[Federal Register Volume 62, Number 77 (Tuesday, April 22, 1997)]
[Notices]
[Pages 19628-19631]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-10332]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-313]


Entergy Operations, Inc.; 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-51, issued to Entergy Operations, Inc. (the licensee), for 
operation of Arkansas Nuclear One, Unit 1, located in Pope County, 
Arkansas.
    The proposed amendment would permit steam generator tubes with 
intergranular corrosion indications that may exceed through-wall limits 
to remain in service until the next refueling outage.
    The proposed amendment is being processed under exigent 
circumstances for the following reason. During the 1R13 refueling 
outage, an eddy current technique was used for the satisfactory 
completion of the ANO-1 steam generator inspection surveillance. The 
technique used had been qualified per Appendix H of the EPRI ``PWR 
Steam Generator Tube Examination Guidelines.'' This technique was used 
to depth size all intergranular attack flaws within the upper 
tubesheet. As required by the technical specifications, all upper tube 
sheet IGA indications with a depth size of greater than the plugging 
limit as determined by the qualified sizing technique, were also 
removed from service by plugging.
    During the steam generator inspections, three tube samples 
containing upper tubesheet IGA flaws were removed from the ``B'' OTSG 
and sent offsite to be analyzed for future development of an alternate 
repair criteria and to further support the qualified eddy current 
sizing technique employed during refueling outages. The preliminary 
destructive examination results were recently received by the ANO 
staff. This data arrived approximately 5 months after the resumption of 
operation following the steam generator inspections that occurred 
during 1R13. These results indicate that the flaw depths do not 
correlate well with the depths sized using the qualified eddy current 
technique. Upon further review, ANO has determined that the application 
of the sizing criterion is no longer valid. With the qualified sizing 
technique invalidated, there is a potential that tubes could have been 
left in service with indications that have through-wall depths greater 
than the plugging limit specified in the technical specifications. This 
would be considered a condition that is not allowed by the technical 
specifications. Prior to the receipt of the preliminary destructive 
examination results, ANO had no reason to question the adequacy of the 
steam generator inspections that occurred during 1R13.
    Based on the developments described above, on April 9, 1997, the 
NRC verbally issued a Notice of Enforcement Discretion (NOED). The NOED 
was documented by letter dated April 11, 1997. The NOED expressed NRC's 
intention to exercise discretion in enforcing compliance with portions 
of the technical specifications related to steam generator tubes. The 
NOED will remain in effect until an exigent technical specification 
amendment is processed but in no case later than May 7, 1997.

[[Page 19629]]

    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 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:

    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion 
of these standards as they relate to this amendment request follows:
    Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The steam generators are used to remove heat from the reactor 
coolant system during normal operation and during accident 
conditions. The steam generator tubing forms a substantial portion 
of the reactor coolant pressure boundary. A steam generator tube 
failure is a violation of the reactor coolant pressure boundary and 
is a specific accident analyzed in the ANO-1 Safety Analysis Report.
    The purpose of the periodic surveillance performed on the steam 
generator in accordance with ANO-1 Technical Specification 4.18, is 
to ensure that the structural integrity of this portion of the 
reactor coolant system (RCS) will be maintained. The technical 
specification (TS) plugging limit of 40% of the nominal tube wall 
thickness requires tubes to be repaired or removed from service 
because the tube may become unserviceable prior to the next 
inspection. Unserviceable is defined in the TS as the condition of a 
tube if it leaks or contains a defect large enough to affect its 
structural integrity in the event of an operating basis earthquake, 
a loss-of-coolant accident, or a steam line break.[sic] Of these 
accidents, the most severe condition with respect to patch 
intergranular attack (IGA) degradation within the upper tube sheet 
is the main steam line break (MSLB). During this event the 
differential pressure across the tube could be as high as 2500 psid. 
The rupture of a tube during this event could permit the flow of 
reactor coolant into the secondary coolant system thus bypassing the 
containment.
    From testing performed on simulated flaws within the tubesheet 
it has been shown that the patch IGA indications within the upper 
tubesheet left in service during 1R13 with potential depths greater 
than the plugging limit, do not represent structurally significant 
flaws which would increase the probability of a tube failure beyond 
that currently assumed in the ANO-1 Safety Analysis Report.
    Burst tests were conducted on tubing with simulated flaws within 
the tubesheet. In these tests, through-wall holes of varying sizes 
up to 0.5 inch in diameter were drilled in test specimens. The 
flawed specimen tubes were then inserted into a simulated tubesheet 
and pressurized. In all cases the tube burst away from the flaw in 
that portion of tube that was outside the tubesheet. The size of 
these simulated flaws bound the indications left in service within 
the upper tubesheet during 1R13. These tests demonstrate for flaws 
similar to the patch IGA found in the ANO-1 upper tubesheet that the 
tubes will not fail at this location under accident conditions.
    The dose consequences of a MSLB accident are analyzed in the 
ANO-1 accident analysis. This analysis assumes the unit is operating 
with a 1 gpm steam generator tube leak and that the unit has been 
operating with 1% defective fuel.
    Increased leakage during a postulated MSLB accident resulting 
from the patch IGA left in service in the upper tube sheet is not 
expected. IGA has been present in the ANO-1 steam generators for 
many years with no known leakage attributed to this damage 
mechanism. Because of its localized nature and morphology, the flaw 
does not open under accident pressure conditions.
    This change allows continued operation with IGA indications 
within the upper tube sheet with the potential of through-wall 
depths greater than the technical specification plugging limit. 
Continued operation with these flaws present does not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated for ANO-1.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2--Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The steam generators are passive components. The intent of the 
technical specification surveillance requirements are being met by 
this change in that adequate structural and leakage integrity will 
be maintained. Additionally, the proposed change does not introduce 
any new modes of plant operation.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The ANO-1 Technical Specification Bases specify that the 
surveillance requirements (which includes the plugging limits) are 
to ensure the structural integrity of this portion of the RCS 
pressure boundary. The technical specification plugging limit of 40% 
of the nominal tube wall thickness requires tubes to be repaired or 
removed from service because the tube may become unserviceable prior 
to the next inspection. Unserviceable is defined in the technical 
specification as the condition of a tube if it leaks or contains a 
defect large enough to affect its structural integrity in the event 
of an operating basis earthquake, a loss-of-coolant accident, or a 
MSLB.[sic] Of these accidents the most severe condition with respect 
to IGA within the upper tubesheet is the MSLB.
    Testing of tubes with representative IGA flaws removed from ANO-
1 OTSGs during 1R13, showed the flawed tubes to be capable of 
withstanding differential pressures in excess of 10,000 psid without 
the presence of the tubesheet. Testing of simulated through-wall 
flaws of up to 0.5 inch in diameter within a tubesheet showed that 
the tubes always failed outside of the tubesheet. Thus the 
structural requirements listed in the bases of the technical 
specification is satisfied considering this change.
    Leakage under accident conditions would be limited due to the 
small size and morphology of the flaws and would be low enough to 
ensure offsite dose limits are not exceeded.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    In conclusion, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that 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 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 14 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 14-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 14-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

[[Page 19630]]

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 Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications 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 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 May 22, 1997, 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 Tomlinson Library, Arkansas Tech 
University, Russellville, AR 72801. 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 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 14 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 14 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 20555-0001, 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 Dr. William Beckner: 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 20555-0001, 
and to Winston & Strawn, 1400 L Street, N.W., Washington, 20005-3502, 
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 April 11, 1997, 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 Tomlinson Library, Arkansas Tech 
University, Russellville, AR 72801.

    Dated at Rockville, Maryland, this 16th day of April, 1997.


[[Page 19631]]


    For the Nuclear Regulatory Commission.
George Kalman,
Senior Project Manager, Project Directorate IV-1, Division of Reactor 
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 97-10332 Filed 4-21-97; 8:45 am]
BILLING CODE 7590-01-P