[Federal Register Volume 61, Number 61 (Thursday, March 28, 1996)]
[Notices]
[Pages 13888-13890]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-7674]



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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-302]


Florida Power Corporation; 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. 
DPR-72, issued to Florida Power Corporation, et al. (the licensee), for 
operation of the Crystal River Nuclear Generating Plant, Unit No. 3 
(CR3 or the facility) located in Citrus County, Florida.
    Currently, the technical specifications (TS) for CR3 relating to 
the Once Through Steam Generator's (OTSG's) tube inspection acceptance 
criteria, specify repair limit for removing steam generator tubes from 
service. This repair limit is based on a structural evaluation of a 
simplified model of tubes with uniform through wall (T/W) thinning. A 
recent tube-pull examination at CR3 identified a number of low signal-
to-noise (S/N) tube eddy current indications. The licensee indicated 
that these S/N indications are a substantially different morphology 
from the model used to develop the current TS inspection and acceptance 
limit. As a result of the small signal amplitude associated with these 
S/N indications, they could not be accurately sized by conventional 
bobbin coil phase angle.
    By letter dated May 31, 1995, proposed TS changes which involved a 
broad and long-term criteria addressing both wear and Inter-Granular-
Attack (IGA) degradation mechanisms. The licensee's May 31, 1995 
request was noticed in the Federal Register on July 5, 1995 (60 FR 
35071). By letter dated March 21, 1996, the licensee superseded its May 
31, 1995 request and proposed a more focused TS change which would be 
applicable for one cycle duration, and only to Inter-Granular-Attack 
(IGA) degradation mechanisms in a limited region of the OTSG. 
Accordingly, this supersedes that notice in its entirety.
    Specifically, the licensee proposed to:
    A. Revise TS 3.4.12 item d, to read: ``150 gpd primary to secondary 
LEAKAGE through any one steam generator (OTSG).''
    B. Revise TS 5.6.2.10.2, page 5.0-14, ``The results of each sample 
inspection shall be classified into one of the following three 
categories:'' to read: ``The results of each bobbin coil sample 
inspection shall be classified into one of the following three 
categories:''
    C. Revise the Note in TS 5.6.2.10.2, page 5.0-14, ``In all 
inspections, previously degraded tubes whose degradation has not been 
spanned by a sleeve must exhibit a significant increase in the 
applicable imperfection size measurement (>+0.3V bobbin coil amplitude 
increase for first span IGA indications or >10% further wall 
penetration for all other imperfections) to be included in the below 
percentage calculations.''
    D. Revise the second sentence in TS 5.6.2.10.4.a.2, page 5.0-16, 
``Eddy-current * * * as imperfections'' to read: ``Any indication below 
all degraded tube criteria specified in item below may be considered as 
imperfections.''
    E. Revise TS 5.6.2.10.4.a.4, page 5.0-16, to read: ``Degraded Tube 
means a tube containing a first span IGA indication with a bobbin coil 
amplitude [greater than or equal to] 0.65V, an axial extent of [greater 
than or equal to] 0.13 inch, or a circumferential extent of [greater 
than or equal to] 0.3 inch or other imperfections [greater than or 
equal to] 20% of the nominal wall thickness caused by degradation 
except where all such degradation has been spanned by the installation 
of a sleeve.''
    F. Add TS 5.6.2.10.4.a.7 ``First span Inter-Granular-Attack (IGA) 
indication means a bobbin coil indication located between the lower 
tubesheet secondary face and the first tube support plate confirmed by 
MRPC to have a volumetric morphology characteristic of IGA.''
    G. As a result of adding the new TS 5.6.2.10.4.a.7 above, revise 
applicable TS to reflect the new ``first span IGA definition'' term. 
Renumber 5.6.2.10.4.a.8 and 9 to 5.6.2.10.4.a.9 and 10.
    H. Renumber TS 5.6.2.10.4.a.7 to TS 5.6.2.10.4.a.8, and revise to 
read: ``Plugging/Sleeving Limit means the extent of degradation beyond 
which the tube shall be restored to serviceability by the installation 
of a sleeve or removed from service because it may become unserviceable 
prior to the next inspection. The limit for first span IGA indications 
is a bobbin coil amplitude of 1.25V, an axial extent of 0.25 inch, or a 
circumferential extent of 0.6 inch. The limit for indications other 
than first span IGA is equal to 40% of the nominal tube or sleeve wall 
thickness. No more than five thousand sleeves may be installed in each 
OTSG.''
    I. Revise TS 5.7.2.c.2, page 5.0-29, to read: ``Following each 
inservice inspection of steam generator (OTSG) tubes, the NRC shall be 
notified of the following prior to plant ascension into Mode 4.
    1. Number of tubes plugged and sleeved
    2. Crack like indications in the first span
    3. An assessment of growth in the first span indications, and
    4. Results of in-situ pressure testing, if performed.
    The complete results of the OTSG tube inservice inspection shall be 
submitted to the NRC within 90 days following the completion of the 
inspection. The report shall include:
    1. Number and extent of tubes inspected,
    2. Location and percent of wall-thickness penetration for each 
indication of an imperfection,
    3. Location, bobbin coil amplitude, and axial and circumferential 
extent (if determined) for each first span IGA indication, and
    4. Identification of tubes plugged and tubes sleeved.''
    The licensee requested that the above proposed license amendment be 
processed as an emergency or exigent amendment to prevent delay of the 
restart of the facility which is currently in an refueling outage. The 
licensee described the circumstances involving the request and stated 
that its request meets the requirements of 10 CFR 50.91a (5) and (6). 
The licensee stated that the complexity of the issues involved, 
differences between the licensee's and the industry's approach, and 
evolving industry/NRC interactions on the steam generator integrity 
issues resulted in a longer than anticipated NRC staff review time of 
the licensee's previous submittal (May 31, 1995). As a result, staff 
review of the licensee's May 31, 1995 submittal has not been completed. 
Therefore, the licensee proposed this more limited license amendment as 
described herein. 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.
    Pursuant to 10 CFR 50.91(a)(6) for amendments to be granted under 
exigent circumstances, the NRC staff must determine 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

[[Page 13889]]
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. The proposed change will not significantly increase the 
probability or consequences of an accident previously evaluated. The 
relevant accidents are excessive leakage or steam generator tube 
rupture (as a consequence of MSLB [Main steam Line Break] or 
otherwise).
    RG [Regulatory Guide] 1.121 establishes a standard method for 
demonstrating structural integrity under worse-than-DBE [design 
basis Event] conditions. The existing TS is based on this RG. The 
first span, IGA disposition strategy continues to rely on this 
guidance. Current TW [through wall] sizing techniques would allow 
defects greater than the current TS limit of 40% to remain in 
service since these techniques do not accurately measure percent 
wall penetration for small volume indications. The proposed 
disposition strategy is based on measurable eddy current parameters 
of voltage, axial extent, and circumferential extent has been shown 
to provide a higher confidence that unacceptable flaws are removed 
from service. Therefore, the probability of a Steam Generator Tube 
Rupture (SGTR) is not increased and may well be decreased by 
implementation of this S/N disposition strategy.
    The probability of OTSG tube leakage during normal operation or 
accident conditions is not adversely affected by the proposed S/N 
disposition strategy. Operating history indicates essentially no 
primary to secondary leakage through the OTSG tubes at CR-3. Growth 
rate studies imply this trend could be expected to continue. 
However, for conservatism the OTSG leakage limit has been reduced 
from 1 gallon per minute through all OTSGs to 150 gallons per day 
through any one OTSG. This change is consistent with the guidance 
provided in Generic Letter 95-05. Small volume indications which 
might leak during worse-case FWLB [Feedwater Line Break] conditions 
are addressed in the RG 1.121 evaluation. The disposition strategy 
ensure these indications are removed from service as part of the 
inservice inspection. Once detected, the proposed criteria is at 
least as effective in determining those indications which should be 
removed from service as are the existing TS limits.
    The first span IGA disposition strategy is an integral part of 
an overall effort to better address these and similar phenomena in 
OTSGs.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The key `new or different' accidents addressed in this and 
similar proposals is the potential for MSLB-induced multiple SGTR or 
excessive primary-to-secondary leakage during such events. While 
these events are addressed in CR-3 Emergency Operating Procedures 
(EOPs), they are beyond those licensed for the facility.
    However, as noted above, the probability of MSLB induced 
multiple SGTR is reduced by more effective screening and plugging/
sleeving criteria. The probability of detection and identification 
of tubes which should be removed from service is maintained or 
improved by the S/N disposition strategy. The likelihood of adverse 
effects from plugging sound tubes is reduced. The operation of the 
OTSG or related structures, systems or components is otherwise 
unaffected.
    3. The proposed change will not involve a significant reduction 
to any margin of safety.
    The margins of safety defined in RG 1.121, including the 
required pressure used in the structural analysis, are retained. The 
probability of detecting degradation is unchanged since bobbin coil 
methods will continue to be the primary means of initial detection. 
The probability of leakage remains acceptably small. The proposed S/
N disposition strategy is an enhancement to the inservice inspection 
of OTSG tubing that will provide a higher level of confidence that 
tubes exceeding the allowable limits are repaired while sound tubes 
are left in service. Based upon results of the various growth rate 
studies, the probability of an accident at the end of cycle is 
essentially the same as the beginning.

    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.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 15 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 15-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 15-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 take this action, it will publish in the Federal 
Register a notice of 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 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 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 April 29, 1996, 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 
Building, 2120 L Street, NW., Washington, DC, and at the local public 
document room located at the Coastal Region Library, 8619 W. Crystal 
Street, Crystal River, Florida 32629. 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 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 a party to the proceeding; (2) the

[[Page 13890]]
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 the amendment is issued before the expiration of the 30-day 
hearing period, the Commission will make a final determination on the 
issue of no significant hazards consideration. If a hearing is 
requested, 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 Services 
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 Eugene V. Imbro: 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 A. H. Stephens, 
General Counsel, Florida Power Corporation, MAC--A5D, P. O. Box 14042, 
St. Petersburg, Florida 33733, 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 March 21, 1996, which is 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 Coastal Region Library, 8619 W. Crystal 
Street, Crystal River, Florida 32629.

    Dated at Rockville, Maryland, this 25th day of March 1996.

    For the Nuclear Regulatory Commission.
Bart C. Buckley,
Acting Director, Project Directorate II-1, Division of Reactor 
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 96-7674 Filed 3-27-96; 8:45 am]
BILLING CODE 7590-01-P