[Federal Register Volume 69, Number 225 (Tuesday, November 23, 2004)]
[Notices]
[Pages 68180-68193]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-25664]
[[Page 68180]]
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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 October 29, 2004, through November 12, 2004.
The last biweekly notice was published on November 9, 2004 (69 FR
64984).
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. (Note:
Public access to ADAMS has been temporarily suspended so that security
reviews of publicly available documents may be performed and
potentially sensitive information removed. Please check the NRC Web
site for updates on the resumption of ADAMS access.) The filing of
requests for a hearing and petitions for leave to intervene is
discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. (Note: Public access to ADAMS has been
temporarily suspended so that security reviews of publicly available
documents may be performed and potentially sensitive information
removed. Please check the NRC Web site for updates on the resumption of
ADAMS access.) If a request for a hearing or petition for leave to
intervene is filed within 60 days, the Commission or a presiding
officer designated by the Commission or by the Chief Administrative
Judge of the Atomic Safety and Licensing Board Panel, will rule on the
request and/or petition; and the Secretary or the Chief Administrative
Judge of the Atomic Safety and Licensing Board will issue a notice of a
hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases
[[Page 68181]]
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, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to 301-415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. (Note: Public access to ADAMS has been
temporarily suspended so that security reviews of publicly available
documents may be performed and potentially sensitive information
removed. Please check the NRC Web site for updates on the resumption of
ADAMS access.) If you do not have access to ADAMS or if there are
problems in accessing the documents located in ADAMS, contact the NRC
PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: October 20, 2004.
Description of amendment request: The proposed change would revise
Technical Specification (TS) Table 4.1-1 functional testing
surveillance interval from monthly to semi-annually for the following
reactor protection system instrument channels: Table 4.1-1, Item No. 4,
``Power Range Channel,'' Item No. 7, ``Reactor Coolant Temperature
Channel,'' Item No. 8, ``High Reactor Coolant Pressure Channel,'' Item
No. 9, ``Low Reactor Coolant Pressure Channel,'' Item No. 10,'' Flux-
Reactor Coolant Flow Comparator,'' Item No. 11, ``Reactor Coolant
Pressure-Temperature Comparator,'' Item No. 12, ``Pump Flux
Comparator,'' Item No. 13, ``High Reactor Building Pressure Channel,''
Item No. 45, ``Loss of Feedwater Reactor Trip,'' and Item No. 46,
``Turbine Trip/Reactor Trip.'' The TS Section 4.1 Bases would be
revised to reflect the proposed change from monthly to semi-annually
and to specify that one channel is being tested every 46 days on a
continual sequential rotation, which is consistent with the
calculations of BAW-10167A, Supplement 1, and associated Nuclear
Regulatory Commission Safety Evaluation Report that indicate that the
reactor protection system retains a high level of reliability for this
test interval. The proposed change would also revise TS Table 4.1-1
functional testing surveillance interval from monthly to quarterly for
the following reactor protection system reactor trip devices: Table
4.1-1, Item No. 1, ``Protection Channel Coincidence Logic,'' and Item
No. 2, ``Control Rod Drive Trip Breaker and Regulating Rod Power
SCRs.'' The TS Section 4.1 Bases would be revised to reflect the
proposed change from monthly to quarterly testing and to specify that
one channel is being tested every 23 days on a continual sequential
rotation, which is consistent with the calculations of BAW-10167A,
Supplement 3, February 1998, and the NRC SER for BAW-10167A, Supplement
3, dated January 7, 1998, that indicate that the reactor trip system
retains a high level of reliability for this test interval.
Basis for proposed valuated 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 reactor protection system monitors parameters related to
safe operation and trips the reactor to protect the reactor core
against fuel cladding damage. It also assists in protecting against
reactor coolant system damage caused by high system pressure by
limiting energy input to the system through reactor trip action.
Therefore, this change has no impact on the probability of an
accident previously evaluated. The results of the reliability
analyses conducted in accordance
[[Page 68182]]
with NRC [Nuclear Regulatory Commission] approved methodology and
criteria show that the test interval extension of the reactor
protection system instrument channels and reactor trip devices is
not a significant contributor to trip system unavailability or the
risk of core damage. The reactor protection system instrument
channel and reactor trip device functional test surveillance program
will continue to ensure that the reactor protection system is
capable of performing its intended safety function during a design
basis accident.
Therefore, this change has no effect on the consequences of an
accident previously evaluated.
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 involves the reactor protection system
instrument channel and reactor trip device surveillance test
interval, which is not, in and of itself, considered to be an
accident initiator. Postulated failure of the reactor protection
system instrument channel or reactor trip device to function is an
analyzed condition and does not constitute a new or different kind
of accident. The proposed change does not create any new failure
modes not bounded by previously analyzed accidents.
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 results of the reliability analysis conducted in accordance
with NRC approved methodology and criteria show that the test
interval extension of the reactor protection system instrument
channels and reactor trip devices is not a significant contributor
to trip system unavailability or the risk of core damage. The
Technical Specifications will continue to require the reactor
protection system trip setpoints to remain within the assumptions of
the accident analysis and that adequate reliability of the reactor
protection system trip devices is maintained, thus preserving
existing margins of safety.
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: Thomas S. O'Neill, Associate General
Counsel, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville,
IL 60555.
NRC Section Chief: Richard J. Laufer.
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: August 20, 2004.
Description of amendment request: The proposed amendment would
revise the Allowable Values for the following Reactor Protection System
(RPS) instrumentation functions: Intermediate Range Neutron Flux,
Reactor Coolant Flow--Low, Steam Generator Water Level--Low Coincident
with Steam Flow/Feedwater Flow Mismatch, and Intermediate Range Neutron
Flux (P-6) Interlock. Additionally, these changes revise the Allowable
Value for the Engineered Safety Feature Actuation System
Instrumentation function for High Steam Flow in Two Steam Lines
Coincident with Steam Line Pressure--Low. Also the proposed amendment
would delete an unnecessary footnote associated with the applicability
for the Automatic Trip Logic RPS instrumentation function.
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?
The proposal to revise the Allowable Values for the affected
reactor protection and engineered safety feature actuation functions
was developed in accordance with the current setpoint methodology
for HBRSEP [H. B. Robinson Steam Electric Plant], Unit No. 2, thus
ensuring that the probability and consequences of previously
evaluated accidents are not significantly increased. The proposed
deletion of the unnecessary footnote associated with the Automatic
Trip Logic reactor protection instrumentation function does not
change the requirements for operability of this function. Therefore,
the proposed changes do 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?
The proposed changes will continue to ensure that the
operability of the previously described functions will be
appropriately maintained. No physical changes to the HBRSEP, Unit
No. 2, systems, structures, or components are being implemented.
There are no new or different accident initiators or sequences being
created by the proposed Technical Specifications changes. Therefore,
these changes do 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?
The proposed changes, as previously described, ensure that the
margin of safety for the applicable fission product barriers that
are protected by these functions will continue to be maintained.
This conclusion is based on the use of a valid setpoint methodology
for determining the Allowable Values for the reactor protection and
engineered safety feature actuation functions. Therefore, these
changes do not involve a significant reduction in the margin of
safety.
Based on the preceding discussion, the requested changes do not
involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina; Docket
Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1and 2,
Mecklenburg County, North Carolina; Docket Nos. 50-269, 50-270, and 50-
287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: September 28, 2004.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated September 28, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
[[Page 68183]]
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated?
The proposed change eliminates the TS 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.
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 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.
Criterion 3--The proposed change does not involve a significant
reduction in a margin of safety?
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: October 5, 2004.
Description of amendment request: The proposed amendments would
revise Technical Specifications (TSs) 3/4.3.1, ``Reactor Trip System
Instrumentation,'' and 3/4.3.2, ``Engineered Safety Feature Actuation
System Instrumentation,'' to modify steam generator (SG) level
allowable value setpoints. The proposed changes address recent generic
issues involving new SG level uncertainty considerations and margins
associated with Westinghouse-designed SGs.
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?
No. The SG water level-low-low setpoint and allowable value have
been revised to address Westinghouse Nuclear Safety Advisory Letter
NSAL-03-9 and other considerations on steam generator water level
uncertainties. The revised setpoint and allowable value calculations
continues to follow the setpoint methodology previously approved for
BVPS Unit No. 1 and No. 2 while addressing newly identified level
uncertainty considerations. The proposed changes to the SG water
level-low-low Allowable Value for BVPS Unit No. 1 and No. 2 and to
the SG water level-high-high Allowable Value for BVPS Unit No. 2
continue [to] maintain the validity of the safety analysis limits
used in the safety analyses that credit the actuations based on SG
water level.
The proposed changes do not alter the causes for any accident
described in the Updated Final Safety Analysis Report (UFSAR) that
credit the SG water level setpoint actuations. Therefore, they do
not involve a significant increase in the probability of an accident
previously evaluated.
The proposed changes do not alter the accident analyses that
credit the SG water level-low-low setpoint actuation or the
associated accident acceptance criteria. Therefore, they do not
involve a significant increase in the consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The SG water level-low-low setpoint and allowable value have
been revised to address Westinghouse Nuclear Safety Advisory letter
NSAL-03-9 and other considerations on steam generator water level
uncertainties. Implementation of the proposed setpoint changes have
no significant effect on either the configuration of the plant, or
the manner in which the plant is operated. The proposed changes to
the SG water level-low-low allowable value for BVPS Unit No. 1 and
No. 2 and to the SG water level-high-high allowable value for BVPS
Unit No. 2 continue to maintain the validity of the safety analysis
limits used in the safety analyses that credit the actuations based
on SG water level.
Therefore, since the plant configuration is not adversely
changed and the proposed changes do not alter the accident analyses
that credit actuation based on SG water level, 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?
No. The Reactor Trip System and Engineered Safety Feature
Actuation System setpoint analysis methodology and acceptance
criteria provide the margin of safety. The SG water level-low-low
and SG water level-high-high actuation setpoint and allowable value
have been calculated using the same methodology as previously
approved for the BVPS Unit No. 1 and No. 2 while addressing newly
identified considerations needed to protect the limits used in the
safety analyses. The applicable safety analyses have been performed
and show acceptable results. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 25, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification 2.1.1.2 for the dual recirculation loop
and single recirculation loop Safety Limit Minimum Critical Power Ratio
(SLMCPR) values to reflect results of a cycle specific calculation.
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
[[Page 68184]]
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 (SLMCPR) for Cooper Nuclear Station 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 [Nebraska Public Power District]
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 R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Acting Section Chief: Michael K. Webb.
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: April 15, 2004.
Description of amendment request: The proposed change would modify
the Salem Updated Final Safety Analysis Report (UFSAR) with respect to
fire protection requirements for the 4160 Volt Switchgear Rooms, 460
Volt Switchgear Rooms, and the Lower Electrical Penetration Area Rooms.
Specifically, the amendment would reduce the UFSAR description of the
Carbon Dioxide Tank volume from being able to provide two full
discharges to an affected room to one full and one partial discharge to
an affected room. Additionally, the assumed ability of the Carbon
Dioxide system would be reduced from an ability to produce a
CO2 concentration of 50% for 30 minutes to an ability to
produce a CO2 concentration of 27.6% for a length of time
sufficient to suppress a fire and allow the PSEG Nuclear Fire
Department to respond.
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 likelihood of a fire event is not increased since the
proposed change does not alter the fire hazards contained in the
plant. The ability to achieve and maintain safe shutdown in the
event of a fire is not impacted by the reduction of CO2
concentration, since the Fire Brigade will respond in ample time and
extinguish a fire using alternate means. In addition, the proposed
changes to the UFSAR would not change any response to a fire event.
Also, the probability of occurrence or the consequences for an
accident or malfunction of equipment is not increased by the
proposed changes since the response to a fire event would not change
and the fire brigade would continue to respond rapidly to any fires
or fire alarms. Further, the proposed changes do not alter the way
any structure, system, or component (SSC) functions, do not modify
the manner in which the plant is operated, and do not significantly
alter equipment out-of-service time. Changing the CO2
concentration requirement in the 4160 Volt Switchgear Rooms, 460
Volt Switchgear Rooms and Lower Penetration Area Rooms at Salem
Units 1 and 2 does not change the probability or consequences of any
accident and dose consequences are unaffected. No changes to the
design of structures, systems, or components (SSC) are made and
there are no effects on accident mitigation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence 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 possibility of a new or different kind of accident from any
accident or malfunction in the Salem Updated Final Safety Analysis
Report (UFSAR) is not created. The design basis event applicable to
the proposed change is a fire in the 4160 Volt Switchgear Rooms, 460
Volt Switchgear Rooms and Lower Penetration Area Rooms at Salem
Units 1 and 2. Therefore a different accident is not created. In
addition, the proposed changes cannot initiate an accident. Further,
the proposed changes to the UFSAR do not change the design function
or operation of any SSCs.
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 change involve a significant reduction in a margin
of safety?
Response: No.
The reduction in CO2 concentration provides ample
response time for the onsite dedicated fire brigade to respond to a
fire event and a 20% safety factor in CO2 concentration
remains. The proposed changes do not affect the ability to safely
shutdown and maintain the shutdown conditions of either unit
following a fire in the affected areas. The proposed changes do not
rely on compensatory measures or actions deviating from the
licensing or design basis. In addition, the proposed changes do not
change the margin of safety since no SSCs are changed. The results
of accident analysis remain unchanged by the proposed changes to the
UFSAR.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 68185]]
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: James W. Clifford.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: September 17, 2004.
Description of amendment request: The amendment is to support the
replacement of the steam generators (SGs) at Callaway during the
refueling outage in the Fall of 2005. The amendment would (1) change
the affected technical specifications (TSs) such as the reactor core
safety limits (TS 2.1.1), reactor trip system (RTS) and engineered
safety feature actuation system (ESFAS) instrumentation (TSs 3.3.1 and
3.3.2), reactor coolant system (RCS) limits (TS 3.4.1), RCS loops (TSs
3.4.5, 3.4.6, and 3.4.7), RCS operational leakage (TS 3.4.13), SG tube
integrity (new TS 3.4.17), main steam safety valves (TS 3.7.1), SG
surveillance program (TS 5.5.9), containment integrated leakage rate
testing (ILRT) program (TS 5.5.16), and SG inspection report (TS
5.6.10); (2) revise the affected transient analyses such as excessive
increase in secondary steam flow event, loss of normal feedwater event,
transient mass and energy releases, radiological consequences of
associated events, and containment pressure/temperature responses; and
(3) revise nuclear steam and supply system (NSSS) design parameters and
transients, and fatigue usage factors and stresses for the replacement
SGs. The amendment involves the following areas of change to the
license: nuclear steam supply system evaluations for the replacement
steam generators, trip time delay (TTD) elimination for certain RTS and
ESFAS functions, the SG surveillance program in Technical Specification
Task Force (TSTF) No. 449 (TSTF-449), and the post-modification
containment ILRT exception.
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 for the above areas of review, which is presented below
(with the terms defined in the plant Technical Specifications
capitalized):
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Nuclear Steam Supply System Evaluations for Replacement Steam
Generators
As discussed in the NSSS Licensing Report (Appendix A to this
amendment application), all acceptance criteria continue to be met.
All major NSSS components (e.g., Reactor Vessel, Pressurizer, RCPs
[(reactor coolant pumps)], Steam Generators, etc.) have been
assessed with respect to bounding conditions expected for
replacement steam generator (RSG) conditions. In all cases operation
has been found to be acceptable. Major systems and subsystems (e.g.,
safety injection, RHR [residual heat removal], etc.) have been
reviewed and acceptable performance has been verified for their
normal operation and, as applicable, for their safety-related
functions. All reactor trip and ESFAS actuation setpoints have been
assessed, and the proposed setpoint modifications will assure
adequate protection is afforded for all design basis events.
The reactor core safety limits have been revised based on the
RSG project parameters. All of the acceptance criteria for the
accident analyses (e.g., DNBR [departure from nucleate boiling
ratio] limits, fuel centerline temperatures, etc.) continue to be
met with the revised safety limit lines. Therefore, the revised core
safety limit line changes are acceptable. The proposed changes to
the reactor core safety limits will not initiate any accidents;
therefore, they do not increase the probability of an accident
previously evaluated in the FSAR [Callaway Final Safety Analysis
Report]. The comprehensive analytical efforts performed to support
the proposed RSG conditions include a reanalysis or evaluation of
all accident analyses that are impacted by the revised reactor core
safety limits.
The changes in various SG-related RTS and ESFAS Allowable Values
have resulted from the analyses performed to support plant operation
at the proposed RSG conditions. Setpoint uncertainty calculations
confirm the acceptability of these revised Allowable Values. The
affected RTS and ESFAS Allowable Values have been modified to
reflect the results of updated setpoint calculations based on plant-
specific uncertainties, calibration practices, calibration
equipment, and installed hardware and procedures. The Allowable
Values were calculated using the same Westinghouse setpoint
methodology used for the current trip setpoints, but improved in a
conservative fashion to include refinements that better reflect
plant calibration practices and equipment performance. These
refinements include the incorporation of a sensor reference accuracy
term to address repeatability effects when performing a single pass
calibration (i.e., one up and one down pass at several points
verifies linearity and hysteresis, but not repeatability). In
addition, sensor and rack error terms for calibration accuracy and
drift are grouped in the Channel Statistical Allowance equation with
their dependent measurement and test equipment (M&TE) terms, then
combined with the other independent error terms using the square
root sum of the squares (SRSS) methodology. This improved setpoint
methodology has been previously review[ed] and approved by the NRC.
The proposed RTS and ESFAS Allowable Value changes will not initiate
any accidents; therefore, they do not increase the probability of an
accident previously evaluated in the FSAR. The comprehensive
analytical effort performed to support the proposed RSG conditions
included a reanalysis or evaluation of all accident analyses that
are impacted by the revised RTS and ESFAS Allowable Values. All
systems will function as designed.
The decrease in the Maximum Allowable Power for 3 OPERABLE MSSVs
[main steam safety valves] per SG from < 49% of Rated Thermal Power
to < 45% of Rated Thermal Power resulted from the analyses and
evaluations performed to support plant operation at the proposed RSG
conditions. The accident analysis acceptance criteria continue to be
met with these changes. These proposed plant system changes do not
increase the probability of an accident previously evaluated in the
FSAR. The comprehensive analytical effort performed to support the
proposed RSG conditions has included a review and evaluation of all
components and systems (including interface systems and control
systems) that could be affected by this change. All systems will
function as designed. The change in the manner in which the Reactor
Coolant Flow--Low Allowable Value is defined (while retaining the
same numerical value), the change in the manner in which RCS average
temperature is defined and the reduced upper limit for nominal T-avg
[average temperature] at full power conditions in the
Overtemperature [Delta]T [delta temperature] and Overpower [Delta]T
setpoint equations, and the changes to the pressurizer pressure and
RCS average temperature limits in the DNB LCO [departure from
nucleate boiling limiting condition for operation] [TS] 3.4.1 have
also been evaluated. None of these proposed changes will initiate
any accidents; therefore, the probability of an accident has not
been increased.
The potential dose consequences have been analyzed with respect
to the above changes collectively. The dose increases are less than
minimal (i.e., <10% of the margin between the regulatory limits and
the currently reported doses). The applicable dose acceptance
criteria continue to be met.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Trip Time Delay Elimination
This design change will eliminate only the Trip Time Delay
portion of the SG Water Level Low-Low trip functions and return that
portion of the design to condition that existed prior to Callaway
Amendment 43 dated April 14, 1989. The coincidence logic in the
Solid State Protection System will be unaffected. In all other
regards, the design of the RTS and ESFAS instrumentation will be
unaffected. These protection systems will continue to function in a
manner consistent with the plant design basis. All design, material,
and construction standards that were applicable prior to this
amendment request are maintained.
[[Page 68186]]
The probability and consequences of accidents previously
evaluated in the FSAR are not adversely affected because the removal
of the trip time delay circuitry assures a faster response by the
affected trip functions, consistent with the safety analysis
acceptance criteria and the original plant licensing basis.
The proposed change will not affect the probability of any event
initiators. There will be no degradation in the performance of, or
an increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance.
The proposed change will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in
the FSAR.
Therefore, the proposed TTD elimination does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
TSTF-449 Generic Licensing Change Package
This proposed change requires a Steam Generator Program that
includes performance criteria that will provide reasonable assurance
that the steam generator (SG) tubing will retain integrity over the
full range of operating conditions (including startup, operation in
the power range, hot standby, cooldown, and all anticipated
transients included in the design specification). The SG performance
criteria are based on tube structural integrity, accident induced
leakage, and operational LEAKAGE.
A steam generator tube rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis cases for the SGTR event at Callaway Plant, a
primary to secondary LEAKAGE rate of 1 gallon per minute (gpm) to
the unaffected SGs is assumed, in excess of the RCS Operational
LEAKAGE rate limit in TS 3.4.13, and the LEAKAGE rate associated
with a double-ended rupture of a single tube in the ruptured SG is
also assumed. For other design basis accidents such as main steam
line break (MSLB), rod ejection, and reactor coolant pump locked
rotor, the SG tubes are assumed to retain their structural integrity
(i.e., they are assumed not to rupture). These additional analyses
for Callaway Plant assume, as an initial condition, that primary to
secondary LEAKAGE for all SGs is 1 gpm. The accident induced leakage
criterion introduced by the proposed change to TS 5.5.9 accounts for
tubes that may leak during design basis accidents. The accident
induced leakage criterion limits this leakage to no more than the 1
gpm value assumed in the accident analyses.
The SG performance criteria added to TS 5.5.9 identify the
standards against which tube integrity is to be measured. Meeting
the performance criteria provides reasonable assurance that the SG
tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design basis accident. The performance criteria are only a part of
the Steam Generator Program required by the proposed change to TS
5.5.9. The program, defined by NEI [Nuclear Energy Institute] 97-06,
Steam Generator Program Guidelines, includes a framework that
incorporates a balance of prevention, inspection, evaluation,
repair, and leakage monitoring.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in TS 3.4.13 for RCS Operational
leakage and in TS 3.4.16 for DOSE EQUIVALENT I-131 in the primary
coolant to ensure the plant is operated within its analyzed
condition. The radiological consequence analyses at Callaway Plant
assume that the primary to secondary LEAKAGE rate is 1 gpm (more
conservative than the limit in TS 3.4.13), and that the reactor
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS
3.4.16 limits.
The proposed TSTF-449 changes reflect the design of the
replacement SGs, but do not affect their method of operation or
primary or secondary coolant chemistry controls. The proposed
changes update the TS and enhance the requirements for SG
inspections. The proposed changes do not adversely impact the
conclusions of any previously evaluated design basis accident and
are an improvement over the existing TS.
Therefore, this proposed change to implement TSTF-449 does not
affect the consequences of a SGTR accident and the probability of
such an accident is reduced. In addition, this proposed change does
not affect the consequences of an MSLB, rod ejection, reactor
coolant pump locked rotor, or any other accident event involving the
potential release of radioactive fluids from the secondary side of
Callaway Plant. [Therefore, this proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.]
Post-Modification ILRT Exception
This proposed change would provide Callaway Plant with an
exception from performing a post-modification containment integrated
leak rate test following the replacement of the steam generators
during Refuel [Outage] 14.
Integrated leak rate tests are performed to assure the leak-
tightness of the primary containment boundary system, and as such
they are not accident initiators. Therefore, not performing an
integrated leak rate test will not affect the probability of an
accident previously evaluated. The intent of post-modification
integrated leak rate testing requirements is to assure the leak-
tight integrity of the area affected by the modification. For the
Callaway Plant steam generator replacement modification, this intent
will be satisfied by performing the American Society of Mechanical
Engineers code required inspections and tests. Since the leak-
tightness integrity of the primary containment boundary affected by
the steam generator replacement will be assured, there is no change
in the containment boundary's ability to confine radioactive
materials during an accident. Therefore, adding a Technical
Specification exception from the steam generator replacement post-
modification integrated leak rate testing requirements does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Nuclear Steam Supply System Evaluations for Replacement Steam
Generators
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this amendment. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this amendment.
This amendment does not alter the safe performance of the plant
protection systems to trip the reactor when necessary or actuate ESF
[engineered safety feature] systems.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Trip Time Delay Elimination
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this amendment. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this amendment.
This amendment does not alter the safe performance of the plant
protection systems to trip the reactor when necessary or actuate ESF
systems.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
TSTF-449 Generic Licensing Change Package
The proposed performance based requirements are an improvement
over the requirements imposed by the existing TS.
Implementation of the proposed Steam Generator Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the Steam Generator Program will be
an enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
This proposed change does not impact the method of SG operation
or primary or secondary coolant chemistry controls. In addition,
this proposed change does not impact any other plant system or
component. The change enhances SG inspection requirements.
Therefore, this proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Post-Modification ILRT Exception
The proposed change would provide Callaway Plant with an
exception from performing a post-modification containment
[[Page 68187]]
integrated leak rate test following the replacement of the steam
generators during Refuel 14. Providing an exception from performing
a test does not involve a physical change to the plant nor does it
change the operation of the plant. Thus it cannot introduce a new
failure mode. Therefore adding a Technical Specification requirement
that provides an exception from the steam generator replacement
post-modification integrated leak rate testing requirement does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Nuclear Steam Supply System Evaluations for Replacement Steam
Generators
The analyses and evaluations supporting the proposed RSG
conditions reflect the reactor core safety limits. All acceptance
criteria continue to be met.
The analyses supporting the proposed RSG conditions reflect the
proposed RTS and ESFAS Allowable Values. Setpoint calculations
demonstrate that margin exists between these Allowable Values and
the corresponding safety analysis limits used in the RSG analyses.
The calculations are based on plant instrumentation and calibration/
functional test methods and include allowances for the RSG
conditions. All analyses and evaluations supporting the proposed RSG
core safety limits, decrease in maximum allowable power level for 3
operable MSSVs per SG, the change in the manner in which the Reactor
Coolant Flow--Low Allowable Value is defined (while retaining the
same numerical value), the change in the manner in which RCS average
temperature is defined and the reduced upper limit for nominal T-avg
at full power conditions in the Overtemperature [Delta]T and
Overpower [Delta]T setpoint equations, and the changes to the
pressurizer pressure and RCS average temperature limits in the DNB
LCO [TS] 3.4.1 are acceptable. All acceptance criteria continue to
be met. Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
Trip Time Delay Elimination
This proposed change does not eliminate any RTS or ESFAS
surveillances or alter the frequency of those surveillances as
required by the TS. The SG Water Level Low--Low safety analysis
limit of 0% span assumed in the analyses supporting the approval of
the TTD design in Callaway Amendment 43 dated April 14, 1989 is also
used in the RSG analyses discussed above. None of the acceptance
criteria for any accident analysis is changed for TTD elimination.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. The radiological dose consequence
acceptance criteria will continue to be met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
TSTF-449 Generic Licensing Change Package
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. This proposed
change to implement TSTF-449 does not, of itself, affect tube design
or operating environment. The proposed change is expected to result
in an improvement in the tube integrity by implementing the Steam
Generator Program to manage SG tube inspection, assessment, repair
(only under NRC-approved methods, none of which currently apply to
the RSGs), and plugging. The requirements established by the Steam
Generator Program are consistent with those in the applicable design
codes and standards and are an improvement over the requirements in
the existing TS.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by this proposed change.
Post-Modification ILRT Exception
The proposed change would provide Callaway Plant with an
exception from performing a post-modification containment integrated
leak rate test following the replacement of the steam generators
during Refuel 14. The intent of post-modification integrated leak
rate testing requirements is to assure the leak-tight integrity of
the area affected by the modification. This intent will be satisfied
by performing American Society of Mechanical Engineers code required
inspections and tests. The acceptance criterion for American Society
of Mechanical Engineers code system pressure testing for the base
metal and welds is no leakage. In addition, the test pressure for
the system pressure test will be several times that required during
an integrated leak rate test. Since the leak-tight integrity of the
primary containment boundary affected by the steam generator
replacement will be assured, there is no change in the primary
containment boundary's ability to confine radioactive materials
during an accident. Therefore, adding a Technical Specification
requirement that provides an exception from the steam generator
replacement post-modification integrated leak rate testing
requirements 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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Robert A. Gramm.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: September 15, 2004.
Description of amendment request: The proposed changes will change
the Administrative Controls Section of the Technical Specifications
(TS) in order to incorporate title changes, change the location where
the plant-specific titles and TS titles are correlated, and relocate
the unit staff requirements to the Quality Assurance Program. These
proposed changes will support the implementation of proposed Virginia
Electric and Power Company Topical Report DOM-QA-1, ``Nuclear Facility
Quality Assurance Program Description,'' currently under U.S. Nuclear
Regulatory Commission (NRC) staff review.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of North Anna Units 1 and 2 in accordance with the
proposed license amendments would not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change is administrative in nature and does not
affect plant systems, structures or components (SSCs) or plant
operation during normal or accident conditions. The proposed change
only affects the designated titles of personnel, the location of the
TS title and plant-specific title correlation, and the location of
the unit staff qualification requirements. Therefore, this change
has no bearing on the probability of an accident. Management
organizational structure and safety and operational reviews have not
changed and there is no change in the method of plant operation,
operation review, or system design review. As such, this change does
not alter the conclusions of the existing safety analyses and
therefore does not alter the consequences of an accident previously
evaluated.
2. Operation in accordance with the proposed license amendments
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed administrative change continues to ensure that
adequate management oversight exists at the plant in accordance with
the existing Technical
[[Page 68188]]
Specifications. The proposed change only affects the designated
titles of personnel, the location of the TS title and plant-specific
title correlation, and the location of the unit staff qualification
requirements. This change does not impact plant SSCs or plant
operation. Management organizational structure and safety and
operational reviews have not changed and there is no change in the
method of plant operation, operation review, or system design
review. There are no new or different accident scenarios, accident
initiators, nor failure mechanisms that will be introduced due to
this change. Therefore, the proposed change does not create the
possibility of an accident of a different type than evaluated
previously.
3. Operation in accordance with the proposed license amendments
would not involve a significant reduction in a margin of safety.
The proposed change only affects the designated titles of
personnel, the location of the TS title and plant-specific title
correlation, and the location of the unit staff qualification
requirements. This change does not impact plant design, plant
operation or any safety margin. Therefore, the proposed change does
not significantly reduce a margin of safety.
This evaluation concludes that the proposed amendments to the
North Anna Units 1 and 2 Technical Specifications do not involve a
significant increase in the probability or consequences of a
previously evaluated accident, do not create the possibility of a
new or different kind of accident and 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: Mary Jane Ross-Lee (Acting).
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 7, 2004.
Description of amendment request: The proposed change would revise
Technical Specification (TS) 5.3, ``Unit Staff Qualifications,'' to
reinstate the qualification requirements for the shift manager and
control room supervisor positions that were inadvertently eliminated
through Amendment No. 150. Also, TS 5.3 would be revised to reference
this amendment application for the use of the National Academy for
Nuclear Training guideline, ACAD 00-003, Revision 1, ``Guidelines for
Initial Training and Qualification of Licensed Operators.'' Various
other TSs would be revised to make corrections that were identified by
the NRC staff in its letter dated January 28, 2004, and additional
reviews performed by the licensee.
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.
Unit Staff Qualifications
The proposed change is an administrative change to reinstate the
qualification requirements for specific control room positions that
were inadvertently eliminated through the issuance of Amendment No.
150 and utilize Revision 1 to ACAD 00-003, ``Guidelines for Initial
Training and Qualification of Licensed Operators.'' The proposed
change does not directly impact accidents previously evaluated.
WCNOC's [Wolf Creek Nuclear Operating Corporation's] licensed
operator training program is accredited by the National Academy for
Nuclear Training and is based on a systems approach to training
consistent with the requirements of 10 CFR 55. Although licensed
operator qualifications and training may have an indirect impact on
accidents previously evaluated, the NRC considered this impact
during the rulemaking process, and by promulgation of the revised 10
CFR 55 rule, concluded that this impact remains acceptable as long
as the licensed operator training program is certified to be
accredited and is based on a systems approach to training.
Corrections
The proposed change involves corrections to the Technical
Specifications that are either associated with the issuance of the
Improved Technical Specifications (Amendment No. 123) or subsequent
amendments. The changes are considered administrative changes and do
not modify, add, delete, or relocate any technical requirements of
the Technical Specifications. As such, administrative changes do not
effect initiators of analyzed events or assumed mitigation of
accident or transient events.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Unit Staff Qualifications
The proposed change is an administrative change to reinstate the
current requirements of specific control room positions and allow
the use of Revision 1 of ACAD 00-003 for initial training and
qualification of licensed operators. WCNOC's licensed operator
training program is accredited by the National Academy for Nuclear
Training and is based on a systems approach to training consistent
with the requirements of 10 CFR 55. Although licensed operator
qualifications and training may have an indirect impact on accidents
previously evaluated, the NRC considered this impact during the
rulemaking process, and by promulgation of the revised 10 CFR 55
rule, concluded that this impact remains acceptable as long as the
licensed operator training program is certified to be accredited and
is based on a systems approach to training.
Corrections
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in methods of governing normal plant operation. The
proposed change will not impose any new or eliminate any old
requirements.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Unit Staff Qualifications
The proposed change is an administrative change to reinstate the
current requirements of specific control room positions and allow
the use of Revision 1 of ACAD 00-003 for initial training and
qualification of licensed operators. As noted previously, WCNOC's
licensed operator training program is accredited and is based on a
systems approach to training consistent with the requirements of 10
CFR 55. Licensed operator qualifications and training can have an
indirect impact on the margin of safety. However, the NRC considered
this impact during the rulemaking process, and by promulgation of
the revised 10 CFR 55 rule, determined that this impact remains
acceptable when licensees maintain a licensed operator training
program that is accredited and based on a systems approach to
training.
Corrections
The proposed change will not reduce a margin of safety because
they have no effect on any safety analysis assumptions. The change
is administrative in nature.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
[[Page 68189]]
NRC Section Chief: Robert Gramm.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: October 22, 2004.
Description of amendment request: The proposed amendment would
revise the allowed outage times of Technical Specification 3.3.3.6,
``Accident Monitoring Instrumentation,'' to be consistent with the
completion times in the related specification in NUREG-1431, Revision
3, ``Standard Technical Specifications Westinghouse Plants.''
Date of publication of individual notice in Federal Register:
November 2, 2004 (69 FR 63560).
Expiration date of individual notice: December 2, 2004 (public
comments) and January 3, 2005 (hearing requests).
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected]. (Note: Public access
to ADAMS has been temporarily suspended so that security reviews of
publicly available documents may be performed and potentially sensitive
information removed. Please check the NRC Web site for updates on the
resumption of ADAMS access.)
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: December 23, 2003, as
supplemented by letter dated June 16, 2004.
Brief description of amendment: The amendment revised Section 4.5.D
of the Technical Specifications to specify testing the main steam
isolation valves at a pressure lower than Pa, the calculated peak
containment internal pressure related to the design-basis loss-of-
coolant accident.
Date of Issuance: November 2, 2004.
Effective date: November 2, 2004 and shall be implemented within 30
days of issuance
Amendment No.: 250.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7518).
The June 16, 2004, letter provided clarifying information within
the scope of the original application and did not change the staff's
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of this amendment is contained in a
Safety Evaluation dated November 2, 2004.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: January 15, 2004, as
supplemented by letter dated March 15, 2004.
Brief description of amendments: The amendments revise the
Technical Specifications associated with the control rod drive trip
devices. The amendments are needed to support implementation of the
reactor trip breaker replacement.
Date of Issuance: November 2, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 341, 343, 342.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19566). The supplement dated March 15, 2004, provided clarifying
information that did not change the scope of the January 15, 2004,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 2, 2004.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: March 22, 2004, as supplemented
July 23 and October 11, 2004.
Brief description of amendments: The amendments modified Technical
Specification (TS) requirements to adopt the provisions of Industry/TS
Task
[[Page 68190]]
Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: November 4, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 263 and 144.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the TSs.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53108). The supplemental letters dated July 23 and October 11, 2004,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 4, 2004.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410,
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New
York
Date of application for amendments: January 8, 2004 (2 letters), as
supplemented by letter dated June 17, 2004.
Brief description of amendments: The amendments approve
implementation of the Boiling Water Reactor Vessel and Internals
Project Reactor Pressure Vessel Integrated Surveillance Program as the
basis for demonstrating the units' compliance with the requirements of
appendix H to Title 10 of the Code of Federal Regulations.
Specifically, the amendments approved the wording proposed by the
licensee to update the units' Updated Safety Analysis Reports. In
addition, the Unit 1 amendment also revised the Technical
Specifications to delete any reference to plant-specific surveillance
requirements.
Date of issuance: November 8, 2004.
Effective date: As of the date of issuance. Integrated Surveillance
Program shall be implemented within 90 days of issuance. The units'
Final Safety Analysis Report (Updated) shall be updated in accordance
with 10 CFR 50.71(e).
Amendment Nos.: 184 and 114.
Facility Operating License Nos. DPR-63 and NPF-69: Amendments
revise the Technical Specifications (for Unit 1), the operating license
(for Unit 2), and approve revision of licensing basis for both units.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7524). The June 17, 2004, letter provided clarifying information
within the scope of the original application and did not change the
staff's initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in two Safety Evaluations, both dated November 8, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: December 23, 2003, as
supplemented June 21, 2004.
Brief description of amendment: The amendment changes Technical
Specification (TS) Limiting Condition for Operation (LCO) Tables 3.2.1
and 3.2.4 to (1) eliminate the reactor head cooling containment
isolation function from the TSs, (2) correct and clarify the
description of the number of instrument channels per trip system as
defined in the TSs, and (3) revise an existing LCO for radiation
monitors used to isolate reactor building ventilation and initiate the
standby gas treatment system.
Date of issuance: November 2, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 140.
Facility Operating License No. DPR-22. Amendment revised the TSs.
Date of initial notice in Federal Register: March 30, 2004 (69 FR
16621).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 2, 2004.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: March 18, 2004, and its
supplements dated August 18 and 20, and September 17, 2004.
Brief description of amendments: The amendments authorize revisions
to the Final Safety Analysis Report (FSAR) Update to incorporate the
NRC approval of a permanently revised steam generator voltage-based
repair criteria probability of detection (POD) method. The revised POD
method is referred to as the probability of prior cycle detection
method. In addition, a reporting requirement is added to the DCPP
Technical Specifications as TS 5.6.10.i.
Date of issuance: October 28, 2004.
Effective date: October 28, 2004, and shall be implemented within
30 days of the date of issuance. The implementation of the amendment
includes the incorporation into the FSAR Update the changes discussed
above, as described in the licensee's application dated March 18, 2004,
and its supplements dated August 18 and 20, and September 17, 2004, and
evaluated in the staff's Safety Evaluation attached to the amendments.
Amendment Nos.: Unit 1-177; Unit 2-179.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the FSAR Update and the Technical Specifications.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34704).
The August 18 and 20, and September 17, 2004, supplemental letters
provided additional clarifying information, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 28, 2004.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: December 22, 2003, as
supplemented by letters dated June 18, July 15, and September 8, 2004.
Brief description of amendments: The amendment added TS 3.3.1.3,
``Oscillation Power Range Monitor (OPRM) Instrumentation,'' and changed
TS 3.4.1, ``Recirculation Loops Operating,'' and TS 5.6.5, ``Core
Operating Limits Report,'' to remove specifications and information
related to current stability specifications which will no longer be
needed with the operation of the OPRM system.
Date of issuance: November 9, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 217 and 192.
[[Page 68191]]
Facility Operating License Nos. NPF-14 and NPF-22: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: January 20, 2004 (69 FR
2745). The supplements dated June 18, July 15, and September 8, 2004,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 9, 2004.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: March 31, 2003, as supplemented
by letter dated July 30, 2004.
Brief description of amendment: The amendment revised the reactor
pressure vessel pressure-temperature limits and extends their validity
to 32 effective full power years.
Date of issuance: November 1, 2004.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 157.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 8, 2004 (69 FR
32076). The July 30, 2004 letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination or expand the application beyond the scope
of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 1, 2004.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland. Publicly available
records will be accessible from the Agencywide Documents Access and
Management System's (ADAMS) Public Electronic Reading Room on the
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html.
If you do not have access to ADAMS or if there are problems in
accessing the documents located in ADAMS, contact the NRC Public
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or
by e-mail to [email protected]. (Note: Public access to ADAMS has been
temporarily suspended so that security reviews of publicly available
documents may be performed and potentially sensitive information
removed. Please check the NRC Web site for updates on the resumption of
ADAMS access.)
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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
[[Page 68192]]
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, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1-800-397-4209, 301-415-4737, or by e-mail to [email protected]. (Note:
Public access to ADAMS has been temporarily suspended so that security
reviews of publicly available documents may be performed and
potentially sensitive information removed. Please check the NRC Web
site for updates on the resumption of ADAMS access.) If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or a presiding officer designated by the Commission or
by the Chief Administrative Judge of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the Chief Administrative Judge of the Atomic Safety and
Licensing Board will issue a notice of a hearing or an appropriate
order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify 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
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
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.\1\ 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 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.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to 301-415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or 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).
STP Nuclear Operating Company, Docket No. 50-499, South Texas Project,
Unit 2, Matagorda County, Texas
Date of amendment request: September 30, 2004.
Description of amendment request: The amendment changes Technical
Specification 4.4.4.2 to expand the range of conditions under which
quarterly testing of block valves for the pressurizer power operated
relief valves would be unnecessary.
Date of issuance: October 21, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 153.
[[Page 68193]]
Facility Operating License No. NPF-80: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. October 6, 2004 (69 FR 59969). The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by December 6, 2004, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated October 21, 2004.
Attorney for licensee: Mr. John E. Matthews, Morgan, Lewis &
Bokius, LLP, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Michael K. Webb, Acting.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: November 2, 2004.
Description of amendment request: The amendment revises Technical
Specification Limiting Condition for Operation 3.4.3, ``Primary Coolant
System (PCS) Pressure and Temperature (P/T) Limits'' to add
restrictions to the cooldown rate limits. This amendment supports plant
restart following repairs of two reactor vessel closure head control
rod drive nozzle penetrations at the Palisades Nuclear Power Plant.
Date of issuance: November 8, 2004.
Effective date: As of the date of issuance and shall be implemented
immediately.
Amendment No.: 218.
Facility Operating License No. DPR-20: Amendment revises the
Technical Specification.
Public comments requested as to proposed no significant hazards
consideration (NSHC):
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated November 8,
2004.
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.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of amendment request: November 4, 2004.
Description of amendment request: The proposed amendment extended
the implementation period for License Amendment 294 to May 15, 2005.
Date of issuance: November 9, 2004.
Effective date: As of date of issuance, to be implemented by May
15, 2005.
Amendment No.: 297.
Facility Operating License No. DPR-77: Amendment revises the
implementation date for License Amendment No. 294.
Public comments requested as to proposed no significant hazards
consideration (NSHC):
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated November 9,
2004.
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.
Dated at Rockville, Maryland, this 15th day of November, 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-25664 Filed 11-22-04; 8:45 am]
BILLING CODE 7590-01-P