[Federal Register Volume 69, Number 119 (Tuesday, June 22, 2004)]
[Notices]
[Pages 34696-34712]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-13753]
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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, May 28, 2004, through June 10, 2004. The
last biweekly notice was published on June 8, 2004 (69 FR 32070).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be
[[Page 34697]]
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/. If a request for a
hearing or petition for leave to intervene is filed within 60 days, the
Commission or a presiding officer designated by the Commission or by
the Chief Administrative Judge of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [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. 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].
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: May 5, 2004.
Description of amendment request: The proposed change will revise
Technical Specification Surveillance Requirement (SR) 4.0.5.a for
inservice inspection (ISI) and testing of American Society of
Mechanical Engineers (ASME) Code Class 1, 2, and 3 components, to
include a reference to the ASME Code for Operation and Maintenance of
Nuclear Power Plants (OM Code) in addition to Section XI of the ASME
Boiler and Pressure Vessel Code and applicable Addenda as required by
Title 10 of the Code of Federal Regulations (10 CFR), Section
50.55a(g).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 34698]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the Technical Specification SR 4.0.5.a
and the associated Bases are requested to add a reference to the
ASME OM Code and applicable Addenda for inservice inspection of ASME
Code Class 1, 2, and 3 components.
The existing Technical [Specification] requires inservice
inspection of ASME Code Class 1, 2, and 3, components and inservice
testing of ASME Code Class 1, 2 and 3 pumps and valves as required
by 10 CFR 50.55a. The purposes of the inservice inspection and
inservice testing programs are to assess the operational readiness
of pumps and valves, to detect degradation that might affect
component operability, and to maintain safety margins with
provisions for increased surveillance and corrective action. 10 CFR
50.55a defines the requirements for applying industry codes and
standards to each licensed nuclear power facility. The initial HNP
[Shearon Harris Nuclear Power Plant, Unit 1] ISI program was
developed in accordance with NRC regulations (10 CFR
50.55a(g)(4)(i)) to comply with the 1983 Edition of the ASME Boiler
and Pressure Vessel Code, including Addenda through the Summer of
1983 and is reflected in the existing Technical Specifications and
associated Bases sections.
The current, second ten-year interval HNP ISI program was
developed in accordance with the 1989 Edition (no Addenda) of ASME
Boiler and Pressure Vessel Code, Section XI. Subarticles IWF-1200
and IWF-5300 require the examination and testing of snubbers per the
first Addenda of ASME/ANSI [American National Standards Institute]
OM-1987, Part 4 (published in 1988), generally referred to as ``OM-
4.'' HNP Relief Request 2RG-008, Revision 1, grants HNP the ability
to retain the snubber testing and examination program in Technical
Specification 3/4.7.8.
The 1995 Edition with 1996 Addenda of the ASME OM Code,
Subsection ISTD, is the applicable Code per Code Case OMN-13. HNP
plans to utilize the 1995 Edition with 1996 Addenda of the ASME OM
Code for snubber visual examinations as an approved alternative to
the snubber visual examination requirements of the 1989 Edition of
ASME Section XI and as modified by HNP Relief Request 2RG-008,
Revision 1. Code Case OMN-13 has been evaluated and approved by the
NRC in Reg Guide 1.192.
The proposed change to Technical Specification SR 4.0.5.a is
also administrative in nature. The proposed changes comply with
approved codes and standards. As a result, there will be no affect
on plant safety.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The changes to Technical Specification SR 4.0.5.a and Bases
section 4.0.5 and are being proposed to reference the ASME OM Code
in addition to Section XI of the ASME Boiler and Pressure Vessel
Code. The proposed changes are administrative in nature and do not
adversely affect accident initiators or precursors nor alter the
design assumptions, conditions, or configuration of the facility.
The use of the ASME OM Code 1995 Edition with 1996 Addenda,
Subsection ISTD, with incorporation of the snubber visual
examination frequency of Code Case OMN-13 will result in an
improvement in personnel safety and dose reduction.
This change will have no operational impact, therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The changes to Technical Specification SR 4.0.5.a and Bases
section 4.0.5 do not involve a reduction in the margin of safety. As
previously identified, the subject changes are administrative in
nature and will add a reference to the ASME OM Code in Technical
Specification SR 4.0.5.a. Therefore, the proposed changes to the
Technical Specifications and Bases will not result in a reduction in
the margin of safety.
Based on the above, HNP concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: William Burton (Acting).
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: January 30, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.6.2, ``Secondary Containment
Isolation Instrumentation,'' Condition C, to add the words, ``not
met,'' to the end of the sentence, ``Required Action and associated
Completion Time.'' The omission of the words, ``not met,'' was an
oversight during the change to the Improved Standard Technical
Specifications (ISTS), NUREG 1433.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change corrects the sentence in Condition C of TS
3.3.6.2 by indicating that when this condition is not met, certain
actions are required. This terminology is prevalent throughout the
ISTS and is implied in this section as well. No changes in operating
practices or physical plant equipment are created as a result of
this terminology addition. 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 type of accident from any accident previously evaluated?
Response: No.
This proposed change is a correction of an action statement in
TS 3.3.6.2. No physical change in plant equipment will result from
this proposed change. Therefore, the proposed change does not create
the possibility of a new or different type of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is editorial in nature and only provides a
correction to an action statement in the Secondary Containment
Isolation Instrumentation involving inoperable channels and
automatic functions to agree with NUREG 1433. 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: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: March 19, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.6.1, ``Primary Containment
Isolation Instrumentation,'' to correct a formatting error introduced
during conversion to Improved Technical Specifications (ITS)
[[Page 34699]]
by replacing ``1 per room'' with ``2'' for the Required Channels Per
Trip System for the Reactor Water Cleanup (RWCU) Area Ventilation
Differential Temperature--High primary containment isolation
instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change restores the number of Required Channels Per
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1,
Primary Containment Isolation Instrumentation, to its pre-ITS value
and adds an explanatory note. No changes in operating practices or
physical plant equipment are created as a result of this change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different type of accident from any accident previously evaluated?
Response: No.
The proposed change restores the number of Required Channels Per
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1,
Primary Containment Isolation Instrumentation, to its pre-ITS value
and adds an explanatory note. No physical change in plant equipment
will result from this proposed change. Therefore, the proposed
change does not create the possibility of a new or different type of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is administrative in nature and only
provides a correction to Table 3.3.6.1-1 of TS 3.3.6.1, Primary
Containment Isolation Instrumentation, as well as an explanatory
note. 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: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: May 19, 2004.
Description of amendment request: The proposed change revises
Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' to
permit a longer completion time for the Division 1 and Division 2
diesel generators (DGs). This is a risk-informed TS change that would
extend the DG completion time from 72 hours (the current limit) to 14
days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change does not adversely affect the design of the
DGs, the operational characteristics or function of the DGs, the
interfaces between the DGs and other plant systems, or the
reliability of the DGs. Required Actions and the associated
Completion Times are not initiating conditions for any accident
previously evaluated, and the DGs are not initiators of any
previously evaluated accidents.
The DGs support the mitigation of the consequences of previously
evaluated accidents that involve a loss of offsite power. The
consequences of a previously analyzed accident will not be
significantly affected by the extended DG Completion Time since the
remaining DGs will continue to be capable of performing their
accident mitigation function as assumed in the accident analysis.
Thus, the consequences of accidents previously analyzed are
unchanged between the existing TS requirements and the proposed
changes. The consequences of an accident are independent of the time
the DGs are out of service as long as there are adequate DGs
available.
Based on the above, the proposed change to extend the DG allowed
Completion Time during plant operation will not involve a
significant increase in accident probabilities or consequences.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
No new accidents would be created since no changes are being
made to the plant that would introduce any new accident causal
mechanisms. This amendment request does not impact any plant systems
that are accident initiators; neither does it adversely impact any
accident mitigating systems. The addition of an independent AACSBC
[alternate AC source to the Division 1 and Division 2 battery
chargers] will provide added time for responding to a loss of all AC
power assumed in the accident analyses. The design of the AACSBC
will contain features and administrative controls to maintain the
separation and protection of emergency AC distribution systems and
does not create the possibility of a new or different kind of
accident from any previously evaluated.
Based on the above, implementation of the proposed changes will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
Throughout the period of the current TS Completion Time, when one DG
is out-of-service during power operation, the margin of safety is
managed by limiting the allowed outage time and other concurrent
power source outages within the TS. This time period is a temporary
relaxation of the single failure criteria, which, consistent with
overall system reliability considerations, provides a limited time
to repair the equipment and conduct testing. The extension of the
current TS Completion Time to 14 days has been determined not to be
a significant reduction in the margin of safety. The proposed
changes will not result in a significant decrease in DG availability
so that the assumptions regarding DG availability are not impacted.
Probabilistic Risk Assessment (PRA) methods, and a deterministic
analysis were utilized to fully evaluate the effect of the proposed
DG Completion Time extension. The results of the analysis show no
significant increase in Core Damage Frequency (CDF) and Large Early
Release Frequency (LERF). Energy Northwest has proposed a number of
risk management actions to reduce the possibility of a plant
transient; a loss of high-pressure injection and cooling systems, a
loss of other on-site power sources, or a loss of offsite power
during the period the DG is out-of-service.
Based on the above, the change to the TS Completion Time does
not result in a significant reduction in the margin of safety. This
is based on our management of plant risk, the reliability of the
other diesel generators, and the inclusion of risk management
actions.
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 C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: May 12, 2004.
[[Page 34700]]
Description of amendment request: The proposed amendment would
change the reactor core analytical methods used to determine the core
operating limits, reflect the changes allowed by Technical
Specification Task Force (TSTF) Traveler No. 363, ``Revised Topical
Report References in ITS [Improved Standard Technical Specifications]
5.6.5, COLR [Core Operating Limits Report],'' and delete the Index from
the Technical Specifications (TSs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
TS 6.9.5.1, Core Operating Limits Report (COLR)
The proposed amendment, in part, identifies a change in the
nuclear physics codes used to confirm the values of selected cycle-
specific reactor physics parameter limits and includes minor
editorial changes which do not alter the intent of stated
requirements. The proposed change also allows the use of methods
required for the implementation of ZIRLO clad fuel rods. Inasmuch as
the proposed change includes codes that have been previously
approved by the NRC [Nuclear Regulatory Commission] for CE
[Combustion Engineering] cores, the amendment is administrative in
nature and has no impact on any plant configuration or system
performance relied upon to mitigate the consequences of an accident.
Parameter limits specified in the COLR for this amendment are not
changed from the values presently required by TSs. Future changes to
the calculated values of such limits may only be made using NRC
approved methodologies, must be consistent with all applicable
safety analysis limits, and are controlled by the 10 CFR 50.59
process. Assumptions used for accident initiators and/or safety
analysis acceptance criteria are not altered by this change.
The proposed change also implements NRC approved TSTF Traveler
No. 363. This is an administrative change that will allow specific
details, such as the revision number, revision date, and supplement
number of topical reports that are referenced in the TSs, to be
deleted and relocated in the cycle specific COLR. This proposed
change does not result in any changes to the assumptions used to
evaluated accident initiators and/or safety analysis acceptance
criteria.
Index
The proposed deletion of the Index is purely administrative and
does not impact the accident analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
TS 6.9.5.1, Core Operating Limits Report (COLR)
The proposed change, in part, identifies a change in the nuclear
physics codes used to confirm the values of selected cycle-specific
reactor physics parameter limits. The proposed change also allows
the use of methods required for the implementation of ZIRLO clad
fuel rods. Neither of these changes results in a change to the
physical plant or to the modes of operation defined in the facility
license.
The proposed change also implements TSTF Traveler No. 363. The
proposed change does not result in changes to the physical plant or
to the modes of operation defined in the facility license nor does
it involve the addition of new equipment or the modification of
existing equipment.
Index
The proposed deletion of the Index is purely administrative has
no affect on existing equipment.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
TS 6.9.5.1, Core Operating Limits Report (COLR)
The proposed changes to change the nuclear physics code package
and to add a topical report to support the use of ZIRLO do not amend
the cycle specific parameter limits located in the COLR from the
values presently required by the TS. The individual specifications
continue to require operation of the plant within the bounds of the
limits specified in COLR. Benchmarking has shown that uncertainties
for the Westinghouse Physics code system yields are essentially the
same or less than those obtained for the current ROCS/DIT
methodology. Future changes to the values of these limits by the
licensee may only be developed using NRC approved methodologies,
must remain consistent with all applicable plant safety analysis
limits addressed in the Safety Analysis Report, and are further
controlled by the 10 CFR 50.59 process. The relocation of the
supplement numbers, revision numbers, and approval dates of the
analytical methods listed in the COLR does not affect the margin of
safety. The analysis will continue to be performed using NRC
approved methodology. Safety analysis acceptance criteria are not
being altered by this amendment.
Index
The proposed deletion of the Index, which is an administrative
document, does not impact any TS values or safety limits.
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: February 27, 2004.
Description of amendment request: This amendment request
incorporates a revision to the Technical Specifications and licensing
and design bases that supports a full-scope application of an
Alternative Source Term (AST) methodology.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The Nuclear Regulatory Commission (NRC) staff has
reviewed the licensee's analysis against the standards of 10 CFR
50.92(c). The NRC staff's review is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Adoption of the AST and those plant systems affected by
implementation of the AST do not initiate design-basis accidents
(DBAs). The proposed changes do not affect the design or manner in
which the facility is operated; rather, once the occurrence of an
accident has been postulated, the new AST is an input to analyses
that evaluate the radiological consequences. Therefore, the proposed
changes do not involve an increase in the probability of an accident
previously evaluated.
The structures, systems and components (SSCs) affected by the
proposed change act as mitigators to the consequences of accidents.
Based on the revised analyses, the proposed changes do revise
certain performance requirements; however, the proposed changes
involve different acceptance criteria. There cannot, therefore, be a
direct comparison to determine if the proposed change would result
in an increase in consequences over the current design. However, the
licensee's analysis proposes that, with implementation of AST, all
regulatory acceptance criteria continue to be met. Therefore, any
potential increase in consequences would not be considered
significant.
[[Page 34701]]
Therefore, the proposed changes do not result in 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?
Implementation of AST does not affect the design function or
mode of operations of SSCs in the facility prior to a postulated
accident. Since SSCs are operated essentially the same after the AST
implementation, no new failure modes are created by this proposed
change.
Therefore, the proposed changes will 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 a
margin of safety?
The changes proposed are associated with a revision to the
licensing basis. These changes would modify the input to DBA
analyses from the original source term to the AST. Based on the
revised analyses, the proposed changes involve different acceptance
criteria. There cannot, therefore, be a direct comparison to
determine if the proposed change would result in a reduction in a
margin of safety. However, the licensee's analysis proposes that,
with implementation of AST, all regulatory acceptance criteria
continue to be met. The dose consequences of the accident analyses
revised in support of the proposed changes are subject to the
acceptance criteria in 10 CFR 50.67, ``Accident source term,''
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for
Evaluating Design Basis Accidents at Nuclear Power Reactors,'' and
Standard Review Plan 15.0.1, ``Radiological Consequence Analyses
Using Alternative Source Terms.'' Thus, by meeting the applicable
regulatory limits for AST, any potential decrease in a margin of
safety would not be considered significant.
Therefore, because the proposed changes continue to result in
dose consequences within the applicable regulatory limits, the
changes are considered to not result in a significant reduction in a
margin of safety.
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. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: James W. Clifford.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: April 8, 2004.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS), Section 6, Administrative
Controls, to relocate (1) the Plant Operations Review Committee and
Nuclear Review Board requirements, (2) the program/procedure review and
approval requirements, and (3) the record retention requirements to the
Quality Assurance Topical Report, the document controlling the
licensee's quality assurance program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed changes involve the relocation of
several administrative requirements from the Technical
Specifications (TS) to a document subject to the control of 10 CFR
50.54(a), and is therefore, administrative in nature. The relocated
requirements involve the onsite and offsite organization's review
and audit, the review and approval of procedures, and the retention
of records. The change will not alter the physical design or
operational procedures associated with any plant structure, system,
or component. The change does not reduce the duties and
responsibilities of the organizations performing the review, audit,
and approval functions essential to ensuring the safe operation of
the plant.
Therefore, this proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No. The proposed changes are administrative in nature.
The changes do not alter the physical design, safety limits, or
safety analysis assumptions, associated with the operation of the
plant. Accordingly, the changes do not introduce any new accident
initiators, nor do they reduce or adversely affect the capabilities
of any plant structure, system, or component to perform their safety
function.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No. The proposed changes conform to NRC regulatory
guidance regarding the content of plant Technical Specifications.
The guidance is presented in Administrative Letter 95-06, and NUREG-
1433, Rev. 2. The relocation of these administrative requirements
will not reduce the quality assurance commitments as accepted by the
NRC, nor reduce administrative controls essential to the safe
operation of the plant. Future changes to these administrative
requirements will be performed in accordance with NRC regulation 10
CFR 50.54(a), consistent with the guidance identified above.
Accordingly, the relocation results in an equivalent level of
regulatory control.
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: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: James W. Clifford.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 17, 2004.
Description of amendment request: The proposed amendment would
revise the operating license and Technical Specifications (TSs) to
support an increase in the licensed power from 3411 megawatts thermal
(MWt) to 3587 MWt. This represents an increase of approximately 5.2
percent above the current rated licensed thermal power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The Nuclear Regulatory Commission (NRC) staff has
reviewed the licensee's analysis against the standards of 10 CFR
50.92(c). The NRC staff's review is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Plant structures, systems and components (SSCs) have been
verified to be capable of performing their intended design functions
at uprated power conditions. Where necessary, some components will
be modified prior to implementation of uprated power operations to
accommodate the revised operating conditions. The analysis indicated
that operation at uprated power conditions will not adversely affect
the capability of plant equipment. Current TS surveillance
requirements ensure frequent and adequate monitoring of system and
component operability. All systems will continue to be operated in
accordance with current design requirements under uprated
conditions; therefore, no new components or system interactions have
been identified that could lead to an increase in the probability of
any accident previously evaluated in the Updated Final Safety
Analysis Report (UFSAR).
The radiological consequences were reviewed for design basis
accidents (DBAs) previously analyzed in the UFSAR. The analysis
showed that the resultant radiological consequences for both loss-
of-coolant accidents (LOCAs) and non-LOCAs remain either unchanged
or have increased due to operation at uprated power conditions. Any
increase in the radiological
[[Page 34702]]
consequences of DBAs is not considered significant because plant
operation at uprated power conditions continue to meet established
regulatory limits.
Therefore, the proposed changes do not result in 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?
The configuration, operation, and accident response of the SSCs
are unchanged by operation at uprated power conditions or by the
associated proposed TS changes. Analyses of transient events have
confirmed that no transient event results in a new sequence of
events that could lead to a new accident scenario.
The effect of operation at uprated power conditions on plant
equipment has been evaluated. No new operating mode, safety-related
equipment lineup, accident scenario, or equipment failure mode was
identified as a result of operating at uprated conditions. In
addition, operation at uprated power conditions does not create any
new failure modes that could lead to a different kind of accident.
Minor plant modifications, to support implementation of uprated
power conditions, will be made as required to existing systems and
components. The basic design function of all SSCs remains unchanged
and no new safety-related equipment or systems will be installed
which could potentially introduce new failure modes or accident
sequences.
Based on this analysis, it is concluded that no new accident
scenarios, failure mechanisms or limiting single failures are
introduced as a result of the proposed changes. The proposed TS
changes do not have an adverse effect on any safety. Therefore, the
proposed changes will 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 a
margin of safety?
A comprehensive analysis was performed to support the power
uprate program at the Seabrook Station. This analysis identified and
defined the major input parameters to the Nuclear Steam Supply
System (NSSS), reviewed NSSS design transients, and reviewed the
capabilities of the NSSS fluid systems, NSSS/BOP (balance-of-plant)
interfaces, and NSSS and BOP components. The nuclear and thermal
hydraulic performance of nuclear fuel was also reviewed to confirm
acceptable results. Only minor plant modifications, to support
implementation of uprated power conditions, will be made as required
to existing systems and components. Changes in setpoints for
actuation of equipment do not adversely affect the outcome of any
postulated accident. The analysis indicated that all NSSS and BOP
systems and components will continue to operate within existing
design and safety limits at uprated power conditions.
The margin of safety of the reactor coolant pressure boundary is
maintained under uprated power conditions. The design pressure of
the reactor pressure vessel and reactor coolant system will not be
challenged as the pressure mitigating systems were confirmed to be
sufficiently sized to adequately control pressure under uprated
power conditions.
The radiological consequences were reviewed for DBAs previously
analyzed in the UFSAR. The analysis showed that the radiological
consequences of DBAs continue to meet established regulatory limits
at uprated power conditions.
The analyses supporting the power uprate program have
demonstrated that all systems and components are capable of safely
operating at uprated power conditions. All DBA acceptance criteria
will continue to be met. Therefore, it is concluded that the
proposed changes do not result in a significant reduction in the
margin of safety.
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: M.S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: James W. Clifford.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 27, 2004.
Description of amendment request: The proposed amendment would
revise the Cooper Nuclear Station (CNS) Technical Specifications (TS).
The proposed amendment would lower the reactor vessel water level at
which the reactor water cleanup (RWCU) system isolates, secondary
containment isolates, and the control room emergency filter system
(CREFS) starts. General Electric (GE) Service Information Letter (SIL)
No. 131 discussed problems that result from isolation of the RWCU and
start of the standby gas treatment (SGT) system, in conjunction with
isolation of secondary containment. The SIL recommended that the vessel
water level at which these actions occur be lowered, thereby
eliminating these problems and the resulting unnecessary complications
with scram recovery. The proposed changes to the CNS TS are in
accordance with SIL 131 Recommendations 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The values of various plant parameters at which piping
connected to the reactor vessel and containment isolates and air-
filtering systems start are not accident precursors. Thus, lowering
the reactor vessel water level at which RWCU and secondary
containment isolate and SGT and CREFS initiate has no impact on the
probability of a design basis accident evaluated in the CNS Station
Safety Analysis. Therefore the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
The proposed logic changes involve no changes to the logic of
the Reactor Protection System that initiates automatic reactor
shutdown in response to an accident. The proposed logic changes
involve no changes to the logic of the Emergency Core Cooling System
(ECCS) that initiates automatic actions to ensure adequate core
cooling and containment integrity in response to an accident. The
CNS response to the design basis accidents (DBAs) addressed in the
Station Safety Analysis with the proposed changes to the logic was
evaluated. This evaluation has demonstrated that there is no
increase in the offsite radiological doses to the public resulting
from these accidents.
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?
The proposed lowering of the level of water in the reactor
vessel at which certain automatic actions would occur changes the
operation of various systems at CNS. However, the change in system
operation is not significant. Currently automatic actions occur in
the RWCU System, SGT System, CREFS, and secondary containment in
response to reactor vessel water level. Changing the level at which
these automatic actions occur is not a significant change in the
systems operation. Hardware changes needed to implement the modified
logic are minor. Lowering the reactor vessel water level for these
actions does not introduce a new mode of plant operation and does
not create a potential for any new failure mechanisms, malfunctions,
or accident initiators. Making this change does not involve adding
new systems to the CNS design.
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?
The safety margin associated with dose consequences to the
public following DBAs is based on, in part, automatic operation of
systems that shut down the reactor, automatic initiation of ECCS,
and automatic isolation of primary and secondary containment. The
proposed changes to the CNS TS make no changes that affect the
automatic shutdown of the reactor or the
[[Page 34703]]
automatic initiation and operation of ECCS. The plant response to
DBAs with the proposed revisions to the RWCU isolation (primary
containment) and the SGT and the CREFS initiation (secondary
containment) have been evaluated and shown to not result in any
increase in dose to the public. The safety margin associated with
dose consequences to the control room operators is based on
automatic isolation of secondary containment, and initiation of
CREFS. The plant response to DBAs with the proposed revisions to the
RWCU isolation (primary containment) and SGT and CREFS initiation
(secondary containment) have been evaluated and shown to not result
in any increase in dose to the control room operators.
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 Section Chief: Robert A. Gramm.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 14, 2004.
Description of amendment request: The proposed amendment will
relocate the requirements of Technical Specification (TS) 3.3(1)a,
``Reactor Coolant System and Other Components Subject to ASME XI Boiler
& Pressure Vessel Code Inspection and Testing Surveillance,''
concerning inservice inspection of ASME Class 1, 2, and 3 components
and TS 3.4, ``Reactor Coolant System Integrity Testing,'' concerning
reactor coolant system integrity testing to the Fort Calhoun Station
(FCS) Updated Safety Analysis Report (USAR). These TSs do not meet the
criterion in 10 CFR 50.36(c)(2)(ii) for inclusion in the TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment relocates the requirements of TS 3.3(1)a
concerning inservice inspection of ASME Class 1, 2, and 3 components
and TS 3.4 concerning reactor coolant system integrity testing to
the FCS USAR. These TSs are directed toward prevention of component
degradation and continued long term maintenance of acceptable
structural conditions. It is not necessary to retain these TSs to
ensure immediate operability of safety systems. Therefore these TSs
do not meet the criteria set forth in 10 CFR 50.36(c)(2)(ii) for
inclusion in the TS. The requirements are being relocated from TS to
the FCS USAR, which will be maintained pursuant to 10 CFR 50.59,
thereby reducing the level of regulatory control. [This reduction in
the] level of regulatory control has no impact on the probability or
consequences of an accident previously evaluated. Therefore, the
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change relocates requirements of TS 3.3(1)a
concerning inservice inspection of ASME Class 1, 2, and 3 components
and TS 3.4 concerning reactor coolant system integrity testing that
do not meet the criteria for inclusion in TS set forth in 10 CFR
50.36(c)(2)(ii). The change does not involve a physical alteration
of the plant (no new or different type of equipment will be
installed) or make changes in the methods governing normal plant
operation. The change will not impose different requirements, and
adequate control of information will be maintained. This change will
not alter assumptions made in the safety analysis and licensing
basis. Therefore, the change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change relocates requirements of TS 3.3(1)a
concerning inservice inspection of ASME Class 1, 2, and 3 components
and TS 3.4 concerning reactor coolant system integrity testing that
do not meet the criteria for inclusion in TS set forth in 10 CFR
50.36(c)(2)(ii). The change will not reduce a margin of safety since
the location of a requirement has no impact on any safety analysis
assumptions. In addition, the relocated requirements of TS 3.3(1)a
and TS 3.4 concerning inservice inspection and testing of ASME Class
1, 2, and 3 components remain the same as the existing TS. Since any
future changes to these requirements will be evaluated per the
requirements of 10 CFR 50.59, there will be no reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 21, 2004.
Description of amendment request: The proposed amendment would add
information to the Technical Specification (TS) Basis for TS 2.4,
``Containment Cooling,'' to allow containment spray pumps to be secured
during a loss-of-coolant accident (LOCA) to minimize the potential for
containment sump clogging when certain conditions are met. NRC Bulletin
2003-01, ``Potential Impact of Debris Blockage on Emergency Sump
Recirculation at Pressurized Water Reactors,'' required that operators
of pressurized water reactor (PWR) plants state that the emergency core
cooling systems (ECCS) and the containment spray (CS) recirculation
functions meet applicable regulatory requirements with respect to
adverse post-accident debris blockage or describe interim compensatory
measures to reduce the risk associated with the potentially degraded or
non-conforming ECCS and CS recirculation functions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes will not [significantly] increase the
probability or consequences of any accident based on the following:
The proposed compensatory action is only taken following a LOCA
if all safeguards have functioned and if an excess of CS flow exists
above that required to control containment pressure, temperature,
and remove the accident source term. The proposed action is only
taken if the worst-case single failure has not occurred indicating
maximum containment cooling and SI [safety injection] flow
delivered, and minimum source term due to no severe core damage. The
proposed action occurs following the peak containment pressure
transient, therefore, the action has no impact on the peak
containment pressure analysis. A quantitative analysis of the change
in LOCA consequences due to suspension of CS flow for 10 minutes has
not been performed. However, the prerequisite conditions for taking
this action provide reasonable assurance that the loss of the
remaining CS train for ten minutes will not result in a significant
increase in the LOCA consequences. Therefore, the proposed changes
will not [significantly] increase the probability or consequence of
any accident.
[[Page 34704]]
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed revision does not involve physical changes to any
equipment required to mitigate the consequences of an accident, nor
alter how design basis accident events are postulated. The proposed
change alters the method of controlling an Engineered Safety Feature
following a design basis event so that manual actions are
substituted for automatic actions. Reasonable assurance exists that
these manual actions can be taken in a timely manner to allow
continued CS system operation to provide containment cooling and
source term reduction with no significant increases in the
radiological consequences or approaching of design containment
limits. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change alters the method of controlling an
Engineered Safety Feature following a design basis event so that
manual actions are substituted for automatic actions. The proposed
actions are only taken following a LOCA if all safeguards have
functioned and if an excess of CS flow exists above that required to
control containment pressure, temperature, and remove the accident
source term. The prerequisite conditions for taking this action
provide reasonable assurance that the loss of the remaining CS train
will not result in a reduction in the margin of safety for
radiological consequences or containment design parameters.
Therefore, the proposed changes do not involve a significant
reduction to 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: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: March 18, 2004.
Description of amendment requests: The proposed amendments would
authorize updates of the Diablo Canyon Power Plant (DCPP) Final Safety
Analysis Report (FSAR) Update to use on a permanent basis, a revised
steam generator (SG) voltage-based repair criteria probability of
detection (POD) method using plant specific SG tube inspection results,
referred to as the probability of prior cycle detection (POPCD) method.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The use of a revised steam generator (SG) voltage-based repair
criteria probability of detection (POD) method, the probability of
prior cycle detection (POPCD) method, to determine the beginning of
cycle (BOC) indication voltage distribution for the Diablo Canyon
Power Plant (DCPP) Units 1 and 2 operational assessments does not
increase the probability of an accident. Based on industry and plant
specific bobbin detection data for outside diameter stress corrosion
cracks (ODSCC) within the SG tube support plate (TSP) region, large
voltage bobbin indications which individually can challenge
structural or leakage integrity can be detected with near 100
percent certainty. Since large voltage ODSCC bobbin indications
within the SG TSP can be detected, they will not be left in service,
and therefore these indications should not be included in the
voltage distribution for the purpose of operational assessments. The
POPCD method improves the estimate of potentially undetected
indications for operational assessments, but does not directly
affect the inspection results. Since large voltage indications are
detected, they will not result in an increase in the probability of
a steam generator tube rupture (SGTR) accident or an increase in the
consequences of a SGTR or main steam line break (MSLB) accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The use of the POPCD method to determine the BOC voltage
distribution for the DCPP Units 1 and 2 operational assessments
concerns the SG tubes and can only affect numerical predictions of
probabilities for the SGTR accident. Since the SGTR accident is
already considered in the Final Safety Analysis Report Update, there
[is] no possibility to create a design basis accident that has not
been previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The use of the POPCD method to determine the BOC voltage
distribution for the DCPP Units 1 and 2 operational assessments does
not involve a significant reduction in a margin of safety. The
applicable margin of safety potentially impacted is the Technical
Specification 5.6.10, ``Steam Generator (SG) Tube Inspection
Report,'' projected end-of-cycle leakage for a MSLB [main steam line
break] accident and the projected end-of-cycle probability of burst.
Based on industry and plant specific bobbin detection data for ODSCC
within the SG TSP region, large voltage bobbin indications that can
individually challenge structural or leakage integrity can be
detected with near 100 percent certainty and will not be left in
service. Therefore these indications should not be included in the
voltage distribution for the purpose of operational assessments.
Since these large voltage indications are detected, they will not
result in a significant increase in the actual end-of-cycle leakage
for a MSLB accident or the actual end-of-cycle probability of burst.
The POPCD method approach to POD considers the potential for missing
indications that might challenge structural or leakage integrity by
applying the POPCD data from successive inspections. If a large
indication was missed in one inspection, it would continue to grow
until finally detected in a later inspection.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: April 27, 2004.
Description of amendment request: The proposed change will revise
the Safety Limit Minimum Critical Power Ratio (SLMCPR) values for two
recirculation loop and one recirculation loop operation. Each safety
limit value will be applicable for all fuel types in the Hope Creek
Generating Station core. In the amendment request, PSEG Nuclear LLC
requested changes to the Technical Specifications to support the use of
GE14 fuel and General Electric Company (GE) reload analysis methods
beginning with the upcoming Cycle 13.
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?
[[Page 34705]]
Response: No.
The SLMCPR ensures that no mechanistic fuel damage occurs in the
core if the limit is not violated. The revised SLMCPR values
maintain the appropriate conservative margin to boiling transition
and the probability of fuel damage is not increased. The derivation
of the revised SLMCPR values specified in the Technical
Specifications has been performed using NRC approved methods and
uncertainties. The analysis methodology incorporates appropriate
cycle-specific parameters and uncertainties in determining the
revised SLMCPR values. The analyses do not change the method of
operating the plant and have no effect on the probability of an
accident initiating event or transient. The revised SLMCPR values do
not affect the performance of systems or components used to mitigate
the consequences of accidents previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or radiological consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The revised SLMCPR values specified in the Technical
Specifications have been calculated in accordance with NRC approved
methods and uncertainties. The changes do not involve any new method
for operating the facility and do not involve any facility
modifications. No new initiating events or anticipated operational
occurrences result from these changes.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The revised SLMCPR values are calculated using NRC approved
methods and uncertainties. The revised SLMCPR values continue to
ensure that greater than 99.9% of all fuel rods in the core are
expected to avoid boiling transition if the safety limits are not
violated, thereby maintaining the fuel cladding integrity during
normal plant operation and anticipated operational occurrences.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
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 26, 2004.
Description of amendment request: The proposed change will revise
the Salem Unit Nos. 1 and 2 source term used for design basis
radiological analysis, in accordance with the provisions of 10 CFR
50.67, ``Accident Source Term''. The proposed change will also revise
certain requirements in the Technical Specifications (TSs) and the
Updated Final Safety Analysis Report (UFSAR) based on the radiological
dose analysis margins obtained in the Alternate Source Term
application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the three standards of 10 CFR 50.92(c). The NRC staff's
analysis is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The alternative source term analysis does not change the design
of the plant or affect the performance of the systems or components
used to mitigate the consequences of accidents previously evaluated.
The analyses do not change the method of operating the plant and has
no effect on the probability of an accident initiating event or a
transient. The alternative source term calculations demonstrate the
radiological consequences to the design basis accidents specified in
the plant's UFSAR will still remain well below the radiological
limits specified in 10 CFR 100.11. Therefore, since the radiological
consequences are well below the specified limits and the probability
of an accident is unchanged, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed amendment is not the result of a hardware design
change, nor does it lead to the need for a hardware design change.
There is no change in the methods or procedures by which the unit is
operated. As a result, all structures, systems, and components will
continue to perform as previously analyzed by the licensee, and
previously evaluated and accepted by the NRC staff. Therefore, the
proposed amendment will not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed changes result in operation in accordance with
regulatory guidelines and support the revisions to the radiological
analysis of the limiting design basis accidents. The radiological
consequences of these accidents are all within the regulatory
acceptance criteria associated with the use of the alternative
source term methodology. Therefore, the proposed changes do not
involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the 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.
Rochester Gas and Electric Corporation, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: March 1, 2004.
Description of amendment request: The proposed amendment would
extend the completion time (CT) from 1 hour to 24 hours for Condition B
of Technical Specification (TS) 3.5.1, ``Accumulators.'' The
accumulators are part of the emergency core cooling system and consist
of tanks partially filled with borated water and pressurized with
nitrogen gas. The contents of the tank are discharged to the reactor
coolant system (RCS) if, as during a loss-of-coolant accident, the
coolant pressure decreases to below the accumulator pressure. Condition
B of TS 3.5.1 specifies a CT to restore an accumulator to operable
status when it has been declared inoperable for a reason other than the
boron concentration of the water in the accumulator not being within
the required range. This change was proposed by the Westinghouse Owners
Group participants in the TS Task Force (TSTF) and is designated TSTF-
370. TSTF-370 is supported by NRC-approved Topical Report WCAP-15049-A,
``Risk-Informed Evaluation of an Extension to Accumulator Completion
Times,'' submitted on May 18, 1999. The NRC staff issued a notice of
opportunity for comment in the Federal Register on July 15, 2002 (67 FR
46542), on possible amendments concerning TSTF-370, including a model
safety evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for
[[Page 34706]]
referencing in license amendment applications in the Federal Register
on March 12, 2003 (68 FR 11880). The licensee affirmed the
applicability of the following NSHC determination in its application
dated March 1, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1 The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously Evaluated
The basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of the WCAP-15049, ``Risk-Informed
Evaluation of an Extension to Accumulator Completion Times,''
evaluation, the proposed change will allow plant operation with an
inoperable accumulator for up to 24 hours, instead of 1 hour, before
being required to begin shutdown. The impact of the increase in the
accumulator CT on core damage frequency for all the cases evaluated
in WCAP-15049 is within the acceptance limit of 1.0E-06/yr for a
total plant core damage frequency (CDF) less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049 for the accumulator CT increase meet the criterion of 5E-
07 in Regulatory Guides (RG) 1.174 [``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis''] and 1.177 [``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications''] for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049, design
basis accumulator success criteria are not considered necessary to
mitigate large break loss-of-coolant accident (LOCA) events, and
were only included in the WCAP-15049 evaluation as a worst case data
point. In addition, WCAP-15049 states that the NRC has indicated
that an incremental conditional core damage frequency (ICCDP)
greater than 5E-07 does not necessarily mean the change is
unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Criterion 2 The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049 evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049 evaluation demonstrates that the small
increase in risk due to increasing the accumulator allowed outage
time (AOT) is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this 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 the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of the WCAP-
15049 evaluation, the proposed change will allow plant operation
with an inoperable accumulator for up to 24 hours, instead of 1
hour, before being required to begin shutdown. The impact of this on
plant risk was evaluated and found to be very small. That is,
increasing the time the accumulators will be unavailable to respond
to a large LOCA event, assuming accumulators are needed to mitigate
the design basis event, has a very small impact on plant risk. Since
the frequency of a design basis large LOCA (a large LOCA with loss
of offsite power) would be significantly lower than the large LOCA
frequency of the WCAP-15049 evaluation, the impact of increasing the
accumulator CT from 1 hour to 24 hours on plant risk due to a design
basis large LOCA would be significantly less than the plant risk
increase presented in the WCAP-15049 evaluation.
Therefore, this change does not involve a significant reduction in
a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
Rochester Gas and Electric Corporation, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: March 1, 2004.
Description of amendment request: The proposed change allows entry
into a mode or other specified condition in the applicability of a
Technical Specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of Title 10 of the Code of Federal
Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition
for Operation (LCO) 3.0.4 exceptions in individual TSs would be
eliminated, several notes or specific exceptions are revised to reflect
the related changes to LCO 3.0.4, and Surveillance Requirement (SR)
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated March 1, 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 34707]]
Criterion 1 The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously Evaluated
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2 The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3 The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: April 28, 2004.
Description of amendment request: The proposed amendments would
relocate requirements related to the Cold Over Pressure Protection
System (COPS) arming temperature from the Technical Specifications
(TSs) to the Pressure and Temperature Limits Report (PTLR) to
facilitate future licensee-controlled changes to the COPS arming
temperature. The licensee also proposed to change the COPS arming
temperature from 350 [deg]F to 220 [deg]F.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes to the Technical Specifications do not
affect any plant equipment, test methods, or plant operation, and
are not initiators of any analyzed accident sequence. COPS will
continue to perform its function as designed to provide cold over
pressure protection, and the pressurizer safety valves will provide
over pressure protection during operation when COPS is not in
service. Operation in accordance with the proposed TS will ensure
that all analyzed accidents will continue to be mitigated by the
Structures, Systems, and Components (SSCs) as previously analyzed.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed changes do not introduce any new equipment,
create new failure modes for existing equipment, or create any new
limiting single failures. COPS will continue to ensure that
appropriate fracture toughness margins are maintained to protect
against reactor vessel failure during low temperature operation. The
proposed changes are consistent with [technical specification task
force] TSTF-233, Revision 0, which was approved by the NRC. Plant
operation will not be altered, and all safety functions will
continue to perform as previously assumed in accident analyses.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The proposed changes will not adversely affect the operation
of plant equipment or the function of any equipment assumed in the
accident analysis. The COPS arming temperature has been established
in accordance with an NRC-approved methodology. No changes are being
made to the cold Over pressure protection analysis and the function
of COPS as assumed in the analysis. Therefore, the proposed changes
do not involve a significant reduction in any margin to 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. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Stephanie M. Coffin, Acting Section Chief.
Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power
Station (YNPS) Franklin County, Massachusetts
Date of amendment request: November 24, 2003, and supplemented
December 10, 2003, December 16, 2003, January 19, 2004, January 20,
2004, February 2, 2004, February 10, 2004, and March 4, 2004.
Description of amendment request: The licensee has proposed to
amend its license to incorporate a new license condition addressing the
license termination plan (LTP). The new license condition would
document the date of NRC approval of the LTP and provide criteria to
determine the need for NRC approval of changes to the approved LTP.
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:
[[Page 34708]]
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Currently, the bounding airborne radioactivity event given in
the YNPS [Yankee Nuclear Power Station] FSAR [Final Safety Analysis
Report] is the materials handling event (FSAR Section 403.5). This
event considered the non-mechanistic release of the contents of the
dominant plant component that could have caused the highest offsite
dose as a result of the release of airborne radioactivity during
handling. The dominant component was the feed and bleed heat
exchanger which has since been removed from the site. The bounding
analysis resulted in an offsite dose at the Exclusion Area Boundary
of about 0.320 rem, significantly less than the EPA Protective
Action Guidelines. Other airborne particulate radwaste or
radioactive materials accidents considered in the FSAR but bounded
by the materials handling event are as follows:
Fire in a sea-land container containing combustible
radioactive material,
Dismantlement activities (i.e., cutting , segmentation)
during decommissioning,
A gas bottle explosion inside containment,
An explosion of a propane tank stored onsite.
All spent fuel is located at the ISFSI [Independent Spent Fuel
Storage Installation] and is stored within fifteen NAC Multi-Purpose
Canisters and associated vertical concrete casks. A sixteenth cask
contains Greater Than Class C material. The NAC-MPC FSAR addresses
the various off-normal and accident events which were postulated in
support of the licensing and certification of the system. In each
case, there were no radiological consequences as a result of a
postulated event.
The requested license amendment is consistent with plant
activities described in the PSDAR [Post Shutdown Decommissioning
Activities Report] and the YNPS FSAR. Accordingly, no systems,
structures, or components that could initiate the previously
evaluated accident or are required to mitigate these accidents are
adversely affected by this proposed change. Therefore, the proposed
change does not involve an increase in the probability or
consequences of any previously evaluated accident.
2. The proposed change does not create the possibility of a new
or different accident from any previously evaluated.
Accident analyses related to decommissioning activities are
addressed in the FSAR. The requested license amendment is consistent
with the plant activities described in the YNPS FSAR and the PSDAR.
The proposed change does not affect plant systems, structures, or
components in a way not previously evaluated. The changes do not
affect any of the parameters or condition that could contribute to
the initiation of an accident. No new accident scenarios are created
nor are any new failure mechanisms created by this activity.
Therefore, the proposed activity does not create the possibility of
a new or different kind of accident than those previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The LTP [License Termination Plan] is a plan for demonstrating
compliance with the radiological criteria for license termination as
provided in 10 CFR 20.1402. The margin of safety defined in the
statements of consideration for the final rule on the Radiological
Criteria for License Termination is described as the margin between
the 100 mrem/yr public dose limit established in 10 CFR 20.1301 for
licensed operation and the 25 mrem/yr dose limit to the average
member of the critical group at a site considered acceptable for
unrestricted use (one of the criteria of 10 CFR 20.1402). This
margin of safety accounts for the potential effect of multiple
sources of radiation exposure to the critical group. Since the
License Termination Plan was designed to comply with the
radiological criteria for license termination for unrestricted use,
the LTP supports this margin of safety.
In addition, the LTP provides the methodologies and criteria
that will be used to perform remediation activities of residual
radioactivity to demonstrate compliance with the ALARA [As Low As
Reasonably Achievable] criterion of 10 CFR 20.1402.
Also, as previously discussed, the bounding accident for
decommissioning is the materials handling event. Since the bounding
decommissioning accident results in more airborne radioactivity than
can be released from other decommissioning events, the margin of
safety associated with the consequences of decommissioning accidents
is not reduced by this activity. Therefore, the proposed change does
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Gerald Garfield, Esq., Day, Berry & Howard,
City Place 1, Hartford, CT 06103.
NRC Section Chief: Claudia Craig.
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 e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois, Docket No. 50-219, Oyster Creek
Generating Station, Ocean County, New Jersey, Three Mile Island Nuclear
Station, Unit 1, Dauphin County, Pennsylvania, Docket No. 50-289
Date of application for amendments: January 30, 2004.
Brief description of amendment: The amendments conformed the
Operating Licenses to reflect the current ownership structure of
AmerGen Energy Company, LLC. Exelon Generation Company currently owns
100% of AmerGen both directly and indirectly as a result of its
purchase on December 22, 2003, of the stock of British Energy U.S.
Holdings, Inc. The amendments deleted the License Conditions that are
no longer valid as a result of the change of the AmerGen ownership.
Date of Issuance: May 27, 2004.
[[Page 34709]]
Effective date: These license amendments are effective as of their
date of issuance.
Amendment Nos.: 160, 243, 249.
Facility Operating License Nos. DPR-16, DPR-50, and NPF-62:
Amendments revised the Operating Licenses.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9859).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated May 27, 2004.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of application for amendment: May 1, 2003, as supplemented
September 25, 2003, November 3, 2003, and February 25, 2004.
Brief description of amendment: The amendment adds Technical
Specification (TS) 3.7.16, ``Spent Fuel Pool Boron Concentration,''
modifies TS 4.3.1, ``Criticality'' and adds an additional license
condition that requires the licensee to develop a long-term coupon
surveillance program for the Carborundum samples.
Date of issuance: June 3, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 267.
Renewed Facility Operating License No. DPR-53: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28846).
The September 25, 2003, November 3, 2003, and February 25, 2004,
letters 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 amendment is contained
in a Safety Evaluation dated June 3, 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: October 16, 2001; as
supplemented by letters dated May 20, September 12, and November 21,
2002; September 22 and November 20, 2003; and February 18 and April 14,
2004.
Brief description of amendments: The amendments revised the
Technical Specifications to incorporate changes resulting from use of
an alternate source term.
Date of Issuance: June 1, 2004.
Effective date: These license amendments are effective as of the
date of issuance and shall be implemented in accordance with the
schedule provided in the licensee's letter dated February 18, 2004.
Amendment Nos.: 338, 339 & 339.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 22, 2002 (67 FR
2922).
The supplements dated May 20, September 12, and November 21, 2002;
and February 18 and April 14, 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 as
published in the Federal Register on January 22, 2002 (67 FR 2922). The
supplements dated September 22, 2003, and November 20, 2003, did change
the NRC staff's proposed no significant hazards consideration
determination. The NRC staff's proposed no significant hazards
consideration determination based on the submittals dated September 22,
2003, and November 20, 2003, were published in the Federal Register on
October 14, 2003 (68 FR 59215), and December 9, 2003 (68 FR 68660),
respectively.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 1, 2004.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: July 29, 2003, and as
supplemented by submittal dated January 14, 2004.
Brief description of amendments: Revise the technical
specifications by adding required actions for inoperable 250 VDC or 125
VDC battery charger, by relocating certain DC power surveillance
requirements and criteria to a licensee controlled program, and by
providing alternative criteria for battery charger testing and battery
monitoring with required actions. Additionally, a new program for
battery monitoring and maintenance is added to the technical
specifications.
Date of issuance: June 8, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 207/199.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 14, 2003 (68 FR
59215).
The supplemental submittal contained clarifying information that
was 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 a Safety Evaluation dated June 8, 2004.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania
Date of application for amendment: February 4, 2003, as
supplemented by letters dated October 24, 2003, and April 6, 2004.
Brief description of amendment: The amendment allowed the
engineered safeguards features actuation system slave relay test
frequency in footnote (1) to Technical Specification (TS) 4.3.2.1.1 to
be changed from once per 92 days to once per 12 months provided a
satisfactory contact loading analysis has been completed, and a
satisfactory slave relay service life has been established, for the
slave relay being tested.
Date of issuance: May 14, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No: 141.
Facility Operating License No. NPF-73. Amendment revised the TSs.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12953).
The supplements dated October 24, 2003, and April 6, 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 amendment is
contained in a Safety Evaluation dated May 14, 2004.
No significant hazards consideration comments received: No.
[[Page 34710]]
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: January 30, 2004.
Brief description of amendment: The amendment eliminates
requirements for hydrogen recombiners and relocates the requirements
for hydrogen and oxygen monitors to the licensee's Commitment Tracking
Program.
Date of issuance: May 21, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 138.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9862).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 21, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: January 30, 2004, supplemented
by letter dated May 6, 2004.
Brief description of amendments: The amendments eliminate
requirements for hydrogen recombiners and relocate the requirements for
hydrogen monitors to the Technical Requirements Manual.
Date of issuance: June 8, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 163 and 154.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9862).
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 amendments is contained
in a Safety Evaluation dated June 8, 2004.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: December 1, 2003, as
supplemented on March 10 and 30, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications to change the peak calculated post accident
primary containment internal pressure values for the primary
containment leakage rate testing program.
Date of issuance: May 28, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 241 and 184.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 20, 2004 (69 FR
2747).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 28, 2004.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: December 30, 2003.
Brief description of amendments: The amendments revised the staff
position titles in Section 5.0 ``Administrative Controls'' of the
Technical Specifications.
Date of issuance: June 3, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 242 and 185.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9865).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 3, 2004.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket No. 50-498, South Texas Project,
Unit 1, Matagorda County, Texas
Date of amendment request: October 16, 2003, as supplemented March
3, 2004.
Brief description of amendments: The amendment provides a one-time
change to Technical Specification 4.4.5.3a to extend the steam
generator inspection interval to 44 months for STP, Unit 1.
Date of issuance: June 8, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: Unit 1--162.
Facility Operating License No. NPF-76: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2003 (68
FR 64139). The supplement dated March 4, 2003, 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 June 8, 2004.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket No. 50-339, North Anna
Power Station, Unit 2, Louisa County, Virginia
Date of application for amendment: January 23, 2004.
Brief description of amendment: This amendment revises Technical
Specification Surveillance Requirements 3.5.1.4, 3.5.4.3, and 3.6.7.3
in order to delete a note that differentiates between the boron
concentrations at North Anna, Units 1 and 2, for the safety injection
accumulators, the refueling water storage tank, and the casing cooling
tank.
Date of issuance: June 4, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No: 218.
Renewed Facility Operating License No. NPF-7: Amendment changes the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 2004 (69 FR
16624).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 4, 2004.
No significant hazards consideration comments received: No.
[[Page 34711]]
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].
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 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://[email protected]/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]. 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
[[Page 34712]]
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).
Duke Energy Corporation, Docket No. 50-270, Oconee Nuclear Station,
Unit 2, Oconee County, South Carolina
Date of amendment request: June 4, 2004.
Description of amendment request: The amendment revised Technical
Specification 3.6.5, ``Reactor Building Spray and Cooling Systems,'' to
add a note that states that Limiting Condition of Operation 3.0.4 is
not applicable.
Date of issuance: June 4, 2004.
Effective date: June 4, 2004.
Amendment No.: 340.
Facility Operating License No. DPR-47: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. 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
June 4, 2004.
Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Stephanie M. Coffin, Acting.
Dated at Rockville, Maryland, this 14th day of June 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-13753 Filed 6-21-04; 8:45 am]
BILLING CODE 7590-01-P