Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 10071-10084 [06-1737]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 39 (Tuesday, February 28, 2006)]
[Notices]
[Pages 10071-10084]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-1737]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 3, 2006, to February 15, 2006. The 
last biweekly notice was published on February 14, 2006 (71 FR 7804).

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

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

[[Page 10072]]

    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: 
Rulemaking 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 email 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.

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

    Date of amendment request: November 30, 2005.
    Description of amendment request: The proposed amendment would 
revise the frequency of the diesel generator automatic trips bypass 
surveillance requirement (SR) 3.8.1.11 from 18 months to 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change decreases the frequency of SR 
3.8.1.11, verification of the DG [diesel generator] automatic trips 
bypass, from 18 months to 24 months. The DG automatic trips bypass 
circuitry is required for DG operability and reliability during 
emergency operation of the DG. The proposed test frequency will 
continue to assure that the DG will perform as required. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated, 
because the factors that are used to determine the probability and 
consequences of accidents are not being affected.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated. There 
are no new or different accident initiators or sequences being 
created by the proposed Technical Specifications change. The 
required surveillance performed at the proposed frequency will 
continue to provide assurance that the trips bypass function is 
operable and is properly supporting operation of the associated DG. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    No. The proposed change does not involve a significant reduction 
in the margin of

[[Page 10073]]

safety. The proposed change will continue to ensure that the DG 
trips bypass function operates as designed. The functionality and 
operability of emergency power system is not being changed. 
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 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 Branch Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: January 4, 2006.
    Description of amendment request: The proposed amendment would 
change the Millstone Power Station, Unit No. 2 Technical Specification 
(TS) 3/4.3.3.8, ``Instrumentation, Accident Monitoring,'' to modify the 
description of the pressurizer power operated relief valves (PORVs) and 
pressurizer safety valves position indicators.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment removes the wording ``Acoustic Monitor,'' 
which provides specific details related to system design, from items 
4 and 6 of TS 3/4.3.3.8, Tables 3.3-11 and 4.3-7. The PORVs and 
Pressurizer Safety Valves position indicators (and the associated 
``Acoustic Monitor'') provide only indications of valve position. 
They do not constitute a design feature that is an initial condition 
for a design basis accident or transient analysis. Furthermore, they 
do not affect the function of the system, equipment in the system or 
actuate to mitigate a design basis accident or transient. Therefore, 
the proposed changes do not increase the probability or consequences 
of an accident previously evaluated.
    Additionally, the TS retains the requirement for the total and 
minimum channels required to be OPERABLE and to verify channel 
OPERABILITY at the designated frequencies. The PORVs and Pressurizer 
Safety Valves are equipped with positive position indication that 
meets the requirements of RG [Regulatory Guide] 1.97.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not impact the capability of existing 
equipment to perform its intended functions. No system setpoints are 
being modified and no changes are being made to the method in which 
plant operations are conducted. No new failure modes that would 
impact accident analyses are introduced by the proposed changes. The 
proposed amendment does not introduce accident initiators or 
malfunctions that would cause a new or different kind of accident. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment removes the wording ``Acoustic Monitor'' 
from items 4 and 6 of TS 3/4.3.3.8, Table[s] 3.3-11 and 4.3-7. The 
proposed changes do not affect any of the assumptions used in the 
accident analysis, nor does it affect any operability requirements 
for equipment important to plant safety. Therefore, the margin of 
safety is not impacted by the proposed amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Branch Chief: Darrell J. Roberts.

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

    Date of amendment request: December 30, 2005.
    Description of amendment request: The proposed amendment 
establishes a combined leakage rate limit for the sum of the four Main 
Steam line leakage rates that is equal to four times the current 
individual Main Steam Isolation Valve (MSIV) leakage rate limit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a change to structures, 
systems, or components that would affect the probability of an 
accident previously evaluated in the Cooper Nuclear Station (CNS) 
Updated Safety Analysis Report (USAR). The proposed amendment 
results in no change in the radiological consequences of the design 
basis Loss-of-Coolant Accident (LOCA) as currently analyzed for CNS. 
That analysis was calculated for a combined Main Steam Isolation 
Valve (MSIV) leakage for determining acceptance to the regulatory 
limits for the offsite and Control Room radiation doses, as 
contained in 10 CFR 100 [Part 100 of Title 10 of the Code of Federal 
Regulations] and 10 CFR 50[,] Appendix A, General Design Criterion 
(GDC) 19. The aggregate Main Steam line leakage rate limit has no 
adverse effect on the environmental qualification of equipment 
important to safety, as provided for in 10 CFR 50.49.
    Based on the above conclusions, this proposed amendment does 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.
    The proposed change does not modify the MSIVs or any other plant 
system or structure associated with this amendment and therefore, 
will not affect their capability to perform their design function. 
The combined total Main Steam line leakage rate is included in the 
current radiological analyses for the assessment of radiation 
exposure following an accident. This License Amendment Request 
revises the allowable leakage rate from a per valve limit to a total 
combined leakage rate limit for all four Main Steam lines but does 
not change the cumulative limit.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The leakage rate limit specified for the MSIVs is used to 
quantify the maximum amount of Secondary Containment bypass leakage 
assumed in the LOCA radiological analysis. Results of the analysis 
are evaluated against the dose limits contained in 10 CFR 50[,] 
Appendix A[,] GDC 19 and 10 CFR 100. The margin of safety in this 
context is considered to be the difference between the calculated 
dose exposures and the limits provided by GDC 19 and 10 CFR 100.
    Therefore, since the proposed combined Main Steam line leakage 
rate limit is unchanged from the assumed maximum leakage rate for 
MSIVs, for the purpose of calculating [a] potential radiation dose, 
the margin of safety is not affected because the postulated 
radiation doses remain the same.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 10074]]

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 C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: January 30, 2006.
    Description of amendment request: The proposed change allows a 
delay time for entering a supported system Technical Specification (TS) 
when the inoperability is due solely to an inoperable snubber, if risk 
is assessed and managed consistent with the program in place for 
complying with the requirements of 10 CFR 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.8 is added to the TS to provide this 
allowance and define the requirements and limitations for its use.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
November 24, 2004 (69 FR 68412), on possible amendments concerning 
TSTF-372, including a model safety evaluation and model no significant 
hazards consideration (NSHC) determination, using the consolidated line 
item improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on May 4, 2005 (70 FR 23252). The 
licensee affirmed the applicability of the following NSHC determination 
in its application dated January 30, 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on 
allowance provided by proposed LCO 3.0.8 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.8. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change. The addition of a 
requirement to assess and manage the risk introduced by this change 
will further minimize possible concerns. Therefore, this change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in Regulatory Guide 1.177. A bounding risk assessment 
was performed to justify the proposed TS changes. The proposed LCO 
3.0.8 defines limitations on the use of the provision and includes a 
requirement for the licensee to assess and manage the risk 
associated with operation with an inoperable snubber. The net change 
to the margin of safety is insignificant. Therefore, this change 
does not involve a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

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: November 11, 2005.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.6.5, ``Containment Spray and 
Cooling Systems''; an existing Condition, two Surveillance 
Requirements, and add a new Condition which will allow continued plant 
operation with TS limitations when two Containment Cooling System fan 
coil units (FCUs), one in each train, are inoperable.
    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 Technical 
Specifications to allow plant operation to continue for a limited 
time period under Technical Specification controls with two fan coil 
units, one fan coil unit from each containment cooling train, 
providing the required cooling function. Analyses demonstrate that 
any two fan coil units, whether they are in the same train or from 
opposite trains, are sufficient to supply the required containment 
cooling following a design basis accident when the plant in the 
proper configuration as required by the proposed Technical 
Specifications.
    The containment cooling system is required for accident 
mitigation and is not an accident initiator, thus revising the 
equipment required to provide the safety function does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    Since the proposed change continues to provide the post-accident 
containment cooling function under Technical Specification controls, 
this change does not involve an increase in the consequences of an 
accident. Thus this change does 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 Technical 
Specifications to allow plant operation to continue for a limited 
time period under Technical Specification controls with two fan coil 
units, one fan coil unit from each containment cooling train, 
providing the required cooling function. Analyses demonstrate that 
any two fan coil units, whether they are in the same train or

[[Page 10075]]

from opposite trains, are sufficient to supply the required 
containment cooling following a design basis accident when the plant 
in the proper configuration as required by the proposed Technical 
Specifications.
    The proposed licensing basis changes do not involve a change in 
the function or use of the containment cooling system. It does 
assure that the containment cooling function is provided during 
plant operations for post-accident mitigation. There are no new 
failure modes or mechanisms created through allowing different 
combinations of fan coil units to provide the cooling function as 
proposed by this Technical Specification change. There are no new 
accident precursors generated by providing the required cooling 
function with an operable fan coil unit from each train.
    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 Technical 
Specifications to allow plant operation to continue for a limited 
time period under Technical Specification controls with two fan coil 
units, one fan coil unit from each containment cooling train, 
providing the required cooling function. Analyses demonstrate that 
any two fan coil units, whether they are in the same train or from 
opposite trains, are sufficient to supply the required containment 
cooling following a design basis accident when the plant in the 
proper configuration as required by the proposed Technical 
Specifications.
    Current plant Technical Specifications allow plant operation to 
continue for 7 days with the containment cooling function provided 
by the two operable fan coil units of a single operable containment 
cooling train. This is acceptable because engineering analyses 
demonstrate that the two fan coil units of a single train can 
provide the required post-accident containment cooling.
    Likewise, engineering analyses demonstrate that any two fan coil 
units from opposite containment cooling trains can also provide the 
required post-accident containment cooling if the cooling water flow 
to the other fan coil unit in each train is isolated. This license 
amendment request proposes Technical Specifications which will allow 
plant operation to continue for 7 days with the containment cooling 
function provided by two fan coils from opposite trains provided the 
cooling water flow to the other fan coil unit in each train is 
isolated. Thus, from a cooling capacity perspective, this proposed 
Technical Specification change does not involve a reduction in a 
margin of safety.
    When inoperable plant systems are under Technical Specification 
controls that limit the time for inoperability, a single failure in 
addition to the inoperable equipment is not postulated. Therefore, 
whether two inoperable fan coil units are in the same train or 
opposite trains does not change the availability of the two 
remaining operable fan coil units. Thus from a Technical 
Specification perspective, this proposed Technical Specification 
change does not involve a reduction in a margin of safety.
    Therefore, based on the considerations given above, 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 Acting Branch Chief: Timothy J. Kobetz.

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

    Date of amendment request: December 19, 2005.
    Description of amendment request: The proposed change will revise 
Fort Calhoun Station, (FCS) Technical Specification 2.4, ``Containment 
Cooling,'' (and associated Bases) to reduce the required number of 
operable Containment Spray (CS) pumps from three to two in order to 
enhance net positive suction head (NPSH) margins. This change will be 
accomplished by disabling the containment spray actuation signal (CSAS) 
automatic start feature of CS pump SI-3C. This change will reduce the 
head loss across the containment sump strainers during the 
recirculation phase of a design-basis accident (DBA) by reducing flow 
rates, and will improve NPSH available (NPSHA).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Containment Spray (CS) system is not an initiator of any 
accident previously evaluated at the Fort Calhoun Station (FCS); the 
CS system is an accident mitigation system. The CS system's 
licensing basis functions are to limit the containment pressure rise 
and reduce the leakage of airborne radioactivity from the 
containment by providing a means for cooling the containment 
following a loss-of-coolant accident (LOCA) or main steam line break 
(MSLB) inside containment. The proposed change disables the CSAS 
automatic start feature of one of the three CS pumps.
    The only FCS safety analysis that currently assumes three CS 
pumps operating to mitigate an accident is the Containment Pressure 
Analysis for a[n] MSLB inside containment. Even though this analysis 
assumes operation of all three CS pumps, it also shows that peak 
containment pressure occurs prior to the CS system starting, 
therefore, the CS system does not mitigate the peak pressure for 
a[n] MSLB. The reviews evaluated both existing AORs [analyses of 
record] and those analyses developed for the Steam Generator 
Replacement (RSG) project. The analysis developed for the RSG 
project that evaluates the Containment Pressure Analysis for MSLB 
inside containment was reviewed for the impact of reducing the 
number of operating CS pumps from three to two. This review 
determined that the RSG MSLB analysis will be acceptable and will 
continue to be bounded by the analysis currently documented in USAR. 
AOR peak pressure is unaffected by implementation of this proposed 
change. Therefore, the combination of the RSG project and this 
containment spray modification will not result in an increase in the 
currently documented peak containment pressure for an MSLB. 
Therefore, the evaluation for the MSLB event has determined that the 
containment pressure response is acceptable with less than three CS 
pumps operating.
    The LOCA analysis source term is based on operation of minimum 
safeguards due to a worst-case single failure. The minimum 
safeguards configuration is unchanged by this modification. 
Following implementation of the proposed change at least one CS pump 
will be available to mitigate a LOCA as currently assumed in the 
analysis, therefore, the proposed change will have no adverse effect 
on the radiological consequences following a LOCA. The analyses that 
establish the radiological consequences for the site are based on a 
Large Break LOCA with a single CS pump in operation, therefore, 
single CS pump operation during a[n] MSLB inside containment is 
bounded by the LOCA analysis.
    Therefore, the proposed change 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 proposed change will reduce the number of operable CS pumps 
from three to two; however, previous accident analyses will remain 
valid. No credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing basis have 
been created and none of the initial condition assumptions of any 
accident evaluated in the safety analysis are impacted.
    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 amendment involve a significant reduction 
in a margin of safety?

[[Page 10076]]

    Response: No.
    The containment building and associated penetrations are 
designed to withstand an internal pressure of 60 psig at 305 [deg]F, 
including all thermal loads resulting from the temperature 
associated with this pressure, with a leakage rate of 0.1 percent by 
weight or less of the contained volume per 24 hours. The CS System 
and the Containment Fan Coolers are credited for maintaining 
containment pressure and temperatures within design limitations, and 
assure that the release of fission products to the environment 
following a design[-]basis accident will not exceed regulatory 
guidelines. The FCS licensing basis credits only one of the three CS 
pumps to limit the containment pressure to below the design value 
for a LOCA. Currently, the FCS licensing basis credits three CS 
pumps for a[n] MSLB, however, the CS system is not credited for 
limiting peak containment pressure for a[n] MSLB.
    The EEQ [electrical equipment qualification] profile developed 
for the current plant configuration bounds those associated with the 
upcoming RSG modification. Both the proposed CS system changes and 
the RSG projects are scheduled for the same refueling outage. The 
thermal lag analysis of equipment performed using the current plant 
configuration demonstrated a large margin between the equipment 
evaluated during the accident versus the conditions under which it 
was tested. The RSG modification will further increase this margin. 
As part of the RSG effort the EEQ analysis will be revised to 
address RSG issues and will include the changes to containment 
spray. When the margins associated with the current analysis as well 
as increases in margin when the new analysis is implemented it is 
expected that the changes to the containment spray system will not 
produce an adverse result. All equipment will remain qualified to 
operate in the accident environment.
    Additionally, the CFCs [containment fan coolers] operate 
independently of the CS system to remove heat from the containment 
atmosphere. The CFCs consist of two redundant trains; each train 
with one air cooling and filtering unit and one air cooling unit, 
for a total of four cooling units. Operation of the CFCs is credited 
in the MSLB containment pressure analysis. The CFCs are not impacted 
by this proposed change. During the MSLB containment spray takes 
place after the peak containment pressure occurs. Therefore, the 
licensing basis capabilities of the Containment Cooling System, 
which consists of the CS and CFCs, is not adversely affected by the 
proposed change; the ability to maintain containment peak pressure 
and temperature and long[-]term containment pressure and temperature 
will be maintained.
    Particulate fission products that are released into the 
containment following a DBA are removed by the CS system for those 
events that result in CS actuation. The water spray strips 
radioactive particles from the atmosphere where they fall to the 
floor and are washed into the containment sump. The radiological 
consequences analysis credits CS system operation for removal of 
particulates from the containment atmosphere during a LOCA. The LOCA 
analysis source term is based on operation of minimum safeguards due 
to a worst-case single failure, and a presumption of core damage. 
Minimum safeguards corresponds to one CS pump and one CS header 
operation and take into account pump degradation, and instrument 
uncertainties. The analyses that establish the radiological 
consequences for the site are not impacted by the proposed 
modification. These analyses are based on a Large Break LOCA with a 
single CS pump in operation. Therefore, single CS pump operation 
bounds the plant configuration following the proposed modification.
    The Large Break LOCA assumes that there will be three CS pumps 
operating when evaluating the effects of containment pressure on 
ECCS [emergency core cooling system] performance. The analysis 
assumes three CS pumps, which minimizes containment pressure, to 
conservatively evaluate ECCS performance in response to a LOCA. The 
use of two CS pumps versus three improves ECCS performance and thus 
increases margin to 10 CFR 50.46 limits on peak clad temperature.
    In summary, following implementation of the proposed change:
     Peak containment pressure for analyzed DBAs will not be 
increased;
     The assumptions used in the environmental qualification 
of equipment exposed to the containment atmosphere following a DBA 
remaining bounding; and
     The radiological consequences for the bounding DBA 
remains unchanged.
     The currently calculated peak clad temperature 
following a LOCA remains bounded by existing analysis.
    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 R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Branch Chief: David Terao.

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon 
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California

    Date of amendment requests: January 13, 2006.
    Description of amendment requests: The proposed amendment would 
revise Technical Specification 5.6.5, ``Core Operating Limits Report 
(COLR),'' by adding WCAP-16009-P-A, ``Realistic Large-Break LOCA [Loss-
of-Coolant Accident] Evaluation Methodology Using the Automated 
Statistical Treatment of Uncertainty Method (ASTRUM),'' dated January 
2005, as an approved analytical method for determining core operating 
limits for Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to allow the use of the best estimate loss-
of-coolant accident (LOCA) analysis methodology using the automated 
statistical treatment of uncertainty methodology (ASTRUM) does not 
involve a physical alteration of any plant equipment or change 
operating practice at Unit 2 of Diablo Canyon Power Plant (DCPP). 
Therefore, there will be no increase in the probability of a LOCA. 
The consequences of a LOCA are not being increased.
    The plant conditions assumed in the analysis are bounded by the 
design conditions for all equipment in Unit 2. That is, it is shown 
that the emergency core cooling system is designed so that its 
calculated cooling performance conforms to the criteria contained in 
10 CFR [Title 10 of the Code of Federal Regulations, Section] 50.46, 
paragraph b. No other accident is potentially affected by this 
change.
    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 accident from any accident previously evaluated?
    Response: No.
    The proposed change would not result in any physical alteration 
to any Unit 2 system, and there would not be a change in the method 
by which any safety [-]related system performs its function. 
Analyses of transient events have confirmed that no transient event 
results in a new sequence of events that could lead to a new 
accident scenario. The parameters assumed in the analysis are within 
the design limits of existing plant equipment.
    In addition, employing the ASTRUM methodology does not create 
any new failure modes that could lead to a different kind of 
accident. The design of all systems remains unchanged and no changes 
are being made to any reactor protection system or engineered 
safeguard features actuation setpoints.
    Based on this review, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed changes.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.

[[Page 10077]]

    It has been shown that the analytic technique used in the 
analysis realistically describes the expected behavior of the DCPP 
Unit 2 reactor system during a postulated LOCA. Uncertainties have 
been accounted for as required by 10 CFR 50.46. A sufficient number 
of LOCAs with different break sizes, different locations, and other 
variations in properties have been analyzed to provide assurance 
that the most severe postulated LOCAs were analyzed. The analysis 
has demonstrated that all acceptance criteria contained in 10 CFR 
50.46[,] paragraph b continue to be 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 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 Branch Chief: David Terao.

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power 
Plant (HBPP), Unit 3 Humboldt County, California

    Date of amendment request: January 19, 2006.
    Description of amendment request: The licensee has proposed to 
revise the Technical Specifications (TS) to correct an editorial error 
in TS 3.1.2, ``Spent Fuel Pool Load Restrictions,'' and to change TS 
5.2.2, ``Facility Staff,'' to allow the Unit 3 control room to be 
temporarily unmanned during emergency conditions that require personnel 
to evacuate buildings for their safety.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed editorial change has no impact on probability or 
consequences of accidents. The following discussion applies to the 
proposed change related to control room evacuation.
    Allowing plant personnel to not continuously man the control 
room has no impact on the probability of an accident from occurring, 
especially acts of nature such as earthquakes and tsunamis.
    The HBPP DSAR, Appendix A, and NRC SER, Section 10, dated April 
29, 1987, evaluate various accidents at HBPP. Because all fuel has 
been removed from the reactor vessel and stored in the spent fuel 
pool, the majority of accidents analyzed pertain to events that 
could only affect spent fuel or the spent fuel pool. All accidents 
affecting spent fuel or the spent fuel pool do not require operator 
action to protect the public health and safety or to maintain 
offsite radiological doses well within regulatory limits. In 
addition, NRC SER, Section 10.7, ``Impact of Tsunami Flooding,'' 
analyzes the impact of tsunami flooding. That analysis identifies a 
likely impact of the tsunami to be a release of the radwaste tank 
radionuclide contents to the bay and some damage to the reactor 
building. For both situations, no operator action is required to 
maintain offsite radiological doses well within regulatory limits.
    Allowing the control room to be temporarily unmanned under 
emergency conditions does not create problems that could increase 
the consequences of an accident. The primary function of manning the 
control room is for an operator to observe and acknowledge alarms. 
Recovery actions to respond to damage to spent fuel, the spent fuel 
pool, or radwaste tanks are taken by personnel outside the control 
room. No recovery actions are required to be taken by the control 
room operator to respond to damage to spent fuel, the spent fuel 
pool, or radwaste tanks.
    Evacuating occupied buildings, including the control room, 
during a tsunami, allows the control room operator to return to the 
control room after the tsunami and assess damage by observing 
indicators and alarms. Upon returning to the control room, the 
operator would be able to direct and monitor recovery efforts from 
the control room that may be necessary to bring plant parameters 
within required specifications.
    If an operator remains in the control room during a tsunami and 
becomes injured, that operator would be unable to direct and monitor 
recovery efforts. Under this scenario, other plant personnel who 
evacuated to higher ground onsite within the OCA would eventually 
return to the plant, including the control room, and perform any 
required recovery functions. Therefore, consequences of a tsunami 
are not increased by not continually manning the control room during 
the event.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident evaluated?
    Response: No.
    The proposed editorial change has no impact on accidents. The 
following discussion applies to the proposed change related to 
control room evacuation.
    As discussed in the response to question 1 above, none of the 
analyzed accidents require operator action to keep offsite 
radiological doses well within regulatory limits. In addition, 
allowing plant personnel to not continuously man the control room 
after an emergency situation has occurred, has no impact on the 
possibility of a new or different kind of accident from occurring. 
If the plant is evacuated, no work activities will be performed in 
the plant. With the plant in SAFSTOR and no work being performed, 
there are no actions required to be taken by personnel manning the 
control room.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: The proposed editorial change has no impact on margin 
of safety. The following discussion applies to the proposed change 
related to control room evacuation.
    NRC SER Section 10.8, ``Accident Analysis Conclusions,'' 
summarizes the consequences from accidents in terms of offsite 
radiological doses. SER Section 10.8 includes the statement, ``The 
(NRC) staff has determined that offsite radiological consequences 
due to a tsunami are within acceptable dose guideline values.'' As 
discussed in the response to question 1 above, none of the analyzed 
accidents require operator action to keep offsite radiological doses 
well within regulatory limits. Therefore, temporarily not manning 
the control room during an emergency will have no impact on the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based upon 
the staff's review of the licensee's analyses as well as the staff's 
own evaluation, the staff concludes that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Richard F. Locke, Esquire, Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Claudia Craig.

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

    Date of amendment request: January 31, 2006.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.8.3.1, ``Onsite Power Distribution-
Operating,'' to extend the allowed outage time (AOT) for an inoperable 
Class 1E vital 120-volt alternating current inverter. The TS currently 
provides an AOT of 24 hours to restore an inoperable inverter. Based on 
risk-informed assessment, the amendments would extend the AOT to 7 
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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed formatting changes to TS 3.8.3.1 Action b and the 
change to the AOT

[[Page 10078]]

for an inoperable inverter to be extended from 24 hours to 7 days do 
not alter any plant equipment or operating practices in such a 
manner that the probability of an accident is increased. The 
proposed changes will not alter assumptions relative to the 
mitigation of an accident or transient event.
    An evaluation was performed to determine the risk significance 
of the proposed change to the AOT. The risk evaluation concludes 
that the [Delta]CDF [core damage frequency] and [Delta]LERF [large 
early release frequency] associated with the proposed changes are 
1.88E-07 and 2.05E-09, respectively, which are characterized as 
``very small changes'' by RG [Regulatory Guide] 1.174. The ICCDP 
[incremental conditional core damage probability] and ICLERP 
[incremental conditional large early release probability] associated 
with the proposed change are 3.63E-07 and 1.08E-08, respectively, 
which are within the acceptance criteria in RG 1.177. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, reactor 
coolant pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed change to TS 
3.8.3.1 to allow the AOT for an inoperable inverter to be extended 
from 24 hours to 7 days has been evaluated for its effect on plant 
safety. The risk-informed evaluation concludes that the [Delta]CDF 
and [Delta]LERF associated with the proposed change are 1.88E-07 and 
2.05E-09, respectively, which are characterized as ``very small 
changes'' by RG 1.174. The ICCDP and ICLERP associated with the 
proposed change are 3.63E-07 and 1.08E-08, respectively, which are 
within the acceptance criteria in RG 1.177. The proposed changes to 
the formatting of TS 3.8.3.1 Action b are administrative only and 
have no impact on margin of safety. Therefore, the proposed changes 
do not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: December 14, 2005 (TS-05-07).
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 5.7.2.19, ``Containment Leakage 
Rate Testing Program,'' to allow a one time, 5-year extension to the 
current 10-year test interval for the performance-based leakage rate 
test program for 10 CFR Part 50, Appendix J, Type A tests.
    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 for extending Type A test frequency does not 
significantly increase the probability of an accident previously 
evaluated since the change is not a modification to plant systems, 
nor a change to plant operation that could initiate an accident.
    TVA performed an evaluation of the risk significance for the 
proposed increase to the WBN Unit 1 Type A test frequency. The 
results of the TVA risk evaluation indicates that the increase in 
Large Early Release Frequency (LERF) remains below the level of risk 
significance defined in the NRC Regulatory Guide 1.174, ``An 
Approach for Using Probabilistic Risk Assessment In Risk-Informed 
Decisions On Plant-Specific Changes to the Licensing Basis.'' TVA's 
evaluation indicates that the calculated increase in frequency for 
all releases (small, large, early and late) and the increase in 
radiation dose to the population are also non-risk significant.
    The proposed test interval extension does not involve a 
significant increase in the consequences of an accident. Research 
documented in NUREG-1493, ``Performance-Based Containment Leakage-
Test Program,'' determined that generically, very few potential 
containment leakage paths fail to be identified by Type A tests. An 
analysis of 144 Type A test results, including 23 failures, found 
that no failures were due to containment liner breach. The NUREG 
concluded that reducing the Type A test frequency to once per 20 
years would lead to an imperceptible increase in risk. Furthermore, 
the NUREG concluded that Type B and C testing provides assurance 
that containment leakage from penetration leak paths (i.e., valves, 
flanges, containment air-locks) identify any leakage that would 
otherwise be detected by the Type A tests.
    In addition to the NUREG conclusions, TVA's American Society of 
Mechanical Engineers (ASME) IWE program performs containment 
inspections in order to detect evidence of degradation that may 
either affect the containment structural integrity or leak 
tightness.
    Therefore, the proposed extension of the Type A test interval 
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 to extend the Type A test interval does not 
create the possibility of a new or different type of accident 
because there are no physical changes made to the plant or plant 
equipment governing normal plant operation. There are no changes to 
the operation of the plant that would introduce a new failure mode 
creating the possibility of a new or different kind of accident. 
Therefore, the proposed extension 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 to extend the Type A test interval will not 
significantly reduce the margin of safety. A generic study 
documented in NUREG-1493 indicates that extending the Type A leak 
test interval to 20 years would result in an imperceptible increase 
in risk to the public. The NUREG also found that, generically, the 
containment leakage rate contributes a very small amount to the 
individual risk and that the decrease in the Type A test frequency 
would have a minimal effect on risk because most potential leakage 
paths are detected by Type C testing.
    Previous Type A leakage tests conducted on WBN Unit 1 indicate 
that leakage from containment have been less than the 10 CFR 50, 
Appendix J leakage limit of 1.0 La. A review of the 
previous Type A test results indicate a stable trend with an 
increase of less than 15 percent of La, well below the 
1.0 La leakage limit.
    Therefore, these test results, in conjunction with the research 
findings from NUREG-1493, provide assurance that the proposed 
extension to the Type A test interval does not involve a significant 
reduction in a margin of safety.

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

[[Page 10079]]

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

    Date of application request: August 26, 2005, as supplemented by 
letter dated December 16, 2005.
    Description of amendment request: The amendment would authorize 
changes to the Final Safety Analysis Report (FSAR) for the Callaway 
Plant, Unit 1, that would revise the methodology for the reactor 
coolant system (RCS) leak detection instrumentation. This revision 
would clarify the requirements of the containment atmosphere gaseous 
radioactivity monitor with regard to the RCS leak detection capability