Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 15940-15955 [E5-1343]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 59 (Tuesday, March 29, 2005)]
[Notices]
[Pages 15940-15955]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E5-1343]


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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 March 4, 2005, through March 17, 2005. The 
last biweekly notice was published on March 15, 2005 (70 FR 12743).

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.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set

[[Page 15941]]

forth with particularity the interest of the petitioner in the 
proceeding, and how that interest may be affected by the results of the 
proceeding. The petition should specifically explain the reasons why 
intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: February 24, 2005.
    Description of amendment request: The licensee proposed to revise 
Table 3.1.1, ``Protective Instrumentation Requirements,'' of the 
Technical Specifications to clarify the conditions under which the 
reactor building closed cooling water (RBCCW) pumps and the service 
water (SW) pumps will trip during a loss-of-coolant accident (LOCA). 
The revised wording would state that the RBCCW and SW pumps will trip 
during a LOCA only if offsite power is unavailable. The licensee also 
proposed to editorially move a footnote on page 3.6-1 to its correct 
place on page 3.6-2.
    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 revision to Technical Specification (TS) Table 
3.1.1 to clarify the tripping of the Service Water (SW) and Reactor 
Building Closed Cooling Water (RBCCW) pumps documents the as-built 
controls for these loads. Amendment No. 42 to the Oyster Creek 
Licensing Application concluded that these pumps are not required to 
perform any functions related to safe plant shutdown. During a loss 
of coolant accident (LOCA) condition, with offsite power available, 
the plant electrical busses have enough capacity and capability to 
supply the SW and RBCCW pumps. This proposed change is an 
administrative change only, and is being made to align the Oyster 
Creek Technical Specifications with the design of the plant. No 
physical changes are being made to the plant. Also, the footnote on 
TS page 3.6-1 would be relocated to TS page 3.6-2 to appear on the 
same TS page as the Specification to which it applies. The proposed 
changes do not alter the physical design or operational procedures 
associated with any plant structure, system, or component.
    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

[[Page 15942]]

accident from any accident previously evaluated?
    Response: No.
    The proposed revision to Technical Specification Table 3.1.1 to 
clarify the tripping of the SW and RBCCW pumps documents as-built 
controls for these loads. These pumps provide cooling to various 
non-safety related plant equipment. Following a LOCA condition, with 
offsite power available, these pumps will help in removing plant 
heat loads. This clarification that the SW and RBCCW pumps do not 
trip during a LOCA, with offsite power available, does not affect 
the Emergency Diesel Generator time delayed loading sequence. The 
relocation of the footnote applicable to Specification 3.6.A.4.1 is 
editorial in nature and has no impact on any accident previously 
evaluated. Accordingly, the proposed changes do not introduce any 
new accident initiators, nor do they reduce or adversely affect the 
capabilities of any plant structure or system in the performance of 
their safety function.
    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.
    The proposed revision to Technical Specification Table 3.1.1 to 
clarify the tripping of the SW and RBCCW pumps documents as-built 
controls for these loads. The NRC Safety Evaluation Report (SER) for 
Amendment 42 to the Oyster Creek Licensing Application concluded 
that it is acceptable to automatically trip the SW and RBCCW pumps 
during a loss of coolant accident. The NRC SER for Technical 
Specification Amendment 60 concluded that the immediate tripping of 
the RBCCW pump and the time delayed tripping of the SW pumps during 
a LOCA was also acceptable. The clarification that the SW and RBCCW 
pumps do not trip during a loss of coolant accident when offsite 
power is available does not reduce any margin of safety because 
these pumps are not required to mitigate the consequences of any 
postulated accident. The relocation of the footnote applicable to 
Specification 3.6.A.4.1 is editorial in nature and has no impact on 
any accident 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: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: February 25, 2005.
    Description of amendment request: The proposed changes would amend 
Operating License DPR-65 for Millstone Power Station, Unit No. 2 (MPS2) 
and Operating License NPF-49 for Millstone Power Station, Unit No. 3 
(MPS3) by incorporating certain administrative changes into the MPS2 
and MPS3 Technical Specifications (TSs).
    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 changes are administrative in nature and do not 
alter any of the requirements of the affected TS[s]. The proposed 
changes do not modify any plant equipment and do not impact any 
failure modes that could lead to an accident. Additionally, the 
proposed changes have no effect on the consequence of any analyzed 
accident since the changes do not affect any equipment related to 
accident mitigation. Based on this discussion, the proposed 
amendment does not increase 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 changes are administrative in nature. They do not 
modify any plant equipment and there is no impact on the capability 
of the existing equipment to perform their 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 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.
    These changes are administrative in nature and do not alter any 
of the requirements of the affected TS[s]. The proposed changes do 
not affect any of the assumptions used in the accident analysis, nor 
do they affect any operability requirements for equipment important 
to plant safety. Therefore, the proposed changes will not result in 
a significant reduction in the margin of safety as defined in the 
bases for technical specifications covered in this license amendment 
request.

    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., Waterford, CT 06141-5127.
    NRC Section Chief: Darrell J. Roberts.

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: March 8, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.13, Primary Containment Leakage 
Rate Testing Program, for the Integrated Leak Rate Testing (ILRT) 
program to add an exception to the commitment to follow the guidelines 
of Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test 
Program.'' The effect of this request would be a one-time extension of 
the interval since the last ILRT from 15 years to 15 years and 4 months 
(i.e., from August 2007 to December 2007).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to TS 5.5.13 allows a one-time extension 
to the current interval for the ILRT. The current interval of 
fifteen years, based on past performance, would be extended on a 
one-time basis to 15-years and 4 months from the date of the last 
test. The proposed extension to the ILRT cannot increase the 
probability of an accident since there are no design or operating 
changes involved and the test is not an accident initiator. The 
proposed extension of the test interval does not involve a 
significant increase in the consequences since analysis has shown 
that, the proposed extension of the ILRT and DWBT [Drywell Bypass 
Test] frequency has a minimal impact on plant risk. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

[[Page 15943]]

    2. Will operation of the facility in accordance with this 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    The proposed extension to the interval for the ILRT does not 
involve any design or operational changes that could lead to a new 
or different kind of accident from any accidents previously 
evaluated. The tests are not being modified, but are only being 
performed after a longer interval. The proposed change does 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 change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    An evaluation of extending the ILRT DWBT surveillance frequency 
from once in 10 years to once in 15 years and 4 months has been 
performed using methodologies based on the approved ILRT 
methodologies. This evaluation assumed that the DWBT frequency was 
being adjusted in conjunction with the ILRT frequency. This analysis 
used realistic, but still conservative, assumptions with regard to 
developing the frequency of leakage classes associated with the ILRT 
and DWBT. The results from this conservative analysis indicates that 
the proposed extension of the ILRT frequency has a minimal impact on 
plant risk and therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    Based on the above, Entergy concludes that the proposed 
amendment(s) present no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    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: Allen G. Howe.

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

    Date of amendment request: December 20, 2004.
    Description of amendment request: Entergy Operations, Inc. is 
proposing that the Arkansas Nuclear One Unit 2 (ANO-2) Facility 
Operating License be amended to revise the requirements for ensuring 
containment structural integrity. The proposed changes modify the 
Containment Structural Integrity Technical Specification (TS) 3.6.1.5 
to delete the existing Surveillance Requirements (SR) and add a new SR 
to verify containment structural integrity in accordance with a new 
Containment Tendon Surveillance Program. A new Containment Tendon 
Surveillance Program is added to TS 6.5.6 and a new reporting 
requirement is being added to TS 6.6.6. The proposed changes are 
generally consistent with NUREG 1432, ``Standard Technical 
Specifications Combustion Engineering Plants,'' Revision 3. This 
request for amendment also contains proposed administrative changes 
related to page numbering.
    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:

    Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The containment building is not considered to be the initiator 
of any accident previously evaluated, but serves to mitigate 
accidents that could allow a release to the environment. The 
proposed TS change will provide for containment tendon inspections 
as required by 10 CFR 50.55a and prevent or inhibit release from the 
containment building as designed. Through appropriate inspections 
and implementation of corrective actions for any degradation 
discovered during the inspections that might lead to containment 
structural failures, the probability or consequences of accidents 
will not be increased.
    Criterion 2--Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed change does not change the design, configuration, 
or method of operation of the plant. By implementing corrective 
actions for any degradation discovered during the required 
inspections of the containment, the possibility of a new or 
different kind of accident will not be created. Implementation of 
the requirements of Subsection IWL of the ASME code [American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code] and 
those of 10 CFR 50.55a(b)(2) provide an equally acceptable 
containment inspection program.
    Criterion 3--Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed change to incorporate the applicable requirements 
of Subsection IWL of the ASME Code and of 10 CFR 50.55a(b)(2) into 
the ANO-2 containment inspection program has no impact on any safety 
analysis assumptions. The addition of structural integrity 
requirements to ANO-2 TS Specification 3.6.1.5 imposes consistent 
requirements with those previously specified in the ANO-2 TSs. The 
requirements of ASME IWL are more restrictive than those currently 
provided in the existing ANO-2 technical specifications. As a 
result, the margin of safety is not reduced by the proposed 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 are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Allen G. Howe.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: February 25, 2005.
    Description of amendment requests: The proposed amendments would 
modify the Technical Specifications by revising the near-end-of-life 
Moderator Temperature Coefficient (MTC) Surveillance Requirement by 
placing a set of conditions on core performance, which, if met, would 
allow conditional exemption from the required MTC measurement.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The probability or consequences of accidents previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR) are 
unaffected by this proposed change because there is no change to any 
equipment response or accident mitigation scenario. There are no 
additional challenges to fission product barrier integrity.
    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.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system.
    Therefore, the proposed change does not create the possibility 
of a new or different

[[Page 15944]]

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 associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
on the availability, operability, or performance of the safety-
related systems and components. A change to a surveillance 
requirement is proposed, but the limiting conditions for operation 
required by the Technical Specifications (TS) are not changed.
    The Technical Specifications Bases are founded in part on the 
ability of the regulatory criteria to be satisfied assuming the 
limiting conditions for operation are met for the various systems. 
Conformance to the regulatory criteria for operation with the 
conditional exemption from the near-end of life moderator 
temperature coefficient (MTC) measurement is demonstrated and the 
regulatory limits are not exceeded. Therefore, the margin of safety 
as defined in the TS is not reduced.
    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: James M. Petro, Jr., Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: March 8, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 2.1.1.2 for the single recirculation 
loop Safety Limit Minimum Critical Power Ratio (SLMCPR) value to 
reflect results of a cycle-specific calculation for Cycle 23 
operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Allen G. Howe.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: February 3, 2005.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications (TSs) by revising TS 6.16.b.1, 
``Radioactive Effluent Controls Program,'' to be consistent with the 
intent of 10 CFR 20 and NUREG-1431, ``Standard Technical Specifications 
Westinghouse Plants'' (STS).
    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?
    NMC [Nuclear Management Company, LLC] Response:
    No. Updating the specification to be consistent with 10 CFR 20 
and the STS has no impact on plant structures, systems, or 
components, does not affect any accident initiators, and does not 
change any safety analysis. Therefore, the changes do not involve an 
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?
    NMC Response:
    No. Updating the specification to be consistent with 10 CFR 20 
and the STS will not change any equipment, require new equipment to 
be installed, or change the way current equipment operates. No 
credible new failure mechanisms, malfunctions, or

[[Page 15945]]

accident initiators are created by the proposed changes. Therefore, 
the changes do 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?
    NMC Response:
    No. Updating the specification to be consistent with 10 CFR 20 
and the STS has no impact on inputs to the safety analysis or to 
automatic plant actions. It also does not impact plant equipment or 
operation. Therefore, the change does not reduce 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: October 15, 2004.
    Description of amendment request: The proposed amendment revises TS 
5.5.6, ``Reactor Coolant Pump Flywheel Inspection Program,'' to extend 
the allowable inspection interval to 20 years.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on October 22, 2003 (68 FR 60422). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated October 15, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    The proposed change does not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, or configuration of the facility, or the manner in which 
the plant is operated and maintained; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.

    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: September 27, 2004.
    Description of amendment request: The proposed amendment would 
revise the reactor coolant pump (RCP) flywheel inspection surveillance 
requirements to extend the allowable inspection interval to 20 years.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on October 22, 2003 (68 FR 60422). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated September 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:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the

[[Page 15946]]

bounding plant configuration case, the risk is still acceptably low.
    The proposed change does not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, or configuration of the facility, or the manner in which 
the plant is operated and maintained; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

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

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

    Docket No. 50-354, Hope Creek Generating Station, Salem County, New 
Jersey Date of amendment request: January 11, 2005. Description of 
amendment request: The proposed amendment would delete the Technical 
Specification (TS) requirements to submit monthly operating reports and 
occupational radiation exposure reports.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in licensing amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated January 11, 2005.
    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 eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating report 
of shutdown experience and operating statistics if the equivalent 
data is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC regulations. The proposed change involves no 
changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

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

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

    Date of amendment request: February 15, 2005.
    Description of amendment request: The proposed amendment will 
revise the Salem, Unit Nos. 1 and 2 Technical Specifications to reflect 
the deletion of Reactor Coolant System (RCS) volume from design 
features Section 5.4.2. This design feature information will continue 
to be maintained in the plant's updated final safety analysis report.
    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 remove this information from T/S 
[technical specifications] does not affect any accident initiators 
or precursors. Elimination of the RCS volume information from the T/
S does not change the methods for plant operation or actions to be 
taken in the event of an accident. The quantity of radioactive 
material available for release in the event of an accident is not 
increased.
    Barriers to release of radioactive material are not eliminated 
or otherwise changed. More detailed RCS component and piping volume 
information is included in the Salem UFSAR [updated final safety 
analysis report], and changes to that information would be evaluated 
prior to implementation in accordance with 10 CFR 50.59.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of accidents 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.

[[Page 15947]]

    The deletion of the RCS volume information from the T/S does not 
change the methods of plant operation or modify plant systems, 
structures, or components. No new methods of plant operation are 
created. As such, the proposed change does not affect any accident 
initiators or precursors or create new accident initiators or 
precursors. More detailed and complete RCS component and piping 
volume information is included in the Salem UFSAR, and any changes 
to that information would be evaluated prior to implementation in 
accordance with 10 CFR 50.59.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The deletion of the RCS volume information from the T/S does not 
affect safety limits or limiting safety system settings. Plant 
operational parameters are not affected. The proposed change does 
not modify the quantity of radioactive material available for 
release in the event of an accident. As such, the change will not 
affect any previous safety margin assumptions or conditions. The 
actual volume of the RCS is not affected by the change, only the 
location of the text describing the volume. More detailed and 
complete RCS component and piping volume information is included in 
the Salem UFSAR, and any changes to that information would be 
evaluated prior to implementation in accordance with 10 CFR 50.59.

    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: Darrell J. Roberts.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of amendment request: January 24, 2005.
    Description of amendment request: The proposed license amendment 
removes unnecessary and obsolete information from the facility license. 
The proposed changes are editorial and administrative in nature and 
will remove inappropriate and unnecessary information from the license 
given that the facility is permanently shutdown and defueled.
    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.
    No. The proposed changes are administrative and involve deleting 
unnecessary and obsolete information from the facility operating 
license. These changes do not affect possible initiating events for 
accidents previously evaluated or alter the configuration or 
operation of the facility. Safety limits, limiting safety system 
settings, and limiting control systems are no longer applicable to 
Rancho Seco in the permanently defueled mode, and are therefore not 
relevant.
    The proposed changes do not affect the boundaries used to 
evaluate compliance with liquid or gaseous effluent limits, and have 
no impact on plant operations. Therefore, the proposed license 
amendment 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 accident from any previously evaluated.
    No. As described above, the proposed changes are administrative. 
The safety analysis for the facility remains complete and accurate. 
There are no physical changes to the facility and the plant 
conditions for which the design basis accidents have been evaluated 
are still valid.
    The operating procedures and emergency procedures are not 
affected. The proposed changes do not affect the emergency planning 
zone, the boundaries used to evaluate compliance with liquid or 
gaseous effluent limits, and have no impact on plant operations. 
Consequently, no new failure modes are introduced as the result of 
the proposed changes. Therefore, the proposed changes will 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.
    No. As described above, the proposed changes are administrative. 
There are no changes to the design or operation of the facility. The 
proposed changes do not affect the emergency planning zone, the 
boundaries used to evaluate compliance with liquid or gaseous 
effluent limits, and have no impact on plant operations. 
Accordingly, neither the design basis nor the accident assumptions 
in the Defueled Safety Analysis Report (DSAR), nor the Technical 
Specification Bases are affected. Therefore, the proposed changes do 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's significant hazards 
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: Arlen Orchard, Esq., General Counsel, 
Sacramento Municipal Utility District, 6201 S Street, P.O. Box 15830, 
Sacramento, CA 95817-1899.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: August 16, 2004 (TS-433).
    Description of amendment request: The proposed amendment extends 
the frequency of ``once-per-cycle'' from 18 months to 24 months in 
several Technical Specification Surveillance Requirements. This change 
will allow the adoption of a 24-month refueling cycle.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed amendment changes the surveillance frequency 
from 18 months to 24 months for Surveillance Requirements in the 
Unit 1 Technical Specification[s] that are normally a function of 
the refueling interval. Under certain circumstances, Surveillance 
Requirement 3.0.2 would allow a maximum surveillance interval of 30 
months for these surveillances. TVA's evaluations have shown that 
the reliability of protective instrumentation and equipment will be 
preserved for the maximum allowable surveillance interval. The 
proposed changes do not involve any change to the design or 
functional requirements of plant systems and the surveillance test 
methods will be unchanged. The proposed changes will not give rise 
to any increase in operating power level, fuel operating limits, or 
effluents. The proposed change does not affect any accident 
precursors. In addition, the proposed changes will not significantly 
increase any radiation levels. Based on the foregoing considerations 
and the evaluations completed in accordance with the guidance of 
Generic Letter 91-04, it is concluded that the proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed amendment does not require a change to the 
plant design, nor the mode of plant operation. The proposed changes 
do not create the possibility of any

[[Page 15948]]

new failure mechanisms. No new external threats or release pathways 
are created. Therefore, the implementation of the proposed amendment 
will not create a possibility for an accident of a new or different 
type than those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed amendment changes the surveillance frequency 
from 18 months to 24 months for Surveillance Requirements in the 
Unit 1 Technical Specification[s] that are normally a function of 
the refueling interval. Under certain circumstances, Surveillance 
Requirement 3.0.2 would allow a maximum surveillance interval of 30 
months for these surveillances. Although the proposed Technical 
Specification changes will result in an increase in the interval 
between surveillance tests, the impact on system availability is 
small based on other, more frequent testing or redundant systems or 
equipment. There is no evidence of any failures that would impact 
the availability of the systems. This change does not alter the 
existing setpoints, Technical Specification allowable values or 
analytical limits. The assumptions in the current safety analyses 
are not impacted and the proposed amendment does not reduce a margin 
of safety. Therefore, the proposed license amendment 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear