[Federal Register Volume 60, Number 27 (Thursday, February 9, 1995)]
[Notices]
[Pages 7801-7803]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-3234]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-413]
Duke Power Company, et al.; Notice of Consideration of Issuance
of Amendment to Facility Operating License, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of an amendment to Facility Operating License No.
NPF-35 issued to Duke Power Company (the licensee) for operation of the
Catawba Nuclear Station, Unit 1, located in York County, South
Carolina.
The proposed amendment request would propose the renewal for
Catawba Unit 1 Cycle 9 operation of the steam generator tube inspection
bobbin probe voltage-based interim plugging criteria that had been
previously approved for Cycle 8. Approval of this amendment will
preclude unnecessary plugging or repairing tubes by sleeving due to the
occurrence of outer diameter initiated stress corrosion cracking
(ODSCC) at the tube support plate elevations in the Catawba Unit 1
steam generators. The interim plugging criteria approved for Cycle 8
and contained in the draft Generic Letter 94-XX, ``Voltage-Based Repair
Criteria for the Repair of Westinghouse Steam Generator Tubes Affected
by Outside Diameter Stress Corrosion Cracking,'' can be summarized as
follows:
Flaw indications with a bobbin coil voltage less than or equal
to 1.0 volt can remain in service without further action. For flaw
indications in excess of 1.0 volt but less than 2.7 volts, the tube
can remain in service provided an RPC inspection of the indication
does not detect ODSCC or any other degradation mode. Crack
indications above 2.7 volts will be plugged or repaired by sleeving,
and do not require RPC confirmation.
This amendment request reflects the ``Requested Actions: for a
licensee that chooses to implement a steam generator tube interim
plugging criteria, as stated in the draft NRC Generic Letter, 94-XX
``Voltage-Based Repair Criteria for the Repair of Westinghouse Steam
Generator Tubes Affected by Outside Diameter Stress Corrosion
Cracking.''
The changes being proposed to the Technical Specification (TS) do
not alter the interim plugging criteria currently stated in the TS
which was approved and utilized during Cycle 8. The primary change to
the TS is to incorporate the guidance of draft Generic Letter 94-XX,
``Voltage-Based Repair Criteria for the Repair of Westinghouse Steam
Generator Tubes Affected by Outside Diameter Stress Corrosion
Cracking,'' which will allow removal of the cycle-specific limitation
currently in the TS.
Before issuance of the proposed license amendment, the Commission
will have made findings required by the Atomic Energy Act of 1954, as
amended (the Act) and the Commission's regulations.
The Commission has made a proposed determination that the amendment
request involves 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. As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
(1) Operation of Catawba Unit 1 in accordance with the proposed
license amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
A single tube rupture is not anticipated during operation of
Catawba Unit 1. Based on the existing data base, the limiting RG
[Regulatory Guide] 1.121 criterion for tube burst capability of 3
times normal operating differential is satisfied with \3/4\''
diameter tubing with bobbin coil indications with signal amplitudes
less than 4.54 volts, regardless of the indicated depth measurement.
This structural limit is based on a lower 95% prediction bound of
the data and using LTL material properties. A 1.0 volt plugging
criteria compares favorably with the structural limit considering
the previously calculated growth rates for ODSCC within the Catawba
Unit 1 steam generators. Assuming a voltage increase of 0.4 volts,
and adding a 14% NDE uncertainty of 0.14 volts (90% cumulative
probability) to the interim plugging criteria [IPC] of 1.0 volt
results in an EOC [end-of-cycle] voltage of approximately 1.6 volts.
This end of cycle voltage compares favorably with the Structural
Limit of 4.54 volts. The applicability of assumed growth rates for
each cycle of operation will be confirmed prior to return to power
of Catawba Unit 1. A similar structural margin is anticipated for
subsequent cycles.
In addition, for an EOC voltage structural limit of 4.54 volts,
applying the 40% growth allowance and the 14% NDE uncertainty
results in a margin between the structural limit and the alternate
repair limit (2.7 volts), which is well within the structural limit.
This repair limit will be applied for IPC implementation to repair
bobbin indications greater than 2.7 volts independent of RPC
confirmation of the indication.
Concerning SLB [steamline break] leakage in support of
implementation to the interim plugging criteria, it will be
determined whether the distribution of cracking indications at the
tube support plate intersections at the end of a cycle are projected
to be such that primary to secondary leakage would result in site
boundary doses within the pertinent 10 CFR 100 limits. The SLB
leakage rate calculation methodology * * * will be used to calculate
End of Cycle SLB leakage. Based on EOC 8 projections, it is
calculated that leakage during a postulated SLB event at the EOC 8
will be limited to approximately 1.61 gpm which is shown to result
in acceptable dose consequences. [An] SLB leakage of 17.5 gpm in the
faulted loop results in dose consequences which are less than the
pertinent 10 CFR 100 limits. Similar results are expected for
subsequent cycles and confirmation of leak rates will be performed
prior to placing the [s]team generators in service. [[Page 7802]]
Therefore, renewal of the proposed 1.0 volt interim plugging
criteria does not adversely affect steam generator tube integrity
and results in acceptable dose consequences. The proposed amendment
does not result in any increase in the probability or consequences
of an accident previously evaluated within the Catawba Unit FSAR
[Final Safety Analysis Report].
(2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Renewal of the proposed steam generator tube interim plugging
criteria does not introduce any significant changes to the plant
design basis. Use of the criteria does not provide a mechanism which
could result in an accident outside of the region of the tube
support plat elevations--no ODSCC is occurring outside the thickness
of the tube support plates. Neither a single or multiple tube
rupture event would be expected in a steam generator in which the
plugging criteria has been applied (during all plant conditions).
Upon application of the interim plugging criteria, no primary to
secondary leakage during normal operation is anticipated during all
plant conditions due to degradation at the tube support plate
elevations in the Catawba Unit 1 steam generators. However,
additional conservatism is built into the existing operating leakage
limit with regard to protection against the maximum permissible
single crack length which may be achieved during operation due, in
large part, to the potential occurrence of through-wall cracks at
locations other than the tube support plate intersections.
Application of the 1.0 volt interim steam generator tube
plugging criteria at Catawba Unit 1 is not expected to result in
tube burst during all plant conditions during operation. Tube burst
margins are expected to meet RG 1.121 acceptance criteria. The
limiting consequence of the application of the interim plugging
criteria is a potential for SLB leakage. The methodology for
calculating SLB leak rate uses a voltage-to-leakage correlation and
this methodology has previously been reviewed and approved by the
NRC. The SLB leakage value will be confirmed to be less than
allowable levels prior to return to power of Catawba Unit 1. No
unacceptable leakage is anticipated at normal operating or RCP
locked rotor conditions.
Therefore, as the existing tube integrity criteria and accident
analyses assumptions and results will continue to be met, the
proposed license amendment does not create the possibility at a new
or different kind of accident from any previously evaluated.
(3) The proposed license amendment does not involve a
significant reduction in [a] margin of safety.
The use of the voltage based bobbin probe interim tube support
plate elevation plugging criteria at Catawba Unit 1 is demonstrated
to maintain steam generator tube integrity commensurate with the
criteria of Regulatory Guide 1.121. [Regulatory Guide] 1.121
describes a method acceptable to the NRC staff for meeting GDCs
[General Design Criteria] 14, 15, 31, and 32, by reducing the
probability or the consequences of steam generator tube rupture.
This is accomplished by determining the limiting conditions of
degradation of steam generator tubing, as established by inservice
inspection, for which tubes with unacceptable cracking should be
removed from service. Implementation of the bobbin probe voltage
based interim tube plugging criteria of 1.0 volt is supplemented by
enhanced eddy current inspection guidelines to provide consistency
in voltage normalization, a 100% eddy current inspection at the tube
support plate elevations, and rotating pancake coil inspection
requirements for the larger indications left in service to
characterize the principle degradation as ODSCC. Even under the
worst case conditions, the occurrence of ODSCC at the tube support
plate elevations is not expected to lead to a steam generator tube
rupture event during normal or faulted plant conditions.
Based on the analyses for Cycle 8, the expected leakage values
and the leakage conditions required to be confirmed during accidents
creating high differential pressures across the steam generator
tubes (e.g. SLB), dose analysis confirm the maximum permissible
leakage will result in offsite dose consequences within the
guideline values. [An] MSLB accident with assumed leakage growth in
the faulted generator results in the EAB and LPZ doses remaining
within 10% of the 10 CFR 100 values of 25 Rem whole body and 300 Rem
thyroid for the accident-initiated iodine spike, and 10 CFR 100
values for the pre-accident iodine spike.
The distribution of crack indications at the tube support plate
elevations will be confirmed to result in acceptable primary to
secondary leakage during all plant conditions and that radiological
consequences are not adversely impacted.
Renewal of the tube support plate elevation plugging criteria
for operation at Catawba Unit 1 will decrease the number of tubes
which must be repaired by sleeving or taken out of service by
plugging. The installation of steam generator tube plugs reduce the
RCS flow margin. Thus, implementation of the alternate plugging
criteria will maintain the margin of flow that would otherwise be
reduced in the event of increased tube plugging.
Based on the above, it is concluded that the proposed license
amendment requested does not result in a significant reduction in
margin with respect to plant safety as defined in the Final Safety
Analysis Report or any Bases of the plant Technical Specifications.
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.
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 on 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.
Should the Commission takes 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 Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of the Federal Register notice. Written comments may also be
delivered to Room T-6 D22, 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 NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By March 13, 1995, 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 which is
available at the Commission's Public Document Room, the Gelman
Buidling, 2120 L Street, NW., Washington, DC, and at the local public
document room located at the York County Library, 138 East Black
Street, Rock Hill, South Carolina. If a request for a hearing or
petition for leave to intervene is filed by the above date, the
[[Page 7803]] 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 notice of hearing or an appropriate order.
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 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, Attention: Docketing and Service
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to Herbert N. Berkow: petitioner's name and telephone
number, date petition was mailed, plant name, and publication date and
page number of this Federal Register notice. A copy of the petition
should also be sent to the Office of the General Counsel, U.S. Nuclear
Regulatory Commission, Washington, DC 20555, and to Mr. Albert Carr,
Duke Power Company, 422 South Church Street, Charlotte, North Carolina,
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for hearing will not
be entertained absent a determination by the Commission, the presiding
officer or the presiding Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
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 dated November 29, 1994, as supplemented
January 12 and 27, 1995, which are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room
located at the New York County Library, 138 East Black Street, Rock
Hill, South Carolina.
Dated at Rockville, Maryland, this 3rd day of February 1995.
For the Nuclear Regulatory Commission.
Robert E. Martin,
Project Manager, Project Directorate II-3, Division of Reactor
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 94-3234 Filed 2-8-94; 8:45 am]
BILLING CODE 7590-01-M
