[Federal Register Volume 68, Number 4 (Tuesday, January 7, 2003)]
[Notices]
[Pages 798-816]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-156]
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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 Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, December 13, through December 26, 2002. The
last biweekly notice was published on December 24, 2002 (67 FR 78515).
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.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing 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, 11555 Rockville
Pike (first floor), Rockville, Maryland. The filing of requests for a
hearing and petitions for leave to intervene is discussed below.
By February 6, 2003, 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 2.714,\1\ which is
available at the Commission's PDR, located at One White Flint North,
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 by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a
[[Page 799]]
notice of a hearing or an appropriate order.
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\1\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714 (d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April
29, 2002.
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As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland, by the above date. Because of continuing
disruptions in delivery of mail to United States Government offices, it
is requested that petitions for leave to intervene and requests for
hearing be transmitted to the Secretary of the Commission either by
means of facsimile transmission to 301-415-1101 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 Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and because of continuing disruptions in delivery of mail to United
States Government offices, 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 filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
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, 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, 304-415-4737 or by e-mail to [email protected].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendment request: November 13, 2002, as supplemented
November 20, 2002
Description of amendment request: The proposed amendments delete
requirements from the technical specifications (TS) and other elements
of the licensing bases to maintain a Post-Accident Sampling System
(PASS). Licensees were generally required to implement PASS upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the TS for nuclear power reactors currently licensed to
operate. Lessons learned and improvements implemented over the last 20
years have shown that the information obtained from PASS can be readily
obtained through other means or is of little use in the assessment and
mitigation of accident conditions.
The changes are based on NRC-approved Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post-Accident Sampling System
(PASS).'' The NRC staff issued a notice of opportunity for comment in
the Federal Register on December 27, 2001 (66 FR 66949), on possible
amendments
[[Page 800]]
concerning TSTF-413, 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 March 20,
2002 (67 FR 13027). The licensee affirmed the applicability of the
following NSHC determination in its application dated November 13,
2002.
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 PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
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 elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radioisotopes
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen G. Howe.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Located in Mecklenburg County, North Carolina
Date of amendment request: December 2, 2002.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) for Administrative Controls in
Section 5.0 concerning Responsibility, Unit Staff, Unit Staff
Qualifications, and Controls of the High Radiation Area.
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:
As required by 10 CFR 50.91(a)(1), this analysis is provided to
demonstrate that the proposed license amendment does not involve a
significant hazard.
Conformance of the proposed amendment to the standards for a
determination of no significant hazards, as defined in 10 CFR 50.92,
is shown in the following:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. Implementation of this amendment would not involve a
significant increase in the probability or consequences of an
accident previously evaluated. Approval of this amendment will have
no effect on accident probabilities or consequences since the
changes are purely administrative in nature.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. Implementation of this amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No physical
changes are being made to the plant. Therefore, the introduction of
any new accident scenarios does not exist. The amendment does not
impact any plant systems that are accident initiators nor does it
adversely impact any accident mitigating system. This amendment is
purely administrative in nature.
(3) Does the proposed change involve a significant reduction in
margin of safety?
No. Implementation of this amendment will 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 situation. These barriers include the fuel cladding, the
reactor coolant system, and the containment system. The performance
of these fission product barriers will not be impacted by
implementation of this amendment. System[s] and components are not
affected and therefore are capable of performing as designed. This
amendment is purely administrative nature, it will have no effect on
any safety margins.
Conclusion.
Based on the preceding analysis, it is concluded that the
proposed license amendment does not involve a Significant
[[Page 801]]
Hazards Consideration Finding as defined in 10 CFR 50.92.
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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Located in Mecklenburg County, North Carolina
Date of amendment request: December 12, 2002.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) for TS Table 3.3.2-1 Footnote
(c) to correct an editorial error, TS 3.4.3 is revised to update the
Reactor Coolant System Pressure-Temperature limits for use up to 34
Effective Full Power Years (EFPY) and TS 3.4.12 is revised to update
the Low Temperature Over-Pressure limits for use up to 34 EFPY.
Associated changes are also proposed for the TS Bases.
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:
Duke has evaluated whether or not a significant hazards
consideration is involved with the proposed amendments by focusing
on the three standards set forth in 10 CFR 50.92, ``issuance of
amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the reactor coolant system (RCS)
pressure and temperature (P-T) limits and low temperature
overpressure protection (LTOP) limits are developed utilizing the
methodology of American Society of Mechanical Engineers (ASME)
Section XI, Appendix G, in conjunction with the methodology of ASME
Code Case N-641. Usage of these methodologies provides compliance
with the underlying intent of 10 CFR [Part] 50 Appendix G and
provides operational limits established to prevent non-ductile
failure of the reactor vessel. The Loss of Coolant Accident analysis
and other accident analyses in the Updated Final Safety Analysis
Report (UFSAR) do not assume failure of the reactor vessel. The P-T
and LTOP limits are not initiators or contributors to accident
analyses addressed in the UFSAR. The proposed changes do not alter
any assumption previously made in the radiological consequence
evaluations nor affect the mitigation of the radiological
consequences of an accident previously evaluated. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The changes to RCS P-T limits and LTOP limits are proposed to
prevent non-ductile failure of the reactor vessel. The proposed
changes do not modify the RCS pressure boundary, nor make any
physical changes to the facility. The proposed changes do not
introduce any new mode of system operation or failure mechanism.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes are developed utilizing the methodology of
ASME Section XI, Appendix G, in conjunction with the methodology of
ASME Code Case N-461. Usage of these methodologies provides
compliance with the underlying intent of 10 CFR [Part] 50 Appendix G
and provides operational limits established to prevent non-ductile
failure of the reactor vessel. This Code case constitutes relaxation
from the current requirements of 10 CFR [Part] 50 Appendix G. The
alternate methodology allowed by the Code case is based on industry
experience gained since the inception of the 10 CFR [Part] 50
Appendix G requirements and replaces some requirements that have now
been determined to be excessively conservative. The more appropriate
assumptions and provisions allowed by the Code case maintain a
margin of safety that is consistent with the intent of 10 CFR [Part]
50 Appendix G. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
Based on the above, Duke concludes that the proposed amendments
present no significant hazards consideration under the standards set
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of amendment request: December 12, 2002.
Description of amendment request: The proposed amendment would
revise the Facility Operating License and Technical Specifications
(TSs) to increase the licensed core thermal power level to 3114.4
megawatts (MWt), which is a 1.4% increase above the currently
authorized power level of 3071.4 MWt. The proposed power uprate
involves the improvement in the core power uncertainty allowance
originally required for the emergency core cooling system (ECCS)
evaluations performed in accordance with Appendix K, ``ECCS Evaluation
Models,'' to Part 50 of Title 10 of the Code of Federal Regulations. In
addition, changes would be made in TS Sections 1.1, 2.1, 2.3, 3.1, 3.4,
6.9, and the applicable TS Bases would be revised to account for the
change in power level.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed 1.4% increase in maximum core thermal power is
based on the use of instrumentation that supports a reduction in the
measurement uncertainty value assumed in certain safety analyses.
The affected analyses now use an uncertainty value of 2% which was
required by 10 CFR [Part] 50 Appendix K at the time that the plant
was originally licensed. At that time, measurement of feedwater
flowrate in the plant secondary side used differential pressure-type
flow venturis. The plant secondary side thermal calorimetric is used
to determine reactor thermal power. A June 2000 revision to 10 CFR
[Part] 50 Appendix K permitted the use of lower uncertainty values
in the affected analyses, if the reduced value can be justified.
Entergy Nuclear Operations (ENO) has implemented the use of Caldon,
Inc. Leading Edge Flowmeter (LEFM) technology to measure feedwater
flowrate. The LEFM measures fluid velocity by measuring the transit
time of ultrasonic pulses introduced into the fluid stream. The LEFM
Check System implemented at Indian Point 2 has a demonstrated
measurement accuracy of 0.6%. Based on this measurement accuracy,
the licensed thermal power can be increased 1.4% by reducing the
assumed uncertainty used in safety analyses
[[Page 802]]
with respect to core thermal power from 2.0% to 0.6%. This results
in a net increase in licensed reactor core thermal power; from
3071.4 MWt to 3114.4 MWt. The LEFM and the flow venturi
instrumentation are used to collect data and there is no automatic
initiation function performed by this instrumentation. Use of the
LEFM instrumentation is therefore not an accident initiator and does
not increase the probability of occurrence of an existing analyzed
accident. Also, the LEFM instrumentation and the venturi
instrumentation do not mitigate accidents so that the consequences
of previously analyzed accidents are not increased.
Analyses and evaluations associated with the proposed change to
core thermal power have demonstrated that applicable acceptance
criteria for plant systems, components, and analyses (including the
Final Safety Analysis Report [FSAR] Chapter 14 safety analyses) will
continue to be met for the proposed 1.4% increase in licensed core
thermal power for Indian Point 2. The subject increase in core
thermal power will not result in conditions that could adversely
affect the integrity (material, design, and construction standards)
or the operational performance of any potentially affected system,
component or analysis. Therefore, the probability of an accident
previously evaluated is not affected by this change. The subject
increase in core thermal power will not adversely affect the ability
of any safety-related system to meet its intended safety function.
Further, the radiological dose evaluations in support of this power
uprate effort show that the current FSAR Chapter 14 radiological
analyses are unaffected, and that the current dose analyses of
record bound plant operation with the subject increase in licensed
core thermal power level.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed license amendment increases the maximum allowed
core thermal power through the use of feedwater flow instrumentation
that supports a reduction in the measurement uncertainty assumed in
certain safety analyses. The LEFM Check System instrumentation has
greater measurement accuracy than the differential pressure-type
flow venturi instrumentation that was originally used so that the
measurement uncertainty assumed in certain analyses can be
correspondingly reduced. Both the venturi and LEFM flow
instrumentation provide data that is used by plant operators to
monitor the thermal output of the plant. The instrumentation does
not perform an automatic actuation function and there are no output
signals to plant safety systems or control systems. Therefore,
instrumentation malfunction or failure does not introduce new
accident scenarios or equipment failure mechanisms. Operation,
maintenance, or failure of either instrumentation system does not
have an adverse effect on safety-related systems or any structures,
systems, and components required for transient or accident
mitigation.
Operating the plant at a new maximum core thermal power of
3114.4 MWt, which is 1.4% greater than the current maximum of 3071.4
MWt, is bounded by existing or updated analyses which demonstrate
that established limits and acceptance criteria continue to be met.
Operating at the new power level does not create new or different
accident initiators and existing credible malfunctions are bounded
by existing or updated analyses or evaluations.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The evaluations and analyses associated with the proposed
increase in maximum core thermal power demonstrate that applicable
acceptance criteria will continue to be met. The existing licensed
maximum core thermal power level incorporates a 2% measurement
uncertainty for the analysis of loss-of-coolant-accidents as
originally required by Appendix K of 10 CFR [Part] 50. The
regulations have subsequently been revised to allow the option of
justifying smaller measurement uncertainties by using more accurate
instrumentation to calculate reactor thermal power. Certain analyses
that already assume a bounding core power level because of the 2%
measurement uncertainty are not changed as a result of the proposed
increase in core thermal power. Use of the LEFM instrumentation with
improved measurement accuracy supports the use of a smaller
measurement uncertainty assumption in the safety analyses. Other
analyses were updated or evaluations were performed to demonstrate
that nuclear steam supply and balance-of-plant systems and
components will continue to perform, under normal and credible
transient conditions, within established 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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania
Date of amendment request: November 21, 2002.
Description of amendment request: Exelon Generation Company, LLC,
the licensee, is proposing a change to the Limerick Generating Station
(LGS), Unit 2, Technical Specifications (TSs) contained in Appendix A
to the Operating License. This proposed change will revise the TS
section on safety limits to incorporate revised safety limit minimum
critical power ratios (SLMCPRs) due to the cycle-specific analysis
performed by Global Nuclear Fuel for LGS, Unit 2, Cycle 8.
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 TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The derivation of the cycle specific Safety Limit Minimum
Critical Power Ratios (SLMCPRs) for incorporation into the Technical
Specifications (TS), and their use to determine cycle specific
thermal limits, has been performed using the methodology discussed
in ``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-14-US,
June, 2000, which incorporates Amendment 25. Amendment 25 was
approved by the NRC [Nuclear Regulatory Commission] in a March 11,
1999 safety evaluation report.
The basis of the SLMCPR calculation is to ensure that greater
than 99.9% of all fuel rods in the core avoid transition boiling if
the limit is not violated. The new SLMCPRs preserve the existing
margin to transition boiling. The GE-14 fuel is in compliance with
Amendment 22 to ``General Electric Standard Application for Reactor
Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-
24011-P-A-14-US, June, 2000, which provides the fuel licensing
acceptance criteria. The probability of fuel damage will not be
increased as a result of this change. Therefore, the proposed TS
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The SLMCPR is a TS numerical value, calculated to ensure that
transition boiling does not occur in 99.9% of all fuel rods in the
core if the limit is not violated. The new SLMCPRs are calculated
using NRC approved methodology discussed in ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which
incorporates Amendment 25. Additionally, the GE-14 fuel is in
compliance with Amendment 22 to ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and
U.S. Supplement, NEDE-24011-P-A-14-US, June 2000, which provides the
fuel licensing acceptance criteria. The SLMCPR is
[[Page 803]]
not an accident initiator, and its revision will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
There is no significant reduction in the margin of safety
previously approved by the NRC as a result of the proposed change to
the SLMCPRs, which includes the use of GE-14 fuel. The new SLMCPRs
are calculated using methodology discussed in ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which
incorporates Amendment 25. The SLMCPRs ensure that greater than
99.9% of all fuel rods in the core will avoid transition boiling if
the limit is not violated when all uncertainties are considered,
thereby preserving the fuel cladding integrity. Therefore, the
proposed TS change will not involve a significant reduction in the
margin of safety previously approved by the NRC.
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. Edward Cullen, Vice President & General
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett
Square, PA 19348.
NRC Section Chief: James W. Andersen.
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: November 26, 2002.
Description of amendment request: The proposed amendments revise
Technical Specification (TS) 3.1.3.1, Control Rod Operability,'' by
adding required actions for scram discharge volume (SDV) vent and drain
valves to align with those in NUREG-1433, ``Standard Technical
Specification, General Electric Plants, BWR/4,'' Revision 2.
Additionally, modifications are proposed to change TS 3.6.3, ``Primary
Containment Isolation Valves,'' to clarify the relationship between TS
3.1.3.1 and TS 3.6.3 regarding SDV vent and drain valve.
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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The scram discharge volume (SDV) and control rod drive (CRD)
system, including the associated SDV vent and drain isolation
valves, are not initiators to any accident sequence analyzed in the
Updated Final Safety Analysis Report (UFSAR). Operation in
accordance with the proposed Technical Specification (TS) ensures
that the SDV and control rods are capable of performing their
function as described in the UFSAR; therefore, the mitigative
functions supported by the SDV and control rods will continue to
provide the protection assumed by the analysis. The addition of
specific TS actions to be taken for inoperable SDV vent or drain
isolation valves will not challenge the ability of the SDV and
control rods to perform their design function. Appropriate
monitoring and maintenance, consistent with industry standards, will
continue to be performed. In addition, the CRD system including the
SDV isolation valves is within the scope of 10 CFR 50.65,
``Requirements for monitoring the effectiveness of maintenance at
nuclear power plants,'' which will ensure the control of maintenance
activities associated with the CRD system and SDV isolation valves.
Under the proposed TS changes, the SDV vent and drain lines may
be unisolated under administrative control. This allows any
accumulated water in the line to be drained, to preclude a reactor
scram on SDV high level. This is acceptable since the administrative
controls ensure the valve can be closed quickly, by a dedicated
operator, if a scram occurs with the valve open. The 8-hour
allowable outage time to isolate the line is based on the low
probability of a scram occurring while the line is not isolated and
unlikelihood of significant CRD seal leakage.
The proposed changes do not involve any physical change to
structures, systems, or components (SSCs) and do not alter the
method of operation or control of SSCs. The current assumptions in
the safety analysis regarding accident initiators and mitigation of
accidents are unaffected by these proposed changes. No additional
failure modes or mechanisms are being introduced and the likelihood
of previously analyzed failures remains unchanged.
The integrity of fission product barriers, plant configuration,
and operating procedures as described in the UFSAR will not be
affected by these proposed changes. Therefore, the consequences of
previously analyzed accidents will not increase because of these
proposed changes.
Based on the above discussion, the proposed TS changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a physical alteration of the
plant. No new equipment is being introduced, and installed equipment
is not being operated in a new or different manner. There are no
setpoints, at which protective or mitigative actions are initiated,
affected by these proposed changes. These proposed changes will not
alter the manner in which equipment operation is initiated, nor will
the function demands on credited equipment be changed. Any
alteration in procedures will continue to ensure that the plant
remains within analyzed limits, and no change is required to the
procedures relied upon to respond to an off-normal event as
described in the UFSAR. As such, no new failure modes are being
introduced. The changes do not alter assumptions made in the safety
analysis and licensing basis.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes are acceptable because the
operability of the SDV and SDV isolation valves is unaffected, there
is no detrimental impact on any equipment design parameter, and the
plant will still be required to operate within assumed conditions.
Operation in accordance with the proposed TS ensures that the SDV
and control rods are capable of performing their functions as
described in the UFSAR. Therefore, the support of the SDV and
control rods in the plant response to analyzed events will continue
to provide the margins of safety assumed by the analysis. The
additions to TS for inoperable SDV vent and drain isolation valves
will not challenge the ability of the SDV or control rods to perform
their design function. Appropriate monitoring and maintenance,
consistent with industry standards, will continue to be performed.
In addition, CRD system, including the SDV vent and drain isolation
valves, are within the scope of 10 CFR 50.65, ``Requirements for
monitoring the effectiveness of maintenance at nuclear power
plants,'' which will ensure the control of maintenance activities
associated with the CRD system. This provides sufficient management
control of the requirements that assure the control rods and CRD
system are maintained in a highly reliable condition. Although there
is an increase in allowable outage time, this increase was evaluated
and determined not to be a significant reduction in a margin of
safety.
The proposed TS Actions for inoperable SDV vent and drain
isolation valves are reasonable and consistent with approved
standards, guidance and regulations.
Based on the above discussion, the proposed TS 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.
[[Page 804]]
Attorney for licensee: Mr. Edward Cullen, Vice President & General
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett
Square, PA 19348.
NRC Section Chief: James W. Andersen.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: June 4, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) Surveillance Requirement (SR)
4.0.3 to extend the delay period, before entering a Limiting Condition
for Operation, following a missed surveillance. The delay period would
be extended from ``* * * up to 24 hours to permit completion of the
surveillance when the allowable (equipment inoperability) outage time
limits of the ACTION requirements are less than 24 hours'' to ``* * *
up to 24 hours or up to the limit of the specified frequency, whichever
is greater.'' In addition, the following requirement would be added to
SR 4.0.3: ``A risk evaluation shall be performed for any Surveillance
delayed greater than 24 hours, and the risk impact shall be managed.''
The proposed amendment is consistent with TS Task Force traveler TSTF-
358, which has been approved by the Nuclear Regulatory Commission
(NRC). The TS Bases will be revised under the licensee's existing TS
Bases control program to be consistent with the bases for TSTF-358.
Basis for proposed no significant hazards consideration
determination: The NRC staff issued a notice of opportunity for comment
in the Federal Register on June 14, 2001 (66 FR 32400), on possible
amendments concerning missed surveillances, 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 September 28, 2001 (66 FR 49714). The licensee reviewed the
model NSHC presented in the Federal Register and concluded that it is
applicable to Davis-Besse. The model NSHC determination was
incorporated by reference into its application dated June 4, 2002, to
satisfy the requirements of 10 CFR 50.91(a), and is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance 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)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
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 extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function. 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.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: December 9, 2002.
Description of amendment request: The proposed amendment utilizes
the Alternate Source Term radiological calculations to update the
design basis analysis in the Updated Safety Analysis Report for the
Fuel Handling Accident. Regulatory Guide 1.183, ``Alternative
Radiological Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Reactors,'' was utilized in the development of the
proposed amendment.
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. This proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment involves implementation of the
Alternative Source Term for the Fuel Handling Accident at the Perry
Nuclear Power Plant (PNPP). There are no physical design
modifications to the plant associated with the proposed amendment.
The revised calculations do not impact the initiators of a Fuel
Handling Accident in any way. They also do not impact the initiators
for any other design basis events. Therefore, because design basis
accident initiators are not being altered by adoption of the
Alternative Source Term analyses, the
[[Page 805]]
probability of an accident previously evaluated is not affected.
With respect to consequences, the only previously evaluated
accident that could be affected is the Fuel Handling Accident. The
Alternative Source Term is an input to calculations used to evaluate
the consequences of an accident, and does not by itself affect the
plant response, or the actual pathway of the radiation released from
the fuel. It does however, better represent the physical
characteristics of the release, so that appropriate mitigation
techniques may be applied. For the Fuel Handling Accident, the AST
analyses demonstrate acceptable doses, within regulatory limits,
after 24 hours of radiological decay, without credit for
Containment/Fuel Handling Building integrity, filtration system
operability, or Control Room automatic isolation. Therefore, the
consequences of an accident previously evaluated are not
significantly increased.
Based on the above conclusions, this proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. This proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment does not involve a physical alteration of
the plant (no new or different type of equipment will be installed
and there are no physical modifications to existing equipment
associated with the proposed changes). Also, no changes are proposed
to the methods governing plant/system operation during handling of
recently irradiated fuel, so no new initiators or precursors of a
new or different kind of accident are created. New equipment or
personnel failure modes that might initiate a new type of accident
are not created as a result of the proposed amendment.
Thus, this amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. This proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed amendment is associated with the implementation of
a new licensing basis for PNPP Fuel Handling Accidents. Approval of
the change from the original source term to a new source term taken
from Regulatory Guide 1.183 is being requested. The results of the
accident analyses, revised in support of the proposed license
amendment, are subject to revised acceptance criteria. The analyses
have been performed using conservative methodologies, as specified
in Regulatory Guide 1.183. Safety margins have been evaluated and
analytical conservatism has been utilized to ensure that the
analyses adequately bound the postulated limiting event scenario.
The dose consequences of the limiting Fuel Handling Accident remains
within the acceptance criteria presented in 10 CFR 50.67, ``Accident
Source Term,'' and Regulatory Guide 1.183.
The proposed changes continue to ensure that the doses at the
exclusion area and low population zone boundaries, as well as the
Control Room, are within corresponding regulatory limits. For the
Fuel Handling Accident, Regulatory Guide 1.183 conservatively sets
the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ)
limits below the 10 CFR 50.67 limit, and sets the Control Room limit
consistent with 10 CFR 50.67.
Since the proposed amendment continues to ensure the doses at
the EAB, LPZ and Control Room are within corresponding regulatory
limits, the proposed license amendment does not involve a
significant reduction in a margin of safety.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: October 23, 2002.
Description of amendment request: The proposed amendment would
revise Crystal River Unit 3 Improved Technical Specifications (ITS)
4.2.1, ``Fuel Assemblies,'' and ITS 4.2.2, ``Control Rods,'' to permit
the use of Framatome ANP M5 advanced alloy for fuel rod cladding and
fuel assembly structural components.
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:
Florida Power Corporation (FPC) has evaluated the proposed
License Amendment Request (LAR), which consists of the identified
Technical Specification changes and exemption requests, against the
criteria of 10 CFR 50.92(c). The Technical Specification changes are
categorized as follows:
1. Modification of Section 4.2.1, DESIGN FEATURES, Fuel
Assemblies, and to include the M5 advanced alloy for fuel rod
cladding and fuel assembly structural material[.]
2. Removal of design information such as maximum fuel
enrichment, nominal active fuel length, maximum individual rod
weight, and details of Control Rod content. Adopting the wording
from the Standard ITS.
3. Addition to ITS 4.2.1 of the following sentence: ``A limited
number of lead test assemblies that have not completed
representative testing may be placed in nonlimiting core regions.''
Crystal River Unit 3 does not intend to load lead test assemblies in
the upcoming fuel cycle (Cycle 14). This sentence is being added for
consistency with NUREG 1430, Revision 2.
FPC has concluded that this proposed LAR does not involve a
significant hazards consideration. The following is a discussion of
how each of the criteria is satisfied.
(1) [Does not] [i]nvolve a significant increase in the
probability or consequences of an accident previously evaluated.
M5 advanced alloy: Topical reports BAW-10227P-A, ``Evaluation of
Advanced Cladding and Structural Material (M5) in PWR [Pressurized
Water Reactor] Reactor Fuel,'' February 2000 and BAW-10179P-A,
Revision 4, ``Safety Criteria and Methodology for Acceptable Cycle
Reload Analyses,'' March 2001 provide the licensing basis for the
Framatome ANP (FRA-ANP) advanced cladding and structural material,
designated M5. The M5 material can be used for fuel rod cladding, as
well as for fuel assembly spacer grids, fuel rod end plugs, and fuel
assembly guide and instrument tubes. By letter dated August 2, 2001
(Reference 4), the NRC approved BAW-10179P-A, Revision 4, for
referencing in license applications. BAW-10179P-A, Revision 4
incorporates BAW-10227P-A. The M5 material was shown in these
documents to have equivalent or superior properties to the current
Zircaloy-4 material. The cladding itself is not an accident
initiator and does not affect accident probability. The M5 cladding
has been shown to meet all 10 CFR 50.46 design criteria and,
therefore, will not increase the consequences of an accident.
Removal of design parameters of maximum fuel enrichment, active
fuel length, rod weight and Control Rod content: This change moves
design features from Improved Technical Specifications (ITS) to the
Final Safety Analysis Report (FSAR) and other design documents and
analyses. The Framatome ANP enhanced fuel design will involve
increased rod weight and active fuel length. The approved Framatome
ANP topical report, BAW-10179P-A, ``Safety Criteria and Methodology
for Acceptable Cycle Reload Analyses,'' will continue to be used to
ensure that the required safety limits for the fuel are satisfied.
Therefore, the relocation of design information does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Addition of a limited number of lead test assemblies: This
change is administrative in nature and is proposed for consistency
with the ITS standard. Crystal River Unit 3 does not intend to load
lead test assemblies in the upcoming fuel cycle. When lead test
assemblies are to be loaded, the approved Framatome ANP topical
report BAW-10179P-A will be used to ensure that all applicable
limits of the safety analysis are met and that the lead test
assemblies are placed in nonlimiting core locations. Applicable
mixed core penalties and core operating limits will be developed and
applied. Therefore, use of lead test assemblies will not involve a
significant
[[Page 806]]
increase in the probability or consequences of an accident
previously evaluated.
(2) [Does not] [c]reate the possibility of a new or different
kind of accident from any accident previously evaluated.
M5 advanced alloy: Topical report BAW-10227P-A demonstrated that
the material properties of the M5 alloy are not significantly
different from those of Zircaloy-4. Therefore, M5 fuel rod cladding
and fuel assembly structural components will perform similarly to
those fabricated from Zircaloy-4, thus precluding the possibility of
the fuel becoming an accident initiator and causing a new or
different type of accident.
Removal of design parameters of maximum fuel enrichment, active
fuel length, rod weight and Control Rod content: This change moves
design features from ITS to the FSAR and other design documents and
analyses or adds consistency with the standard ITS. The location of
this information does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The approved FRA-ANP topical report, BAW-10179P-A will continue to
be used to ensure that the required safety limits are satisfied.
Therefore, these changes do not involve the possibility of a new or
different kind of accident from any accident previously evaluated.
Addition of a limited number of lead test assemblies: This
change is administrative in nature and it is proposed for
consistency with the ITS standard. Crystal River Unit 3 does not
intend to load lead test assemblies in the upcoming fuel cycle. When
lead test assemblies are to be loaded, they will be designed and
manufactured to ensure compatibility with the co-resident fuel
assemblies, core internal structures, and fuel handling and storage
equipment. The approved Framatome ANP topical report BAW-10179P-A
will be used to ensure that the lead test assemblies meet all
applicable limits of the safety analysis and that the lead test
assemblies are placed in non-limiting core locations. Applicable
mixed core penalties and core operating limits will be developed and
applied. Therefore, use of lead test assemblies will not involve the
possibility of a new or different kind of accident from any
previously evaluated.
(3) [Does not] [i]nvolve a significant reduction in a margin of
safety.
M5 advanced alloy: The proposed changes will not involve a
significant reduction in the margin of safety because it has been
demonstrated that the material properties of the M5 alloy are not
significantly different from those of Zircaloy-4. The M5 alloy is
expected to perform similarly or better [than] Zircaloy-4 for all
normal operating and accident scenarios, including both non-LOCA
[loss-of-coolant accident] and LOCA scenarios. For LOCA scenarios,
where the slight differences in M5 material properties relative to
Zircaloy-4 could have some impact on the overall accident scenario,
plant-specific LOCA analyses will be performed prior to the use of
fuel assemblies with fuel rods or fuel assembly components
containing M5. These LOCA analyses, required by ITS 5.6.2.18, ``Core
Operating Limits Report (COLR),'' will demonstrate that all
applicable margins of safety will be maintained by the use of the M5
alloy.
Removal of design parameters of maximum fuel enrichment, active
fuel length, rod weight and Control Rod content: Approved
methodologies will be used in the cycle-specific safety analysis to
evaluate the use of the M5 advanced alloy, and account for various
assembly differences (various rod weights and active fuel lengths).
The location of the design information does not affect the margin of
safety.
Addition of a limited number of lead test assemblies: This
change is administrative in nature and is proposed for consistency
with the ITS standard. Crystal River Unit 3 does not intend to load
lead test assemblies in the upcoming fuel cycle. When lead test
assemblies are to be loaded, the approved Framatome ANP topical
report BAW-10179P-A will be used to ensure that all applicable
limits of the safety analysis are met and that the lead test
assemblies are placed in nonlimiting core locations. Applicable
mixed core penalties and core operating limits will be developed and
applied. There will be no significant reduction in the margin of
safety when a limited number of lead test assemblies are utilized.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: R. Alexander Glenn, Associate General
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St.
Petersburg, Florida 33733-4042.
NRC Section Chief: Allen G. Howe.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of amendment request: November 25, 2002.
Description of amendment request: The proposed license amendment
would modify plant Technical Specifications (TSs) and the associated
spent fuel pool (SFP) criticality analyses to eliminate credit for the
BoraflexTM neutron absorber in SFP fuel storage racks and
credit specific rules to control fuel assembly positioning in the SFP
racks. TS 3.9.11 is revised to add a Limiting Condition for Operation
for the SFP soluble boron concentration and require periodic
surveillance of this parameter. This submittal provides justification
for removing the description of the poison material in the spent fuel
racks from Section 5 of the Unit 1 TSs, that was requested to be added
by the licensee's cask pit spent fuel storage rack submittal dated
October 23, 2002. In addition, a new SFP dilution analysis was
performed that supports the criticality analysis requirement for a
minimum soluble boron concentration.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Would operation of the facility in accordance with the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
No. The proposed amendment to eliminate reliance on
BoraflexTM and to credit SFP soluble boron for reactivity
control in the spent fuel pool storage racks was evaluated for
impact on the following previously evaluated events:
[sbull] A fuel handling accident (FHA)
[sbull] A fuel mispositioning event
[sbull] A cask drop accident
[sbull] A loss of spent fuel pool cooling
The proposed amendment does not modify the facility. A new
criticality analysis credits existing soluble boron in the SFP water
and specific fuel positioning rules for reactivity control, without
requiring any physical changes to the fuel storage racks. The
amendment does not change any rack module location or any module's
designation as Region 1 or Region 2 storage. There is no significant
increase in the probability of a fuel handling accident in the SFP
that is caused by crediting soluble boron and new fuel positioning
rules, rather than BoraflexTM, for reactivity control.
The probability of a fuel handling accident is a function of the
equipment design and procedures used when handling irradiated fuel.
Neither of these features is affected when soluble boron, instead of
BoraflexTM, is credited for reactivity control in the
SFP.
There is no increase in the probability of an accidental fuel
assembly mispositioning when crediting the presence of soluble boron
in fuel pool water for reactivity control. Fuel assembly selection
and manipulation will continue to be controlled by approved fuel
handling procedures; these procedures require the identification of
a verified target location prior to grappling the assembly. Fuel
placement will be in accordance with the revised TS.
There is no increase in the consequences of either an FHA or an
accidental mispositioning of a fuel assembly into the SFP racks.
Consequences of a FHA are not increased because the proposed
amendment does not change the fuel fission product inventory, local
meteorological conditions, or the fission product partition factor
provided by fuel pool water. The consequences of an accidental
misload are not increased because the criticality analysis
demonstrates that the fuel array will remain sub-critical, even if
the pool contains a boron concentration below the minimum level
required by Technical Specifications. The TS will ensure that an
adequate SFP soluble boron concentration is maintained for all
conditions.
The proposed fuel positioning rules do not cause the total
radionuclide inventory present in the spent fuel pool to increase,
or
[[Page 807]]
alter the type or mass of casks that may be placed in the fuel pool,
or alter any facet of operation of the spent fuel cask crane. No
characteristics of the existing spent fuel cask drop analysis for
Unit 1 are affected by the proposed fuel positioning rules or by
credit for soluble boron. Therefore, there is no increase in either
the probability or the consequences of a cask drop accident caused
by this change.
The proposed change does not increase either the probability or
the consequences of a loss of normal SFP cooling. The proposed fuel
positioning rules do not require any interaction with the fuel pool
cooling system. Credit for a portion of the existing soluble boron
concentration does not change its interaction with the fuel pool
cooling system. The ability to detect and mitigate a loss of SFP
cooling event is unchanged, and the revised criticality analysis
considered the effects of boiling in the SFP and found them
acceptable.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Would operation of the facility in accordance with the
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated?
No. The proposed change does not modify the physical plant,
nuclear fuel, or the design function and operation of the spent fuel
pool storage racks at St. Lucie Unit 1. A TS controlled minimum
concentration of soluble boron has always been required in the St.
Lucie Unit 1 spent fuel pool; as such, the possibility of an
inadvertent fuel pool dilution event has always existed. However,
the spent fuel pool dilution analysis that accompanies this
submittal demonstrates that no credible dilution event could
increase fuel pool reactivity such that the effective neutron
multiplication factor (keff) exceeds 0.95. Therefore,
implementation of credit for soluble boron to control reactivity in
the SFP will not create the possibility of a new or different type
of criticality accident.
The limiting fuel assembly mispositioning event does not
represent a new or different type of accident. The mispositioning of
a fuel assembly within the fuel storage racks has always been
possible. The locations of SFP rack modules and the specific modules
assigned to each storage region remain unchanged; analysis results
show that the storage racks remain subcritical, with substantial
margin, following a worst case fuel misloading event. Therefore, a
fuel assembly misload event that involves new fuel storage
arrangements required by the criticality analysis does not result in
a new or different type of criticality accident.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
(3) Would operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
No. The revised fuel positioning requirements proposed by this
license amendment provide sufficient safety margin to ensure that
the spent fuel pool storage racks will always remain subcritical. To
comply with the requirements of 10 CFR 50.68 when crediting soluble
boron, the current TS reactivity limit for the fuel storage racks
(i.e., keff less than or equal to 0.95 when flooded with
unborated water) will be replaced with two separate limits
(keff less than 1.0 when flooded with unborated water,
and keff less than or equal to 0.95 when flooded with
water containing 500 ppm boron).
The proposed amendment maintains the 0.95 reactivity limit by a
combination of restrictions on fuel characteristics and fuel
positioning, storage cell geometry and by crediting a portion of the
soluble boron in the SFP, rather than by crediting Boraflex.
The proposed license amendment does not reduce the margin of
safety provided by the soluble boron normally present in fuel pool
water; the TS minimum permissible boron concentration is not
decreased. The TS minimum required value of 1720 ppm is
substantially greater than the 500 ppm value required by the updated
criticality analysis to assure keff remains = 0.95 for
non-accident conditions; it is also substantially greater than the
soluble boron concentration necessary to compensate at a 95%
probability, with a 95 percent confidence for the limiting
postulated reactivity anomaly in the fuel pool storage racks.
No credible dilution of the fuel pool can result in an SFP
soluble boron concentration less than the minimum value required by
the criticality analysis. Therefore, an inadvertent dilution event
can not challenge safety margins.
Based on these evaluations and the supporting analyses,
operating the facility with the proposed amendment does not involve
in a significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Allen G. Howe.
GPU Nuclear Corporation and Saxton Nuclear Experimental Corporation
(SNEC), Docket No. 50-146, Saxton Nuclear Experimental Facility (SNEF),
Bedford County, Pennsylvania
Date of amendment request: April 22, 2002, as supplemented on
December 5, 2002.
Description of amendment request: The proposed amendment would
allow removal of the upper half of the SNEF containment vessel and make
a change to the organization to add the position of Vice-President GPU
Nuclear Oversight to reflect the merger of GPU Inc. and FirstEnergy
Corp.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
GPU Nuclear has determined that Technical Specification Change
Request No. 62 involves no significant hazard consideration as
defined in 10 CFR 50.92.
1. The proposed changes to the SNEC Technical Specifications do
not involve a significant increase in the probability of occurrence
or consequences of an accident or malfunction of equipment important
to safety previously analyzed in the safety analysis report.
As described in the change to delete Technical Specification
1.1.2, radiation levels inside the Containment Vessel will be below
that necessary to maintain the Containment Vessel as an Exclusion
Area. Further as required by modified Technical Specification 2.1.1
ventilation controls will be established to monitor and control any
potential releases of airborne radioactivity during activities
involving removal of the upper dome. Finally an analysis has been
performed to determine the dose to a maximally exposed individual
due to an accidental release while cutting the Containment Vessel.
In developing a source term for the event it was assumed that
following the concrete removal process the interior surfaces of the
upper Containment Vessel dome was homogeneously coated with concrete
dust. NUREG 1507 ``Minimum Detectable Concentrations with Typical
Radiation Survey Instruments for Various Contaminants and Field
Conditions'' describes an experiment to determine the attenuation
effects due to dusty conditions. The maximum dust loading presented
was 9.99 mg/cm2 for soil. This value was converted to
concrete dust by comparing the relative densities of the material
(1.5 g/cm3 for soil and 2.3 g/cm3 for
concrete) or 15.3 mg/cm2. This amount of dust coating the
internal surfaces of the Containment Vessel dome (9.05E6
cm2) results in 299 pounds of dust being left in the
Containment Vessel.
Table 1 provides the mix of isotopes remaining at the SNEC
Facility based on the most recent survey results and isotope decay.
During the removal operation a resuspension factor of 1.9E-2/m (as
described in NUREG/CR 0130 ``Technology, Safety and Costs of
Decommissioning a Reference Pressurized Water Reactor Power
Station'', Volume 2, page J-27) was selected to represent the amount
of concrete dust going airborne. This parameter is about one order
of magnitude larger than that used in any other accident analyses
described in the NUREG. This entire volume of dust was assumed to be
released, unfiltered, directly to the environment.
An accident dispersion factor (c/Q) of 3.41E-3 sec/
m3, was also selected as it is the highest, thus most
conservative, value used in the SNEC Facility Offsite Dose
Calculation Manual (ODCM). Additionally composite dose conversion
factors were selected from
[[Page 808]]
Table 5-1 of EPA 400-R-92-001 ``Manual of Protective Action Guides
and Protective Guides for Nuclear Incidents'' (US EPA, May 1992).
Based on the above a calculated dose of 3.23E-4 mrem to the
maximally exposed individual represents a conservative estimate for
an accidental release. For comparison Section 3.1 of the SNEC
Facility USAR estimated the dose from an unfiltered release due to a
material handling event of 1.5 mrem to the maximally exposed
individual.
Thus this proposed change does not involve a significant
increase in the probability of occurrence or consequences of an
accident or malfunction of equipment important to safety previously
analyzed in the SNEC Facility USAR.
For the portions of the amendment that would make a change to
the organization to add the position of Vice-President GPU Nuclear
Oversight to reflect the merger of GPU Inc. and FirstEnergy Corp,
these changes are administrative in nature. As such they have no
effect on the probability of occurrence or consequences of an
accident or malfunction of equipment important to safety.
2. The proposed changes to the SNEC Technical Specifications
will not create the possibility for an accident or malfunction of a
different type than any previously evaluated in the safety analysis
report.
As described in the response to item 1 above, the limiting
accidental release during segmentation of the Containment Vessel
dome involves the direct release of radioactive material to the
environment. This event is similar to both a material handling event
as described in Section 3.1 of the SNEC Facility USAR, and loss of
engineering controls during segmentation as described in Section 3.4
of the SNEC Facility USAR. Thus the possibility of a new accident is
not created.
For the portions of the amendment that would make a change to
the organization to add the position of Vice-President GPU Nuclear
Oversight to reflect the merger of GPU Inc. and FirstEnergy Corp,
these changes are administrative in nature. As such they have no
effect on the possibility of an accident or malfunction of a
different type.
3. The changes will not involve a significant reduction in the
margin of safety as defined in the basis for any technical
specification for SNEC. The SNEC Facility Technical Specifications
do not contain a defined margin of safety. However the implied
margin of safety is to protect members of the public from exposure
to radioactive material.
At the point in time that these Technical Specifications would
take affect general radiation levels in the SNEC Facility
Containment Vessel would be such that the Containment Vessel could
be opened for unrestricted use as defined in 10 CFR 20.1301.
Additionally the dose to a maximally exposed individual from an
accidental release during removal of the Containment Vessel dome is
several orders of magnitude below that from the limiting accidents
defined in the SNEC Facility USAR. Thus the margin of safety is not
reduced.
For the portions of the amendment that would make a change to
the organization to add the position of Vice-President GPU Nuclear
Oversight to reflect the merger of GPU Inc. and FirstEnergy Corp,
these changes are administrative in nature. As such they have no
effect on the margin of safety as defined in the basis for any
technical specification for SNEC.
The NRC staff has reviewed the analysis of the licensees and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for the Licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts, and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Program Director: William D. Beckner.
Table 1.--Maximum Exposed Individual Dose From Cutting the CV
--------------------------------------------------------------------------------------------------------------------------------------------------------
CV concrete
activity (Ci) Fraction CV wall area CV air Instantaneous
Isotope per table 4.13 remaining as concetration concetration release rate Concentration DCF \7\ Offsite dose
SNEC char. dust (uCi) (uCi/m) \2\ (uCi/m) \3\ (uCi/sec) \4\ (uCi/cm) \3\ (mrem)
report
--------------------------------------------------------------------------------------------------------------------------------------------------------
Am-241....................... 8.24e-05...... 4.68e-03...... 5.17e-06...... 9.83e-08..... 2.93e-04..... 9.99e-13..... 1.47e+05..... 1.47e-04
Co-60........................ 4.60e-02...... 2.61e+00...... 2.89e-03...... 5.49e-05..... 1.63e-01..... 5.57e-10..... 7.50e+01..... 4.18e-05
Cs-137....................... 2.38e-01...... 1.35e+01...... 1.49e-02...... 2.84e-04..... 8.46e-01..... 2.88e-09..... 1.14e+01..... 3.28e-05
C-14......................... 5.74e-03...... 3.26e-01...... 3.60e-04...... 6.84e-06..... 2.04e-02..... 6.96e-11..... 6.94e-01..... 4.83e-08
Eu-152....................... 1.42e-03...... 8.07e-02...... 8.91e-05...... 1.69e-06..... 5.05e-03..... 1.72e-11..... 7.50e+01..... 1.29e-06
H-3.......................... 1.29e-01...... 7.33e+00...... 8.10e-03...... 1.54e-04..... 4.58e-01..... 1.56e-09..... 2.14e-02..... 3.34e-08
Ni-63........................ 3.93e-02...... 2.23e+00...... 2.47e-03...... 4.69e-05..... 1.40e-01..... 4.76e-10..... 2.11e+00..... 1.01e-06
Pu-239....................... 5.24e-05...... 2.98e-03...... 3.29e-06...... 6.25e-08..... 1.86e-04..... 6.35e-13..... 1.44e+05..... 9.17e-05
Pu-241....................... 1.84e-04...... 1.05e-02...... 1.15e-05...... 2.19e-07..... 6.54e-04..... 2.23e-12..... 2.75e+03..... 6.13e-06
Sr-90........................ 1.59e-04...... 9.03e-03...... 9.98e-06...... 1.90e-07..... 5.65e-04..... 1.93e-12..... 4.44e+02..... 8.56e-07
-----------------
Total.................... 4.60e-01...... 2.61e+01...... .............. ............. 1.63e+00..... ............. ............. 2.70e+05
--------------------------------------------------------------------------------------------------------------------------------------------------------
\1\ Fraction remaining determined by: (299 lbs dust/5.26E6 lbs total concrete in CV) x 1E6 uCi/Ci x CV concrete activity.
\2\ Area concentration determined by dividing dust fraction remaining by 9.05E2 m\2\ (surface of CV shell being removed).
\3\ Air concentration determined by multiplying CV wall area activity by 1.9E-2/m (NUREG 0130 resuspension factor for dust sweeping).
\4\ Calculated by multiplying CV air specific activity by CV volume (2.98E3 m\3\) instantaneously released in one second.
\5\ Maximum atmospheric dispersion factor (X/Q) is 3.41E-3 sec/m\3\ at the site boundary (200 meters) and in Sector N per SNEC ODCM Revision 5.
\6\ Calculated by multiplying X/Q x activity released in uCi/sec x 1e-6 m\3\/cm\3\.
\7\ Per EPA 400-R-92-001, Table 5-1.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 10, 2002.
Description of amendment request: This proposed amendment would
replace the fire protection (FP) requirements contained in Facility
Operating License (FOL) Section 2.C.(4) with the standard fire
protection FOL condition recommended by Generic Letter 86-10, Section
F, adapted to Cooper Nuclear Station (CNS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change would revise the CNS Operating License
condition concerning
[[Page 809]]
the FP program and its change process. It does not alter the FP
requirements in the FHA [fire hazard analysis] or in the USAR
[updated safety analysis report] including the assumptions
underlying them. Neither does it alter SSCs [structures, systems or
components] relied on by analyses to mitigate accidents or special
events. Since it does not change any of the FP requirements or
analyses, this proposed amendment does not introduce a new initiator
for any of the accidents analyzed in the CNS USAR or considered
therein. Because it does not specifically change any FP requirements
or mitigating SSCs, this proposed amendment does not introduce a new
mechanism for degrading the mitigating features considered for the
accidents analyzed. By introducing no new accident initiators and no
new mechanisms for degradation of mitigating features, no
significant increase in the probability or consequences of an
accident previously evaluated is involved in the proposed change.
Therefore, the proposed change does not result in a significant
increase in radiological doses for any Design Basis Accident and
does not result in a significant increase in the types or amounts of
any effluents that may be released off-site.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed amendment does not physically change the fit, form,
or function of any SSC credited in the accident analyses or in the
FHA, Technical Requirements Manual (TRM), or the USAR. The proposed
change does not alter assumptions or requirements used in the FHA,
TRM, or USAR, nor does it affect the CNS Fire Protection program. It
does not, therefore, alter the FP program or affect the plant's
ability to achieve and maintain safe shutdown in the event of a
fire, and it does not result in a reduction in the level of fire
protection of the facility. Because it does not change FP
requirements, the FP program or fire-mitigating SSCs, this proposed
change does not create the possibility of a new or different kind of
accident from those previously evaluated for CNS.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The proposed amendment does not alter the design features of the
approved FP plan. The proposed amendment does not alter
administrative controls in the CNS Fire Protection program necessary
to ensure required performance of physical barriers during
anticipated operational occurrences and postulated accidents. The
proposed change does not alter the NRC approved Fire Protection
program as described in FP SER [safety evaluation report] dated May
23, 1979, SER Supplement 1 dated November 21, 1980, SER dated
September 21, 1983, SER dated April 16, 1984, SER dated August 21,
1985, SER dated April 10, 1986, SER dated November 7, 1988, SER
dated August 15, 1995. It does not affect the USAR, the TRM, the FHA
or the commitments contained therein. It does not physically change
the fit, form, or function of any SSC credited in the accident
analyses or in these documents. Because it does not change the
requirements, plan or mitigating SSCs, this proposed change does not
involve a significant reduction in a margin of safety.
In summary, the proposed amendment does not involve a
significant increase in the probability or consequences of an
accident or creates the possibility of a new or different kind of
accident or 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: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: November 22, 2002.
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
(TS), Section 4.6, ``Periodic Testing of Emergency Power System.'' This
proposed amendment would allow KNPP to inspect the diesel generators
(DGs) at least once per refueling frequency either while the plant is
operating or during a refueling outage. Current TS requires an
inspection during the refueling outage without exception. In addition,
the proposed amendment would allow KNPP to make administrative changes
to TS Section 4.6. The proposed change provides operational flexiblity
in the schedule of maintenance activities.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The DGs are accident mitigating equipment, not accident
initiating equipment. Consequently, there will be no impact on any
accident probabilities by the approval of the requested amendment.
The proposed change does not affect the performance of any
equipment used to mitigate the consequences of an analyzed accident.
Consequently, no analysis assumptions are violated and there are no
adverse effects on the factors that contribute to off-site or on-
site dose as the result of an accident.
The format, typographical, grammatical, and standardized naming
convention changes in addition to the WORD conversion are
administrative in nature and therefore have no impact on accident
initiators or plant equipment.
Based on the above, the proposed administrative changes and
permitting DG inspections to be performed during plant operation
does not involve a significant increase in the probabilities or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
No new accident mechanisms would be created as a result of NRC
approval of this amendment request since no changes are being made
to the plant that would introduce any new accident mechanisms.
Equipment would be operated in the same configurations with the
exception of the mode in which the inspection is credited. The
inspection will be performed within the current approved Technical
Specification limiting condition for operation (LCO). This amendment
request does not impact any plant systems that are accident
initiators or adversely impact any accident mitigating systems.
The proposed administrative changes do not involve any
modifications to the physical plant or operations. Administrative
changes do not contribute to accident initiators nor do they produce
a new accident scenario. Based on the above, implementation of the
proposed change would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in the 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 situation. These barriers include
fuel cladding, the reactor coolant system, and the containment
system. The proposed change to the inspection timing for the DGs do
not affect the operability requirements for the DGs, as verification
of such operability will continue to be performed as required.
Continued verification of operability supports the capability of the
DGs to perform their required function of providing emergency power
to plant equipment that supports the fission product barriers.
Consequently, the performance of these fission product barriers will
not be impacted by implementation of this license amendment request
and therefore does not involve a significant reduction in the margin
of safety.
The administrative changes do not affect plant equipment or
operation. Therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 810]]
Attorney for licensee: John H. O'Neill, Jr., Esq., Shaw Pittman,
Potts & Trowbridge, 2300 N. Street, NW., Washington, DC 20037-1128.
NRC Section Chief: L. Raghavan.
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: August 27, 2002.
Description of amendment requests: The proposed license amendments
would revise the term ``minimum measured flow per loop'' to ``measured
loop flow'' in the allowable value and nominal trip setpoint for the
Reactor Coolant Flow-Low reactor trip function contained in Table
3.3.1-1, ``Reactor Trip System Instrumentation,'' of Technical
Specification (TS) 3.3.1. In addition, the proposed amendments would
allow for an alternate method for the measurement of reactor coolant
system (RCS) total volumetric flow rate through measurement of the
elbow tap differential pressures on the RCS primary cold legs. The use
of elbow tap differential pressures normalized to Diablo Canyon Power
Plant Cycle 1 and 2 precision flow calorimetrics would improve the
accuracy of the RCS flow measurement through reduction of the effect of
hot leg temperature streaming that is present in the current flow
calorimetric 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 proposed change revises the Technical Specification (TS)
3.3.1 Table 3.3.1-1 term ``minimum measured flow per loop'' to
``measured loop flow'' in the allowable value and nominal trip
setpoint for the Reactor Coolant Flow-Low reactor trip function and
allows an alternate method for the measurement of reactor coolant
system (RCS) total flow to meet TS surveillance requirement (SR) SR
3.4.1.4 through measurement of the elbow tap differential pressures
on the RCS primary cold legs.
The change will not increase the probability of an accident
previously evaluated because adequate RCS flow will still be
assured. The Reactor Coolant Flow-Low reactor trip function
allowable value and nominal trip setpoint are accident mitigation
functions and are not an accident initiator. The elbow tap method to
measure RCS flow and the change to the flow definition associated
with the Reactor Coolant Flow-Low reactor trip function do not
involve a plant modification.
For the elbow tap method to measure RCS flow, sufficient margin
exists to account for all reasonable instrument uncertainties and
therefore the RCS flow will continue to be maintained at a value
which is bounded by the design basis accident initial conditions.
The change to the flow definition associated with the Reactor
Coolant Flow-Low reactor trip function allowable value and nominal
trip setpoint does not change a design basis accident initial
condition or the conditions at the time of reactor trip during a
design basis accident and therefore has no adverse effect on the
design basis accidents which credit the Reactor Coolant Flow-Low
reactor trip setpoint.
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 proposed change to the flow definition associated with the
Reactor Coolant Flow-Low reactor trip function allowable value and
nominal trip setpoint and the proposed elbow tap method to measure
RCS flow will not create the possibility of a new or different type
of accident from any previously evaluated. There are no physical
changes being made to the plant and there are no changes in
operation of the plant that could introduce a new failure mode,
creating an accident which has not been 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 proposed change to the flow definition associated with the
Reactor Coolant Flow-Low reactor trip function allowable value and
nominal trip setpoint and the proposed elbow tap method to measure
RCS flow will not reduce the margin of safety. For the proposed
elbow tap flow method, sufficient margin exists to account for all
reasonable instrument uncertainties and thus the RCS flow will
continue to be maintained at a value which is bounded by the design
basis accident initial conditions, and no adverse effect on the
plant response to design basis accidents is created. The change in
the flow definition associated with the Reactor Coolant Flow-Low
reactor trip function allowable value and nominal trip setpoint does
not change a design basis accident initial condition or the
conditions at the time of reactor trip during a design basis
accident, and therefore has no effect on the plant response to
design basis accidents which credit the Reactor Coolant Flow-Low
reactor trip setpoint. Since the change does not affect the response
to design basis accidents, it does not result in a decrease in
departure from nucleate boiling margin or reactor coolant system
peak pressure margin for the design basis accidents.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
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: November 1, 2002.
Description of amendment requests: The proposed license amendments
would revise Technical Specification (TS) 3.3.1, ``Reactor Trip System
(RTS) Instrumentation,'' and TS 3.3.2, ``Engineered Safety Feature
Actuation System (ESFAS) Instrumentation'' as follows: (1) Revise both
the RTS and ESFAS instrumentation TS and TS Bases to change or clarify
the allowances for bypassing and tripping tested channels with other
channels inoperable; (2) remove Surveillance Requirement 3.3.1.10 from
Function 16.b, ``Turbine Stop Valve Closure;'' (3) correct the nominal
trip setpoint value for Function 16.b, ``Turbine Stop Valve Closure;''
(4) correct the allowable value for the Function 18.f, ``Turbine
Impulse Chamber Pressure, P-13;'' and (5) remove and relocate the
nonsafety-related turbine trip function from Function 5 of Table 3.3.2-
1, ``Turbine Trip and Feedwater Isolation.'' This function will be
relocated to other owner-controlled documents.
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 in the required action statements in the
Limiting Conditions for Operation (LCOs) for the allowable
surveillance testing configurations for both the reactor trip system
(RTS) and engineered safety feature actuation system (ESFAS)
instruments will not change the probability or consequences of an
accident previously evaluated.
The proposed surveillance testing configuration changes only
clarify available surveillance testing configurations and
[[Page 811]]
limitations on those configurations. The changes do not modify how
the RTS and ESFAS functions respond to any accident condition. These
surveillance testing configurations provide greater flexibility to
prevent inadvertent actuation of these functions that could be a
precursor for an accident.
Previous Diablo Canyon Power Plant (DCPP) submittals have been
approved providing for the capability of surveillance testing in
trip and/or in bypass. Surveillance testing in bypass is considered
the preferred method for most Eagle 21 instruments. However, where
testing by tripping a single channel without causing a function
actuation is acceptable, that capability was also maintained.
Although some of the changes may appear to add new allowable
surveillance testing configurations, all of the proposed
configurations are based on the application of the intent behind the
existing Technical Specification (TS) wording. The limitations on
surveillance testing configurations provided by the proposed changes
are to ensure that there are no spurious actuations and that during
testing a valid signal will cause the associated functions to
actuate as designed. None of these configurations place the
associated function in a logic that has not been previously
evaluated and approved.
The proposed elimination of the channel calibration for the
turbine stop valve position switches will not change the probability
or consequences of an accident previously evaluated since these
switches are not subject to drift. These limit switches are
installed with fixed limit setpoints that actuate based on valve
position and they are not calibrated in the field. As a result, a
channel calibration being performed on these switches provides no
useful purpose other than to verify function similar to the
remaining trip actuation device operational test (TADOT). As a
result, performing only the TADOT provides all necessary assurances
of operability.
The correction of the turbine stop valve closure nominal trip
setpoint is administrative in nature and will not change the
probability or consequences of an accident previously evaluated.
This was an oversight in the Improved Technical Specification (ITS)
review and conversion process. The proposed change only returns the
setpoint to the previously evaluated value.
The proposed change to the allowable value for Function 18.f,
``Turbine Impulse Chamber Pressure, P-13,'' is administrative in
nature and will not change the probability or consequences of an
accident previously evaluated. The P-13 intended trip setpoint has
always been maintained at 10 percent and remains unchanged. This
modification is performed to provide consistency with current
methodology and NUREG-1431, and does not affect the operation of the
protective function.
The proposed removal and relocation of the turbine trip function
from ESFAS Function 5 will not change the probability or
consequences of an accident previously evaluated. The turbine trip
function is nonsafety-related and is not credited in any design
bases accident scenario. The proposed change only clarifies
importance of the two trip functions. The proposed changes in this
LAR [License Amendment Request] do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes in the required action statements in the
LCOs for the allowable surveillance testing configurations for both
the RTS and ESFAS instruments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed changes only clarify previously available
surveillance testing configurations and limitations on those
configurations. These clarifications ensure maximum surveillance
testing flexibility to prevent inadvertent actuation of these
functions that could be a precursor for an accident. The changes do
not modify any equipment, hardware or how the RTS and ESFAS
functions respond to any accident condition.
The proposed elimination of the channel calibration for the
turbine stop valve position switches will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. This change does not modify any equipment, hardware or
functions. The switches are installed with fixed limit setpoints
that actuate based on valve position. The switches are not subject
to drift and are not calibrated in the field. As a result, a channel
calibration being performed on these switches provides no useful
purpose other than to verify function similar to the required TADOT.
As a result, performing only the TADOT provides equivalent
assurances of operability.
The correction of the turbine stop valve closure nominal trip
setpoint in Function 16.b, ``Turbine Stop Valve Closure,'' is
administrative in nature and will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. This was an oversight in the ITS review and conversion
process. The proposed change does not modify any hardware or
equipment, and only returns the setpoint to the previously evaluated
value.
The proposed change to the allowable value for Function 18.f,
``Turbine Impulse Chamber Pressure, P-13,'' is administrative in
nature and will not create the possibility of a new or different
kind of accident from any accident previously evaluated. The P-13
intended (nominal) trip setpoint has always been maintained at 10
percent and remains unchanged. This change does not modify any
equipment or hardware. This modification is performed to provide
consistency with current methodology and NUREG-1431, and does not
affect the operation of the protective function.
The proposed removal and relocation of the turbine trip function
from ESFAS Function 5 will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The turbine trip function is nonsafety-related and is not credited
in any design bases accident scenario. The proposed change only
clarifies importance of the two trip functions.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes in the required action statements in the
LCOs for the allowable surveillance testing configurations for both
the RTS and ESFAS instruments will not involve a significant
reduction in a margin of safety. The proposed changes only clarify
previously available surveillance testing configurations and
limitations on those configurations. These clarifications ensure
maximum surveillance testing flexibility to prevent inadvertent
actuation of these functions that could be a precursor for an
accident. The changes do not modify any equipment, hardware or how
the RTS and ESFAS functions respond to any accident condition.
The proposed elimination of the channel calibration for the
turbine stop valve position switches will not involve a significant
reduction in a margin of safety. This change does not modify any
equipment, hardware or functions. The switches are installed with
fixed limit setpoints that actuate based on valve position. The
switches are not subject to drift and are not calibrated in the
field. As a result, a channel calibration being performed on these
switches provides no useful purpose other than to verify function
similar to the required TADOT. As a result, performing only the
TADOT provides equivalent assurances of operability.
The correction of the turbine stop valve closure nominal trip
setpoint in Function 16.b, is administrative in nature and will not
involve a significant reduction in a margin of safety. This was an
oversight in the ITS review and conversion process. The proposed
change does not modify any hardware or equipment, and only returns
the setpoint to the previously evaluated value.
The proposed change to the allowable value for Function 18.f,
``Turbine Impulse Chamber Pressure, P-13,'' is administrative in
nature and will not involve a significant reduction in a margin of
safety. The P-13 intended (nominal) trip setpoint has always been
maintained at 10 percent and remains unchallenged. This change does
not modify any equipment or hardware. This modification is performed
to provide consistency with current methodology and NUREG-1431, and
does not affect the operation of the protective function.
The proposed removal and relocation of the turbine trip function
from ESFAS Function 5 does not involve a significant reduction in a
margin of safety. The turbine trip function is nonsafety-related and
is not credited in any design bases accident scenario. The proposed
change only clarifies importance of the two trip functions.
None of the proposed changes affect the acceptance criteria for
any analyzed event. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection functions.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 812]]
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: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendments request: December 9, 2002.
Description of amendments request: The proposed amendments would
revise Technical Specification 3.7.5, ``Auxiliary Feedwater System,''
Surveillance Requirement (SR) 3.7.5.2 for San Onofre Nuclear Generating
Station, Units 2 and 3. Specifically, the proposed change would change
wording of the Frequency of SR 3.7.5.2 from ``31 days on a Staggered
Test Basis'' to ``In accordance with the Inservice Testing Program.''
Such inservice tests confirm component operability, trend performance,
and detect incipient failures by indicating abnormal performance. This
change is requested to implement recommendations from the Standard
Technical Specifications for Combustion Engineering Plants, NUREG-1432,
Revision 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
In June 2001, the Nuclear Regulatory Commission (NRC) issued
NUREG 1432, Revision 2, ``Standard Technical Specifications
Combustion Engineering Plants.'' For Technical Specification 3.7.5,
``Auxiliary Feedwater (AFW) System,'' Surveillance Requirement (SR)
3.7.5.2 requires verification that each AFW pump's developed head at
the flow test point is greater than or equal to the required
developed head which ensures that AFW pump performance has not
degraded during the cycle. This test confirms one point on the pump
design curve and is indicative of overall performance. This proposed
change will revise San Onofre Nuclear Generating Station (SONGS)
Surveillance Frequency to be consistent with NUREG 1432, Revision 2.
This change in and of itself will have no effect on the probability
or consequences of an accident previously evaluated.
Once this change to the Technical Specification is approved,
changes to the Surveillance Frequency of the AFW pumps would be
controlled in accordance with the Risk-Informed Inservice Testing
Program.
Therefore, the proposed change does not involve a significant
increase in the probability of consequences of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment will not change the design, configuration
or method of operation of the plant. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment will change the SR 3.7.5.2 Frequency from
``31 days on a Staggered Test Basis'' to ``In accordance with the
Inservice Testing Program.'' The proposed change does not change the
operation or surveillance requirements. It does not change the
design function of any of AFW system components. Therefore, the
proposed change does not involve a significant reduction in the
margin of safety.
Based on the above, Southern California Edison concludes that
the proposed amendment present no significant hazards consideration
under the standards set forth in 10 CFR 50.92(c), and, accordingly a
finding of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: December 2, 2002.
Description of amendment request: The proposed amendments change
Technical Specification Surveillance Requirement 3.6.4.1.2 to require
that only one access door in each access opening of the secondary
containment be verified closed every 31 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. [Does] the proposed change [* * *] involve a significant
increase in the probability or consequences of an accident
previously evaluated[?]
The proposed change to Surveillance Requirement SR 3.6.4.1.2
would require that only one of the two secondary containment access
doors be verified closed; presently, both doors are required to be
verified closed. This change is administrative in nature in that it
does not involve, require, or result from any physical change to me
secondary containment boundary or access door configuration. The
change to Surveillance Requirement SR 3.6.4.1.2 is consistent with
TSTF Standard Technical Specification Change Traveler TSTF-18,
Revision 1, and Surveillance Requirement SR 3.6.4.1.3 of Revision 2
of Volume 1 of NUREG-1433. As indicated in the ``Justification''
portion of Standard Technical Specification Change Traveler TSTF-18,
Revision 1, verifying one of the two access doors is closed is
sufficient to ensure that the infiltration of outside air does not
prevent the establishment and preservation of the required negative
pressure within the secondary containment. Indeed, neither the
requirements regarding minimum negative pressure and maximum
infiltration and drawdown time nor the actions required to be taken
should these requirements not be met will be altered by me proposed
Licensing amendment.
Because the physical characteristics and performance
requirements of the secondary containment will not be altered and
the change to Surveillance Requirement SR 3.6.4.1.2 is consistent
with the current revision of NUREG-1433, the proposed Licensing
amendment can not involve a significant increase in the probability
or consequences of any accident previously evaluated.
2. [Does] the proposed change [* * *] create the possibility of
a new or different kind of accident from any previously evaluated[?]
For the reasons previously discussed, neither the secondary
containment boundary nor the access door configuration will be
altered by or because of the proposed change to the surveillance
requirement. Likewise, the requirements defining and governing
secondary containment operability and functionality, that is,
Standby Gas Treatment system flow rate and secondary containment
negative pressure and drawdown limits, will not be changed. The
secondary containment, including its access openings, will remain
physically unaltered; will function as presently described in the
Updated Final Safety Analysis Report [(UFSAR)]; and will be subject
to the same structural and functional requirements. Under these
circumstances, this change can not, and 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
decrease in the margin of safety[?]
[[Page 813]]
The requirements defining and governing secondary containment
operability and functionality, that is, Standby Gas Treatment system
flow rate and secondary containment negative pressure and drawdown
limits, will not be changed. The secondary containment, including
its access openings will function as presently described in the [* *
*] UFSAR and will be subject to the same structural and functional
requirements. Therefore, this change can not, and does not, reduce
any margin of safety associated with the secondary containment
function.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: John A. Nakoski.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: December 4, 2002.
Brief description of amendments: The proposed amendments revise
several of the Required Actions in the Technical Specifications (TS)
that require suspension of operations involving positive reactivity
additions or suspension of operations involving reactor coolant system
(RCS) boron concentration reductions. In addition, the proposed
amendments revise several Limiting Conditions for Operation (LCO) Notes
that preclude reductions in RCS boron concentration.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no hardware changes. The RTS [Reactor Trip System] instrumentation
and reactivity control systems will be unaffected. Protection
systems will continue to function in a manner consistent with the
plant design basis. All design, material, and construction standards
that were applicable prior to the request are maintained.
The probability and consequences of accidents previously
evaluated in the FSAR [Final Safety Analysis Report] are not
adversely affected because the changes to the Required Actions and
LCO Notes assure the limits on SDM [Shutdown Margin] and refueling
boron concentration continue to be met, consistent with the analysis
assumptions and initial conditions included within the safety
analysis and licensing basis. The activities covered by this
amendment application are routine operating evolutions. The proposed
changes do not reduce the capability of reborating the RCS.
The proposed changes will not involve a significant increase in
the probability of any event initiators. The initiating event for an
inadvertent boron dilution event, as discussed in FSAR Section
15.4.6, is a failure in the reactor makeup control system (RMCS) or
operator error such that inventory makeup with the incorrect boron
concentration enters the RCS by way of the CVCS [Chemical and Volume
Control System]. Since the RMCS design is unchanged, there will be
no initiating event frequency increase associated with equipment
failures. However, there could be an increased exposure time per
operating cycle to potential operator errors during TS Conditions
that, heretofore, prohibited positive reactivity additions. As such,
the RTS Instrumentation and RCS Loops TS Bases changes from TSTF
[Technical Specification Task Force]-286, Revision 2, have been
augmented to preclude the introduction of reactor makeup water into
the RCS via the CVCS when one source range neutron flux channel is
inoperable or when no RCS loop is in operation. The equipment and
processes used to implement RCS boration or dilution evolutions are
unchanged and the equipment and processes are commonly used
throughout the applicable MODES under consideration. There will be
no degradation in the performance of, or an increase in the number
of challenges imposed on, safety-related equipment assumed to
function during an accident situation. There will be no change to
normal plant operating parameters or accident mitigation
performance.
The proposed changes will not alter any assumptions or change
any mitigation actions in the radiological consequence evaluations
in the FSAR.
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 accident previously evaluated?
Response: No.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. This amendment will not affect the normal method of plant
operation or change any operating limits. The proposed changes
merely permit the conduct of normal operating evolutions when
additional controls over core reactivity are imposed by the
Technical Specifications. The proposed changes do not introduce any
new equipment into the plant or alter the manner in which existing
equipment will be operated. The changes to operating procedures are
minor, with clarifications provided that required limits must
continue to be met. No performance requirements or response time
limits will be affected. These changes are consistent with
assumptions made in the safety analysis and licensing basis
regarding limits on SDM and refueling boron concentration.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this amendment. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this amendment.
This amendment does not alter the design or performance of the
7300 Process Protection System, Nuclear Instrumentation System, or
Solid State Protection System used in the plant protection systems.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter the limits on SDM or refueling
boron concentration. The nominal trip setpoints specified in the
Technical Specifications Bases and the safety analysis limits
assumed in the transient and accident analyses are unchanged. None
of the acceptance criteria for any accident analysis is changed.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, departure from nucleate boiling ratio (DNBR) limits, heat
flux hot channel factor (FQ), nuclear enthalpy rise hot
channel factor (FDH), loss of coolant accident peak cladding
temperature (LOCA PCT), peak local power density, or any other
margin of safety. The radiological dose consequence acceptance
criteria listed in the Standard Review Plan will continue to be met.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and
[[Page 814]]
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, 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].
Carolina Power & Light Company, Docket No. 50-324, Brunswick Steam
Electric Plant, Unit 2, Brunswick County, North Carolina
Date of amendment request: September 16, 2002.
Brief description of amendment: The amendment revises a license
condition by deleting the requirement to include check valve MVD-V5008
in the facility check valve program.
Date of issuance: December 13, 2002.
Effective date: December 13, 2002.
Amendment No.: 251.
Facility Operating License No. DPR-62: Amendment revises Appendix
B, ``Additional Conditions.''
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68731).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 13, 2002.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: January 31, 2002, as
supplemented on September 18, 2002.
Brief description of amendments: The amendments change the method
of verifying boron concentration of each safety injection tank. Rather
than taking a sample of each tank every 31 days, the revised technical
specification surveillance requirement requires leakage into the tanks
to be monitored every 12 hours and a sample to be taken every 6 months.
Date of issuance: December 19, 2002.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 255 and 232.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 16, 2002. The
September 18, 2002, letter provided clarifying information that did not
change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of these amendments
is contained in a Safety Evaluation dated December 19, 2002.
No significant hazards consideration comments received: No.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of amendment request: September 10, 2001, as supplemented by
letters dated June 19 and November 8, 2002. The supplemental
information provided clarification that did not change the scope or the
initial no significant hazards consideration determination.
Brief description of amendment: The amendment revises TS 3/4.9.7
and the corresponding Bases to address the use of a single-failure-
proof-handling system for the Spent Fuel Building and to remove the
restriction on travel of crane loads in excess of 1800 pounds.
Date of issuance: December 17, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 198.
Facility Operating License No. DPR-61: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 5, 2002 (67 FR
10009).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 17, 2002.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: August 29, 2002.
Brief description of amendments: The amendments revised the
Technical Specifications 3.8.4.7, to modify the note to eliminate the
``once per 60 months'' restriction on replacing the battery service
test by the battery modified performance discharge test. Associated
changes to the TS Bases are also included.
Date of issuance: December 17, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 209 & 190.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68733).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 17, 2002.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of application for amendment: May 30, 2002, as supplemented on
October 31, 2002.
Brief description of amendment: The amendment revised the
requirements in several administrative programs in Technical
Specification Section 6.0, ``Administrative Controls.'' Specifically,
the amendment: (1) Replaced the specific management titles for several
organizational positions with generic titles, (2) replaced the title of
the Quality Assurance Program Description
[[Page 815]]
with a reference to the quality assurance program described or
referenced in the Updated Final Safety Analysis Report, and (3) deleted
the functions of the Station Nuclear Safety and the Nuclear Facilities
Safety Committees and the Vice President-Nuclear Power since their
duties and responsibilities are described in the Quality Assurance
Program Description.
Date of issuance: December 17, 2002.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 235.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42824).
The October 31 supplemental letter provided clarifying information
that did not expand the scope of the amendment or change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 17, 2002.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendment: June 28, 2002, as supplemented
on October 15 (two separate letters), October 17, November 15, and
December 6, 2002.
Brief description of amendment: The amendment increases the
licensed reactor core power level by 1.66 percent from 3250 megawatts
thermal (MWt) to 3304 MWt. The power level increase is considered a
measurement uncertainty recapture power uprate.
Date of issuance: December 20, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 273.
Facility Operating License No. DPR-58: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 23, 2002 (67 FR
48219).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 20, 2002.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 8, 2002.
Brief description of amendment: The amendment revised Technical
Specification 2.7, ``Electrical Systems,'' to increase the amount of
diesel fuel oil required for seven days of emergency diesel generator
operation.
Date of issuance: December 16, 2002.
Effective date: December 16, 2002, and to be implemented within 30
days of issuance.
Amendment No.: 213.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68741).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 2002.
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: June 24, 2002, as supplemented
by letter dated September 24, 2002.
Brief description of amendments: The amendments delete Technical
Specification 5.5.3, ``Post Accident Sampling System (PASS),'' and
thereby eliminate the requirements to have and maintain the PASS at
Plant Hatch.
Date of issuance: December 18, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 235 & 177.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50958).
The supplement dated September 24, 2002, provided clarifying
information that did not change the scope of the June 24, 2002,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 18, 2002.
No significant hazards consideration comments received: No
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendments request: October 24, 2001, as supplemented by
correspondent e-mails dated August 27, 2002, and September 24, 2002.
Brief description of amendments: The amendments consist of
relocating various Technical Specifications (TSs) to the Technical
Specification Requirements Manual (TRM). The amendments will relocate
TSs 3/4.1.3.3, 3/4.3.3.2, 3/4.3.3.11, 3/4.4.7, 3/4.4.9.2, 3/4.3.4.11,
3/4.7.2, 3/4.7.10, 3/4.9.3, 3/4.9.5, 3/4.9.7, 3/4.10.5, and 3/4.11.2.5
to the TRM. Their associated bases will also be relocated to the TRM to
be consistent with relocation of the various TSs. In addition, the
proposed amendment corrects various typographical and page numbering
errors, deletes an outdated one-time exception, and makes minor formal
changes to improve consistency.
Date of issuance: The license amendment is effective as of its date
of issuance and shall be implemented within 6 months from the date of
issuance.
Effective date: December 17, 2002.
Amendment Nos.: Unit 1--145; Unit 2--33.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 5, 2002 (67 FR
5334).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 17, 2002.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: September 3, 2002.
Description of amendment request: The proposed amendment revised
Surveillance Requirement (SR) 3.0.3 to extend the delay period, before
entering a Limiting Condition for Operation, following a missed
surveillance. The delay period is extended from the current limit of
``* * * up to 24 hours or up to the limit of the specified Frequency,
whichever is less'' to ``* * * up to 24 hours or up to the limit of the
specified Frequency, whichever is greater.'' In addition, the following
requirement is added to SR 3.0.3: ``A risk evaluation shall be
performed for
[[Page 816]]
any Surveillance delayed greater than 24 hours and the risk impact
shall be managed.''
Date of issuance: December 23, 2002.
Effective date: Date of issuance, to be implemented within 45 days.
Amendment Nos.: 243, 278, 237.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63698).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 2002.
No significant hazards consideration comments received: No.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 2002.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant ,
Unit 1, Rhea County, Tennessee
Date of application for amendments: September 3, 2002.
Description of amendment request: The proposed amendment revises
Surveillance Requirement (SR) 3.0.3 to extend the delay period, before
entering a Limiting Condition for Operation, following a missed
surveillance. The delay period is extended from the current limit of
``* * * up to 24 hours or up to the limit of the specified Frequency,
whichever is less'' to ``* * * up to 24 hours or up to the limit of the
specified Frequency, whichever is greater.'' In addition, the following
requirement is added to SR 3.0.3: ``A risk evaluation shall be
performed for any Surveillance delayed greater than 24 hours and the
risk impact shall be managed.''
Date of issuance: December 11, 2002.
Effective date: Date of issuance, to be implemented within 45 days.
Amendment No.: 42.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63699).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 11, 2002.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket No. 50-280, Surry
Power Station, Unit 1, Surry County, Virginia
Date of application for amendment: October 15, 2001, as
supplemented November 8, 2001, June 28, 2002, and July 25, 2002 .
Brief Description of amendment: This amendment revises the
Technical Specifications to allow a one-time change in the Appendix J
Type A containment integrated leakage rate test interval from the
required 10 years to a test interval of 15 years at Surry Power
Station, Unit 1.
Date of issuance: December 16, 2002.
Effective date: December 16, 2002.
Amendment No.: 233.
Facility Operating License No. DPR-32: Amendment changes the
Technical Specifications.
Date of initial notice in Federal Register: December 12, 2001 (66
FR 64309). The November 8, 2001, June 28, 2002, and July 25, 2002,
supplements contained clarifying information only and did not change
the initial no significant hazards consideration determination or
expand the scope of the initial application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 2002.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 30th day of December 2002.
For the Nuclear Regulatory Commission.
Stuart A. Richards,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 03-156 Filed 1-6-03; 8:45 am]
BILLING CODE 7590-01-P