ML20137Y920
| ML20137Y920 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/16/1997 |
| From: | Kalman G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137Y924 | List: |
| References | |
| NUDOCS 9704230283 | |
| Download: ML20137Y920 (10) | |
Text
_...
I
. 7590-01-P 1
i UNITED STATES NUCLEAR REGULATORY COMMISSION 1
ENTERGY OPERATIONS. INC.
1 DOCKET NO. 50-313 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 4
l corrosion indications that may exceed through-wall limits to remain in service until the next refueling outage.
t
}
The proposed amendment is being processed under exigent circumstances for i
the following. reason. During the IR13 refueling outage, an eddy current technique was used for the satisfactory completion of the ANO-1 steam generator inspection j
surveillance. The technique used had been qualified per Appendix H of the EPRI "PWR Steam Generator Tube Examination Guidelines." This technique was used to i
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.
1 9704230283 970416 PDR ADOCK 05000313 P
2 i
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 addy 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 th: plugging limit specified in the technical specifications. This would be considerec' 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 (N0ED). The N0ED was documented by 1etter 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.
b b
i 3
l Before issuance of the proposed license amendment, the Commission will have i
4 made findings required by the Atomic Energy Act of 1954, as amended (the Act) and l
the Commission's regulations.
j Pursuant to 10 CFR 50.91(a)(6) for amendments to be granted under exigent 4
j circumstances, the NRC staff must determine that the amendment request involves no l
significant hazards consideration. Under the Commission's regulations in i
1 l
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 j
or consequences of an accident previously evaluated; or (2) create the possibility i
of a new or different kind of accident from any accident previously evaluated; or i
i (3) involve a significant reduction in a margin of safety. As required by 10 CFR i'
50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
i An evaluation of the proposed change has been performed in accordance with j
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 j
to this amendment request follows:
i Criterion 1 - Does Not involve a Significant Increase in'the Probability or Consequences of an Accident Previously Evaluated.
1 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.
l 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 j
Report.
t The purpose of the periodic surveillance performed on the steam generator in i
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
4 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 perfonned on simulated flaws within the tubesheet it has been shown that the patch IGA indications within the upper tubesheet left in service during IR13 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 >atch IGA found in the ANO-1 upper tubesheet that the tubes will not fail at tiis 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 j
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 4
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.
i 4
l 5
Therefore, this change does agi create the possibility of a new or different kind of accident from any previously evaluated.
4 Criterion 3 - Does Not Involve a Significant Reduction in the Margin of i
j Safety.
The ANO-1 Technical Specification Bases specify that the surveillance 1
requirements (which includes the plugging limits) are to ensure the
)
structural integrity of this portion of the RCS pressure boundary. The technickl specification plugging limit of 40% of the nominal tube wall i
thickness requires tubes to be repaired or removed from service because the tube may become unserviceable prior to the next inspection. Unserviceable is i
defined in the technical specification as the condition of a tube if it leaks i
or contains a defect large enough to affect its structural integrity in the i
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 l
within the upper tubesheet is the MSLB.
1 1
Testing of tubes with representative IGA flaws removed from ANO-1 OTSGs j
during IR13, showed the flawed tubes to be capable of withstanding j
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 2
diameter within a tubesheet showed that the tubes always failed outside of j
the tubesheet. Thus the structural requirements listed in the bases of the i
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 D91 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 Egi 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.
~
-.-.. - ~. - -.. -. _ -. -
4 I
i 6
Y Normally, the Comission will not issue the amendment until the expiration of i
i i
the 14-day notice period. However, should circumstances change during the notice
+
i I
period, such that failure to act in a timely way would result, for example, in l
derating or shutdown of the facility, the Comission may issue the license 1
j amendment before the expiration of the 14-day notice period, provided that its 1
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 i
FEDERAL REGISTER a notice of issuance. The Commission expects that the need to take this action will occur very infrequently.
1 l
Written coments may be submitted hr
'l to the Chief, Rules Review 1
and Directives Branch, Division of Freedom of Information and Publications 4
l
. Services, Office of Administration, U.S. Nuclear Regulatory Comission, l
Washington, DC 20555-0001, and should cite the publication date and page number of i
this FEDERAL REGISTER notice. Written coments 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 coments received _ may l
be examined at the NRC Public Document Room, the Gelman Building, 2120 L Street, l
NW., Washington, DC.
l 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 i
i 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 j
and a petition for leave to intervene.
Requests for a hearing and a petition for i
l i
1 7
l 1 eave to intervene shall be filed in accordance with the Comission's " Rules of
}
Practice for Domestic Licensing Proceedings" in 10 CFR Part 2.
Interested persons
\\
j should consult a current copy of 10 CFR 2.714 which is available at the I
j Commission's Public Document Room, the Gelman Building, 2120 L Street, NW.,
j 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.
i s
i 8
l Not later than 14 days prior to the first prehearing conference scheduled in l
the proceeding, a petitioner shall file a supplament to the petition to intervene i
which must include a list of the contentions which are sought to be litigated in j
the matter. Each contention must consist of a specific' statement of the issue of l
law or fact to be raised or controverted.
In addition, the petitioner shall i
j provide a brief explanation of the bases of the contention and a concise statement l
of the alleged facts or expert opinion which support the contention and on which l
the petitioner intends to rely in proving the contention at the hearing.
The petitioner must also provide references to those specific sources and i
i documents of which the petitioner is aware and on which the petitioner intends to i
rely to establish those facts or expert opinion.
Petitioner must provide j
sufficient information to show that a genuine dispute exists with the applicant on i
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, l
if proven, would entitle the petitioner to relief. A petitioner who fails to file j
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 J
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 1
significant hazards consideration.
If a hearing is requested, the final determination will serve to decide when the hearing is held.
.~.-- - -. _.
~. -
i 9
If the final determination is that the amendment reo,uest involves no significant hazards consideration, the Comission 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 l
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 Comission, U.S. Nuclear Regulatory Comission, Washington, DC 20555-0001, Attention: Docketing and Services Branch, or may be delivered to the Comission'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 Comission 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 nu'ber, date m
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 Comission, Washington, DC 20555-0001, a'nd to Winston & Strawn, 1400 L Street, N.W., Washington, 20005-3502, attorney for the licensee.
Montimely 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 Comission, the presiding officer
10 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)(1)-(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 10th day of April,1997.
FOR THE NUCLEAR REGULATORY COMMISSION N
George Kal enior Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation I