ML20057F497
| ML20057F497 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/30/1993 |
| From: | Tim Reed Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUDOCS 9310180100 | |
| Download: ML20057F497 (12) | |
Text
.
f o
7590-01 6
UNITED STATES NUCLEAR REGULATORY COMMISSION i
SOUTHERN NUCLEAR OPERATING COMPANY. INC.
DOCKET NO. 50-364 NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENT TQ FACILITY OPERATING LICENSE. PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION. AND OPPORTUNITY FOR A HEARING l
The U.S. Nuclear Regulatory Comission (the Comission) is considering issuance of an amendment to Facility Operating License No. NPF-8 issued E
to Southern Nuclear Operating Company, Inc. (the licensee), for operation of the Joseph M. Farley Nuclear Plant, Unit 2 (Farley 2), located in Houston l
County, Alabama.
l l
The proposed amendment would modify the Technical Specifications (TS), on an interim basis, to replace the percent tube wall degradation criteria currently employed with a more appropriate method for determining serviceability of tubes with flaw indications at the tube support plate intersections.
The staff finds that exigent circumstances exist in that a failure to act promptly would result in a delay in startup from the current Farley Unit 2 refueling outage. The requested TS amendments are required to support performance of the surveillances neccesary to declare the Farley Unit 2 steam generators operable prior to proceeding to mode 4.
Before issuance of the proposed license amendment, the Comission will j
have mad? findings required by the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations.
150014
~
9310180100 930930 3Fo1 PDR ADOCK 050003e4 P
PDR l((
' Pursuant to 10 CFR 50.91(a)(6) for amendments to be granted under l
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 i
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 i
j significant hazards consideration, which is presented below:
(1) Operation of Farley Unit 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Testing of model boiler specimens for free standing tubes at room temperature conditions show burst pressures as high as approximately 5000 psi for indications of outer diameter stress corrosion cracking with voltage measurements as high as 26.5 volts.
Burst testing performed on pulled tubes with up to 10 volt indications show burst 1
pressures in excess of 5900 psi at room temperature.
Correcting for the effects of temperature on material properties and minimum i
strength levels (as the burst testing was done at room temperature),
tube burst capability significantly exceeds the [ Regulatory Guide]
R.G. 1.21 criterion requiring the maintenance of a margin of three times normal operating pressure differential on tube burst if through-wall cracks are present.
Based on the existing data base, this criterion is satisfied with bobbin coil indications with signal amplitudes several times the 1.0 volt interim plugging criteria, l
regardless of the indicated depth measurement. This structural limit is based on a lower 95% confidence level limit of the data.
The 1.0 threshold volt criteria provides an extremely conservative margin of safety to the structural limit considering expected growth rates of ODSCC [outside diameter stress corrosion cracking] at Farley.
Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by a burst pressure to voltage correlation. However, relative to expected leakage during normal operating conditions, no field leakage has been reported from tubes with indications with a voltage level of under 7.7 volts for a
. 3/4 inch tube with a 1.0 volt correlation to 7/8 inch tubing (as compared to the 1.0 volt proposed interim tube plugging limit).
Relative to the expected leakage during accident condition loadings, the accidents that are affected by primary-to-secondary leakage and steam release to the environment are Loss of External Electrical Loan and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Secondary System Pipe Failare, Steam Generator Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control Rod Drive Mechanism Housing. Of these, the Major Secondary System Pipe Failure is the most limiting for Farley in considering the potential for off-site doses.
The offsite dose analyses for the
{
other events which model primary-to-secondary leakage and steam release from the secondary side to the environment assume that the secondary side remains intact.
The steam generator tubes are not subjected to a sustained increase la differential pressure, as is j
the case following a steam line break event. This increase in differential pressure is responsible for the postulated increase.in l
leakage and associated offsite doses following a steam like break event.
Upon implementation of the interim plugging criteria, it j
l must be verified that the expected distribution of cracking i
indications at the tube support plate intersections are such that primary-to-secondary leakage would result in site boundary dose within the current licensing basis.
Data indicate that a threshold voltage of 2.8 volts would result in through-wall cracks long enough to leak at SLB [ steam line break] conditions. Application of the proposed plugging criteria requires that the current distribution of a number of indications versus voltage be obtained during the refueling outages.
The current voltage is then combined with the rate of change in voltage measurement to establish an end of cycle voltage distribution and, thus, leak rate during SLB pressure i
differential.
If it is found that the potential SLB leakage for degraded intersections planned to be left in service is greater than the current licensing bisis limit, then additional tubes will be plugged or repaired to reduce SLB leakage potential to within the acceptance limit.
(2)
The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Implementation of the proposed interim tube support plate elevation steam generator tube plugging criteria does not introduce any significant changes to the plant design basis.
Use of the criteria does not provide mechanism which could result in an accident outside of the region of the tube support plate elevations.
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). The bobbin probe signal amplitude plugging criteria is established such that operational leakage or excessive leakage during a postulated steam line break condition is not
l D '
l anticipated. SNC [ Southern Nuclear Operating Company] will implement a maximum leakage rate limit of 150 gpd per steam generator on Unit 2 to help preclude the potential for excessive leakage during all plant conditions upon application of the plugging criteria.
The R.G. 1.121 criterion for establishing operational leakage rate limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture.
The 150 god limit provides for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is associated with the longest permissible crack length.
R.G. 1.121 acceptance criteria for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exceeded. The longest permissible crack is the length that provides a factor of safety of three against bursting at normal perating pressure differential. A voltage amplitude of i
I approximately 4 volts for typical ODSCC corresponds to meeting this tube burst requirement at the 95% prediction interval on the burst correlation. Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by the burst pressure versus voltage correlation.
Consequently, typical burst pressure versus through-wall crack length correlations are used below to define the " longest permissible crack" for evaluating operating leakage limits.
The single through-wall crack lengths that result in tube burst at three times normal operating pressure differential and SLB conditions are about 0.42 inch and 0.84 inch, respectively. Normal leakage for these crack lengths would range from 0.11 gallons per minute to 4.5 gallons per minute, respectively, while lower 95%
confidence level leak rates would range from about 0.02 gallons per minute to 0.02 gallons per minute to 0.6 gallons per minute, respectively, l
An operating leak rate of 150 gpd will be implemented in application of the tube plugging limit.
This leakage limit provides for detection of 0.4 inch long cracks at nominal leak rates and 0.6 inch long cracks at the lower 95% confidence level leak rates. Thus, the 150 gpd limit provides for plant shutdown prior to reaching critical crack lengths for SLB conditions at leak rates less than a lower 95%
confidence level and for three times normal operating pressure differential at less than nominal leak rates.
(3)
The proposed license amendment does not involve a significant reduction in margin of safety.
~
The use of the interim tube support plate elevation plugging criteria is demonstrated to maintain steam generator tube integrity commensurate with the requirements of R.G. 1.121.
R.G. 1.21
I 1 describes a method acceptable to the NRC staff for meeting GDCs
[ General Design Criteria] 2, 14, 15, 31, and 32 by reducing the probability of the consequences of steam generator tube rupture.
j This is accomplished by determining the limiting conditions of i
degradation of steam generator tubing, as established by inservice inspection, for which tubes with unacceptable cracking should be removed from service. Upon implementation of the criteria, even I
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. The most limiting effect would be a possible increase in leakage during a steam line break event.
Excessive leakage during a steam line break event, however, is precluded by verifying that, once the criteria are applied, the expected end of cycle
~
distribution of crack indications at the tube support plate
)
elevations would result in minimal, and acceptable primary to secondary leakage during the event, and hence, help to demonstrate radiological conditions are less than a small fraction of the 10.CFR l
100 guideline.
In addressing the combined effects of LOCA + SSE [ loss-of-coolant l
accident plus safe shutdown earthquake] on the steam generator component (as required by GDC 2), it has been determined that tube l
collapse may occur in the steam generators at some plants. This is the case as the tube support plates may become deformed as a result of lateral loads at the wedge supports at the periphery of the plate due to either the LOCA rarefaction wave and/or SSE loadings. Then, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.
Additionally, the margin to burst for the tubes using the interim plugging criteria is comparable to that currently providing by existing technical specifications.
There are two issues associated with steam generator tube coll'pse.
First, the collapse of steam generator tubing reduces the RCS flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA which, in turn, may potentially increase Peak Clad Temperature (PCT).
- Second, there is a potential the partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse.
Consequently, a detailed leak-before-break analysis was performed and it was concluded that the leak-before-break methodology (as permitted by GDC4) is applicable to the Farley Unit 1 and 2 reactor coolant system primary loops and, thus, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in the structural design basis of the plant.
Excluding breaks in the RCS primary loops, the LOCA loads from the large branch line breaks were analyzed at Farley Unit 1 and 2 and were
1 j
4 !
found to be of insufficient magnitude to result in steam generator 4
tube collapse or significant deformation.
4 j
Regardless of whether or not leak-before-break is applied to the i
primary loop piping at Farley, any flow area reduction is expected j
to be minimal (much less than 1%) and PCT margin is available to account for this potential effect.
Based on analyses results, no tubes near wedge locations are expected to collapse or deform to the degree that secondary to primary in-leakage would be increased over current expected levels.
For all other steam generator tubes, the possibility of secondary-to-primary leakage in the event of a LOCA +
SSE event is not significtnt.
In actuality, the amount of
]
secondary-to-primary leakage in the event of a LOCA + SSE is i
expected to be less than that currently allowed, i.e., 500 gpd per steam generator. Furthermore, secondary-to-primary in-leakge would be less than primary-to-secondary leakage for the same pressure differential since the cracks would tend to tighten under a secondary-to-primary pressure differential. Also, the presence of the tube support plate is expected to reduce the amount of in-i leakage.
Addressing the R.G.1.83 considerations, implementation of the tube plugging criteria is supplemented by 100% inspection requirements at the tube support plate elevations having ODSCC indications, reduced operating leak rate limits, eddy current inspection guidelines to provide consistency in voltage normalization, and rotating pancake coil inspection requirements for the larger indications left in service to characterize the principal degradation mechanism as ODSCC.
4 As noted previously, implementation of the tube support plate elevation plugging criteria will decrease the number of tubes which must be taken out of service with tube plugs or repaired.
The installation of steam generator tube plugs or tube sleeves would reduce the RCS flow margin, thus implementation of the interim plugging criteria will maintain the margin of flow that would otherwise be reduced through increased tube plugging or sleeving.
l The NRC staff has reviewed the licensee's analysis and, based on this j
review, it appears that the three standards of 10 CFR 50.92(c)'are satisfied.
Therefore, the NRC staff proposes to determine that the i
amendment request involves no significant hazards consideration.
The Commission is seeking public comments on this proposed 2
e
_. _. - _ - -. - - ~. - - - -, ~ - - - - - - -
4 i
i i
determination. Any coments received within 15 days after the date of t
j publication of this notice will be considered in making any final W
determination.
Normally, the Commission will not issue the amendment until the i
j 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 j
amendment involves no significant hazards consideration. The final determination will consider all public and State coments received.
Should the Commission take this action, it will publish in the FEDERAL REGISTER a j
notice of issuance. The Comission expects that the need to take this j
j action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and i
\\
Directives Branch, Division of Freedom of Information and Publications l
Services, Office of Administration, U.S. Nuclear Regulatory Comission, a
}
Washington, DC 20555, and should cite the publication date and page number of i
this FEDERAL REGISTER notice. Written comments may also be delivered to Room i
j P-223, Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland, from 7:30
]
a.m. to 4:15 p.m. Federal workdays. Copies of written coments received may i
be examined at the NRC Public Document Room, the Gelman Building, 2120 L E
i Street, NW., Washington, DC 20555.
The filing of requests for hearing and petitions for leave to intervene is discussed below.
i By DCT 2 01993
, the licensee may file a request for a i
, - - ~,
-B-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 i
Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555 and at the local public document room located at Houston-Love Memorial Library, 212 W. Burdeshaw Street, P. O. Box 1369, Dothan, Alabama 36302.
If a request for a hearing or petition for leave to intervene is file'd 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
i
- 4 t
l
' possible effect of any order which may be entered in the proceeding on the i
petitioner's interest. The petition should also identify the specific i
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 j
prehearing conference scheduled in the proceeding, but such an amended j
petition must satisfy the specificity requirements described above.
i 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 i
a specific statement of the issue of law or fact to be raised or i
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
{
s the petitioner intends to rely in proving the contention at the he. ring.
j 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 i
be limited to matters within the scope of the amendment under consideration.
The contention must be one which, if proven, would entitle j
1
_~
t
> 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 Comission 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 Comission, U.S. Nuclear Regulatory Comission, Washington, DC 20555, 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 20555, by the above date. Where petitions are filed during the last 10 days of the notice
)
a 1
1 :
i period, it is requested that the petitioner promptly so inform the Commission by a toll-in ee 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 j
addressed to [ APPROPRIATE PD]: petitioner's name and telephone number, j
date petition was mailed, plant name, and publication (ate 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 [ ADDRESS OF ATTORNEY]
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be j
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)(1)-(v) and 2.714(d).
For further details with respect to this action, see the application for amendment dated May 28, 1993, as supplemented July 29, 1993, September 14, 1993 and September 22, 1993, which is available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW.,
Washington, DC 20555 and at the local public document room located at
..,y mm...
,.m...,
,,,,...a,
n b !
Houston-Love Memorial Library, 212 W. Burdeshaw Street, P. O. Box 1369, Dothan, Alabama 36302, i
Dated at Rockville, Maryland, this 30th day of September 1993.
FOR THE NUCLEAR REGULATORY COMMISSION Timothy A. Reed, Project Manager Project Directorate 11-1 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation O
f 1
i e
c
-