[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