel oil currently used at Diablo Canyon Power Plant 
(DCPP). Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube 
Oil, Starting Air, and Turbocharger Air Assist,'' requires, as a 
minimum, a supply of diesel fuel sufficient to support 7-days 
operation of the diesel generators (DGs) to power the minimum 
engineered safety feature (ESF) systems required to mitigate a 
design basis loss-of-coolant accident (LOCA) in one unit and those 
minimum required systems for a concurrent non-LOCA safe shutdown in 
the remaining unit (both units initially in Mode 1 operation). TS 
3.8.3 Condition A requires storage levels to be restored to within 
limits within 48 hours if they fall below the 7-day minimum, but 
remain above minimum limits for a 6-day supply. TS 3.8.3 also 
provides for tank cleaning on a 10-year frequency. During tank 
cleaning, TS 3.8.3 requires maintaining at least a 4-day supply.
    Because CARB and ULS fuel oils have a lower heat content than 
EPA fuel, it was necessary to recalculate the amount of fuel 
required to supply necessary loads for the required 7-day, 6-day, 
and 4-day time periods addressed in TS 3.8.3.
    The DGs and associated support systems, such as the fuel oil 
storage and transfer systems, are designed to mitigate accidents, 
and are not accident initiators. Revising the minimum volumes of 
stored fuel in the storage tanks will not result in any increase in 
the probability of any accident previously evaluated.
    Following implementation of this proposed change, there will be 
no change in the ability of the DGs to supply post-accident loads 
for 7 days, or 6 days if in TS 3.8.3 Condition A, or 4 days during 
tank cleaning. This is identical to the current requirements. 
Therefore, this change will not result in a significant increase in 
the consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Following implementation of this change, the DGs will still be 
able to power the minimum ESF systems required to mitigate a design 
basis LOCA in one unit and those minimum required systems for a 
concurrent non-LOCA safe shutdown in the remaining unit (both units 
initially in Mode 1 operation). The current 7-day, 6-day, and 4-day 
fuel supply requirements will be maintained. The DGs and associated 
fuel oil storage systems are not accident initiators, but are 
designed to mitigate accidents.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Following implementation of this change, the DGs will still have 
sufficient fuel oil supply to power the minimum ESF systems required 
to mitigate a design basis LOCA in one unit and those minimum 
required systems for a concurrent non-LOCA safe shutdown in the 
remaining unit (both units initially in Mode 1 operation). When fuel 
inventory is below that required to support 7 days of operation, the 
required actions depend on whether or not a 6-day supply is 
available, or a 4-day supply is available during tank cleaning. The 
proposed storage limits will maintain these 7-day, 6-day, and 4-day 
fuel supply requirements, including current margins, following the 
change to CARB and ULS fuel oils.
    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 requests involve no 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.

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: October 29, 2004.
    Description of amendment requests: The proposed amendments would 
revise the technical specifications (TS) requirements for handling of 
irradiated fuel in the containment and fuel building, and certain 
specifications related to performing core alterations. These changes 
are based on analysis of the postulated fuel handling and core 
alteration accidents and transients for the Diablo Canyon Power Plant, 
Units 1 and 2. The proposed amendment is consistent with the NRC-
approved Industry/Technical Specification Task Force (TSTF) Standard 
Technical Specifications Change Traveler TSTF-51, Revision 2, ``Revise 
containment requirements during handling irradiated fuel and core 
alterations.'' In addition, editorial corrections to TS 3.1.7, ``Rod 
Position Indication''; TS 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation''; TS 3.4.16, ``RCS Specific Activity''; TS 3.7.3, 
``Main Feedwater Isolation Valves (MFIVs), Main Feedwater Regulating 
Valves (MFRVs), MFRV Bypass Valves and Main Feedwater Pump (MFWP) 
Turbine Stop Valves''; and TS 3.7.13, ``Fuel Handling Building 
Ventilation System (FHBVS),'' are proposed.
    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 involves changes to accident 
mitigation system requirements. These systems are related to 
controlling the release of radioactivity to the environment and are 
not considered to be accident initiators for any previously analyzed 
accident. The proposed changes do not involve physical modifications 
to plant equipment, and do not change the operational methods or 
procedures used for moving irradiated fuel assemblies. As such, 
there are no accident initiators affected by the proposed amendment. 
Therefore, the proposed change does not impact the probability of 
postulated accidents.
    Consistent with the previously approved design basis analysis, 
the reanalysis of the containment fuel handling accident (FHA) 
concludes that radiological consequences of the accident at the 
Exclusion Area Boundary and the Low Population Zone Boundary are 
unchanged and remain well within the 10 CFR 100.11 limits, as 
defined by acceptance criteria in NUREG 0800, Section 15.7.4, and 
within the limits of general design criteria (GDC) 19 of 10 CFR 50, 
Appendix A. However, per this reanalysis, the calculated 30-day 
doses in the control room increased from 11.56 rem to 22.31 rem 
thyroid and from 0.00717 rem to 0.00757 rem whole body. Although 
these calculated doses increased they remain well within the 
acceptable limits of GDC 19 of 10 CFR 50, Appendix A, for the 
control room, which is 30 rem thyroid and 5 rem whole body. As a 
result, the increase in the doses is not considered to be a 
significant increase.
    The results of the core alteration events, other than a FHA, 
remain unchanged from the original design basis, which showed that 
these events do not result in fuel cladding integrity damage or 
radioactive releases. Therefore, the proposed changes do not 
significantly increase the consequences of any previously evaluated 
accident.
    In addition, the editorial corrections have no affect on the 
associated components, structures or systems, and their operation or 
design bases.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

[[Page 404]]

    The proposed change affects a previously evaluated accident 
(i.e., FHA). However, the proposed change does not introduce any new 
modes of plant operation and does not involve physical modifications 
to the plant. The proposed change does not change how design basis 
accidents were postulated nor does the proposed change initiate a 
new kind of accident or failure mode with a unique set of 
conditions.
    In addition, the editorial corrections have no affect on 
associated components, structures or systems, and their operation or 
design bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change imposes controls to ensure that during 
performance of activities that represent situations where 
radioactive releases are postulated, the radiological consequences 
are at or below the established licensing limit. Safety margins and 
analytical conservatisms have been evaluated and are understood. 
Substantial conservatism is retained to ensure that the analysis 
adequately bounds all postulated event scenarios. Specifically, the 
margin of safety for a FHA is the difference between the 10 CFR 
100.11 limits and the licensing limit defined by the NUREG-0800, 
Section 15.7.4. The licensing limit is defined by the NUREG as being 
``well within'' the 10 CFR 100.11 limits, with ``well within'' 
defined as 25 percent of the 10 CFR 100 limits of the FHA. Excess 
margin is the difference between the postulated doses and the 
corresponding licensing limit.
    The proposed applicability requirements continue to ensure that 
the whole-body, thyroid and total effective dose equivalent (TEDE) 
doses at the exclusion area and low population zone boundaries are 
at or below the corresponding licensing limit for both the FHA 
inside containment and in the fuel handling building. In addition, 
control room doses for both FHAs meet GDC 19 criterion. Although the 
control room doses as a result of the FHA inside containment 
reanalysis are somewhat higher then previously approved, they still 
remain well below the GDC-19 limits, therefore, the proposed change 
does not involve a significant reduction in a margin of safety.
    The margin of safety for core alteration events other than the 
FHA remains the same as the original licensing analyses, since the 
proposed change does not impact the TS requirements for systems 
needed to prevent or mitigate such core alteration events.
    In addition, the editorial corrections have no affect on 
associated equipment, components, structures or systems, and their 
operation or margin of safety.Therefore, 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 requests involve no 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 No. 50-388, Susquehanna Steam Electric 
Station, Unit 2, Luzerne County, Pennsylvania

    Date of amendment request: September 8, 2004.
    Description of amendment request: The proposed amendment would 
change the Unit 2 Technical Specifications (TSs) by revising the Unit 2 
Cycle 13 (U2C13) Minimum Critical Power Ratio (MCPR) Safety Limits in 
Section 2.1.1.2 and the references listed in Section 5.6.5.b.
    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 to the MCPR Safety Limits does not directly 
or indirectly affect any plant system, equipment, component, or 
change the processes used to operate the plant. Further, the U2C13 
MCPR Safety Limits are generated using NRC approved methodology and 
meet the applicable acceptance criteria. In addition, the effects of 
channel bow were conservatively addressed by increasing the amount 
of channel bow assumed in the MCPR SL calculation. Thus, this 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Prior to the startup of U2C13, licensing analyses are performed 
(using NRC approved methodology referenced in Technical 
Specification Section 5.6.5.b) to determine changes in the critical 
power ratio as a result of anticipated operational occurrences. 
These results are added to the MCPR Safety Limit values proposed 
herein to generate the MCPR operating limits in the U2C13 COLR [core 
operating limits report]. These limits could be different from those 
specified in the U2C12 COLR. The COLR operating limits thus assure 
that the MCPR Safety Limit will not be exceeded during normal 
operation or anticipated operational occurrences. Postulated 
accidents are also analyzed to confirm NRC acceptance criteria are 
met.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U2C13 core operating limits. The use of this approved methodology 
does not increase the probability or consequences of an accident 
previously evaluated.
    Therefore, this 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?
    Response: No
    The change to the MCPR Safety Limits does not directly or 
indirectly affect any plant system, equipment, or component and 
therefore does not affect the failure modes of any of these systems. 
Thus, the proposed changes do not create the possibility of a 
previously unevaluated operator or a new single failure.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U2C13 core operating limits. The use of this approved methodology 
does not create the possibility of a new or different kind of 
accident.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Since the proposed changes do not alter any plant system, 
equipment, component, or the processes used to operate the plant, 
the proposed change will not jeopardize or degrade the function or 
operation of any plant system or component governed by Technical 
Specifications. The proposed MCPR Safety Limits do not involve a 
significant reduction in the margin of safety as currently defined 
in the Bases of the applicable Technical Specification sections, 
because the MCPR Safety Limits calculated for U2C13 preserve the 
required margin of safety.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U2C13 core operating limits. This approved methodology is used to 
demonstrate that all applicable criteria are met, thus, 
demonstrating that there is no reduction in the margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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.

[[Page 405]]

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

    Date of amendment request: August 23, 2004.
    Description of amendment request: The proposed amendments would 
revise the Surveillance Requirements for Technical Specifications 
3.6.1.3, ``Primary Containment Isolation Valves,'' for Hatch Units 1 
and 2. The proposed amendments would substitute the requirement for 
valve seat replacement with a requirement to perform an Appendix J 
leakage rate test on the valves. Conforming revisions to the Technical 
Specification Bases B 3.6.1.3, ``Primary Containment Isolation Valves'' 
are also included.
    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?
    The proposal would change the Technical Specifications 
Surveillance Requirement for containment purge valves with resilient 
seats. The proposed change does not involve a significant increase 
in the probability or consequence of an accident previously 
evaluated because the extensive industry operating experience 
derived from test results has demonstrated that the resilient seat 
material does not experience aging degradation and cause containment 
isolation valves to leak. Thus, the valves will perform as assumed 
in the accident analyses and therefore, this change does not involve 
a significant increase in the consequences of an accident previously 
evaluated. Further, these valves are not accident initiators, and 
therefore, this change does not involve a significant increase in 
the probability of occurrence of a previously evaluated event.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposal would change the Technical Specifications 
Surveillance Requirement for containment purge valves with resilient 
seats. The proposed change does not involve physical alteration of 
the plant (no new or different type of equipment will be installed 
nor changes in methods governing normal plant operation). In 
particular, it does not require the valves to function in any manner 
other than that which is currently required. Thus, this change does 
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 the margin of safety.
    The proposal would change the Technical Specifications 
Surveillance Requirement for containment purge valves with resilient 
seats. The proposed change does not involve a significant reduction 
in margin of safety because it has no effect on any safety analysis 
bases or assumptions. It does not change the leakage acceptance 
criteria. Sufficient data has been collected to demonstrate that 
resilient seats do not experience aging degradation. Deleting the 
seat replacement requirement will not reduce the margin of safety 
provided by Technical Specifications.
    For the above reasons, the margin of safety is not reduced by 
this proposed Technical Specifications change.

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

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

    Date of amendment request: December 9, 2004.
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Updated Final Safety Analysis Report to include an 
alternate methodology for concrete reinforcement bar splicing. The 
change in methodology applies to restoration of the concrete Shield 
Building dome as part of the upcoming steam generator replacement 
project. The alternate methodology uses a Bar-Lock mechanical splice in 
lieu of the Cadweld splice used for the original design and 
construction of the plant.
    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.
    No changes in event classification, as discussed in the UFSAR 
[Updated Final Safety Analysis Report] Chapter 15, will occur due to 
use of the Bar-Lock couplers.
    The restoration of the temporary concrete construction openings 
in the Shield Building will utilize Bar-Lock couplers to splice new 
rebar to the existing rebar. The Shield Building structure limits 
the release of radioactivity following an accident and protects the 
systems, structures, and components inside containment from external 
events. The accidents of interest are those that rely on the Shield 
Building to limit the release of radioactivity to the environment, 
and those that result from some external events. The design of the 
Shield Building is such that it is not postulated to fail and 
initiate an accident described in the UFSAR.
    The Bar-Lock coupler qualification tests detailed in Topical 
Report 24370-TR-C-001-A demonstrate that the Bar-Lock coupler meets 
the ASME [American Society of Mechanical Engineers] strength 
requirements and is, therefore, acceptable for use in nuclear 
safety-related applications. Based on these test results, it is 
concluded that use of the Bar-Lock couplers in restoring the 
temporary concrete construction openings will not reduce the 
structural capability of the repaired structure. The Shield Building 
will continue to perform its design function as described in the WBN 
UFSAR.
    Therefore, the proposed use of the Bar-Lock couplers will not 
significantly increase 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 design of the Shield Building is such that it is not 
postulated to fail and initiate an accident described in the UFSAR. 
The Bar-Lock couplers are passive devices and as such will not 
initiate or cause an accident.
    The restoration of the temporary concrete construction openings 
in the Shield Building will utilize Bar-Lock couplers to splice new 
rebar to the existing rebar. The Bar-Lock coupler qualification 
tests detailed in Topical Report 24370-TR-C-001-A demonstrate that 
the Bar-Lock coupler meets the ASME strength requirements and is, 
therefore, acceptable for use in nuclear safety-related 
applications. Based on these test results, it is concluded that use 
of the Bar-Lock couplers in restoring the temporary concrete 
construction openings will not reduce the structural capability of 
the Shield Building. The Shield Building will, therefore, continue 
to perform its design functions as described in the WBN UFSAR.
    Therefore, the possibility of a new or different accident 
situation occurring as a result of this condition is not created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    As indicated in the WBN UFSAR, the structural design of the 
reinforced concrete Shield Building is in compliance with the 
proposed ACI-ASME [American Concrete Institute--American Society of 
Mechanical Engineers] (ACI-359) Code for Concrete Reactor Vessels 
and Containment, Article CC-3000, as issued for trial use, April 
1973, for the loading combinations defined in UFSAR Table 3.8.1-1. 
Allowable stresses are based on this code with the exception of 
allowable tangential shear stresses in walls

[[Page 406]]

where the ACI 318-71 code is used. The reinforcing steel conforms to 
the requirements of American Society for Testing Maintenance (ASTM) 
A 615, Grade 60. The WBN UFSAR states that reinforcing bars were lap 
spliced and Cadwelded in accordance with ACI 318-7 requirements for 
strength design.
    The restoration of the temporary concrete construction openings 
in the Shield Building will utilize Bar-Lock couplers to splice new 
rebar to the existing rebar. The restoration of the construction 
openings, including use of the Bar-Lock couplers, will conform to 
the requirements of ACI-359 (April 1973) and ACI 318. Therefore, 
following completion of the modification, the Shield Building will 
continue to comply with ACI-359 (April 1973) and ACI 318 
requirements.
    In addition to conforming to ACI-359 (April 1973) and ACI 318 
requirements, the Bar-Lock coupler qualification tests detailed in 
Topical Report 24370-TR-C-001-A demonstrate that the Bar-Lock 
coupler meets the ASME strength requirements.
    Therefore, a significant reduction in the margin to safety is 
not created by this modification.

    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.

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

    Date of amendment request: October 27, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.'' The Table of 
Contents will also be revised to reflect the deletions.
    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 October 27, 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 a significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert Gramm.

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