ML20115C473
| ML20115C473 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/08/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20115C466 | List: |
| References | |
| NUDOCS 9210190393 | |
| Download: ML20115C473 (15) | |
Text
[psom*o, UN!TED STATliS
- t, NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RMATED TO AMENDMENT NO. S5_ TO FAClllTY OPERATif{G LICENSE NO. NPF-2 SOUTHERN NUCLEAR OPERATING COMPANY. INC.
JOSEPH M. FARLEY NVCLEAR PLANT. UNIT 1 DOCKET NO. 50-348 l-l l
l.
1.0
'NTRODUCTION By letter dated August 24, 1992 (Reference 1), as supplementad on September 29, 1992 (Reference 15), Southern Nuclear Operating Company, n.-
(the licensee), submitted a request for changes to the Joseph M. Farley ;Mear L
Plant (Farley), Unit 1, Technical Specifications.
The requested amendment revises Technical Specification 4.4.6.4, and Bases 3/4.4.6, to allow the I
l implementation of interim steam generator (SG) tube plugging criteria for defects located at the tube support plate (TSP) elevations.
Leakage subsequent to a main steamline break (SLB) is not expected to exceed the currently postulated value of one gallon per minute (gpm).
This amendment is only applicable for the twelf th operating cycle.
The September 29, 1992, letter provided clarifying information concerning SG l
tube pull commitments that did not change the initial proposed no significant hatards consideration determination.
2.0 BACKGROUND
Previous inservice inspections and examinations of the SG tubes at Farley, l
Units 1 and 2, have identifled intergranular stress corrosion cracking (IGSCC) on the outer diameter of the tubes at the TSP intersections.
The licensee refers to this particular form of IGSCC as outer diameter stress corrosion cracking (00 SCC).
ODSCC activity at TSP intersections is a common degradetion phenomenon in SGs l
in a number of nuclear power plants. Approximately 21 tubes, including 57 tube-to-TSP intersections, have been removed from affected SGs across the industry for examination and testing. These include one tube (including 1 TSP i
intersection) from Farley, Unit 1, and seven tubes (13 TSD intersections) i l
removed from Farley, Unit 2.
Each of these pulled tube TSP intersections was sectioned and metallographically examined.
In general, these examinations I
have revealed multiple, segmented, and axial cracks with short lengths for the deepest penetrations.
The 00 SCC is generally confined to within tLe thickness of the TSPs, consistent with the corrosion mechanism which involves the concentration of impurities, including caustics, in the tube-to-TSP crevices.
The Nuclear Regulatory Commission staff (the staff), notes that there is some l
potential for shallow ODSCC for a short distance above or below the TSP.
This has been observed for two of the pulled TSP intersections, including one from k bo$$t,g PCR
. Farley, Unit 1, which exhibited short, very shallow (<10%) crack segments extending 0.25 inch above the TSP.
The pulled tube specimens from Farley, Units 1 and 2, to date have shown minimal intergranular attack (IGA) associated with the_0DSCC.
However, more significant IGA has been' observed to occur with ODSCC on some pulled' tube specimens from other plants.
These results suggest that the degradation develops as IGA plus stress corrosion cracking (SCC), particularly when maximum IGA depths greater than 25% are found.
A large number (>100) of axial cracks aroun" the circumference are commonly found on these tubes.
The maximum depth of-IGA is typically 1/2 to 1/3 of the SCC depth.
Patches of cellular IGA /0DSCC formed by combined axial and circumferential orientation of microcracks are frequently found in pulled tube examinations.
The staff notes, however, that the axial crack segments have been the dominant flaw feature affecting the structural integrity of the pulled tube specimens as evidenced by results of burst tests (discussed in Section 4.3) performed for 29 of the pulled TSP intersections prior to sectioning.
Technical Specification 4.4.6.4.a.6, Plugging or Repair Limit, requires thu tubes with imperfections exteeding 40% of the nominal tube wall thickness be repaired by sleeving or removed from service by plugging.
The licensee stated that this repair criterion would result in unnecessary removal of significant numbers of SG tubes from service.
To preclude this, the licensee developed proposed alternative plugging criteria (APC) that.was submitted by letter dated February 26, 1991 (Reference 2).
This proposal was revised by-le+ter dated November 13, 1991 (Reference 3), (1) to respond to questions and comments from the staff transmitted by letter dated August 8, 1991 (Reference 4), and (2) to reflect additional pulled tube information from the Trojan Nuclear Plant.
The proposed APC involves a voltage amplitude limit of_four volts, as measured by the industry standard eddy current bobbin coil probe (referred to herein as a bobbin) using the 400/100 KHz mix differential channel, in lieu of the current 40% depth-based plugging or repair limit.
These criteria would only apply to axially oriented 00 SCC degradation confined to within the thickness of the TSPs.
Staff comments and questions concerning the November 13, 1991, APC proposal were provided by letter dated January 29, 1992 (Reference 5).
The licensee responded to these questions and comments during a meeting on February 6, 1992.
By letter dated February 20, 1992 (Reference 6), the licensee submitted an updated technical support document for the APC, WCAP-12871, Revision 2 (Reference 7). WCAP-12871, Revision 2, is intended to support a 3.6 volt repair limit; however, the. licensee has not yet revised its 4.0 volt APC proposal in Reference 2.
In their August 24, 1992, letter, the licensee requested interim modifications to the tube repair limits in the Farley, Unit 1, Technical Specifications for-the twelfth operating cycle only, pending completion of the staff's review of the APC proposal.
The proposed modifications to the tube repair limits are described in detail in Section 3.0 of this Safety Evaluation, and include a one volt repair criterion for flaws confined to the thickness of the TSP in lieu of the currently applicable depth-based limit of 40%.
The August.24, 1992, letter also' submitted WCAP-13464 (Reference 8), which supports an alternate approach to that contained in WCAP-12871, Revision 2, for calculating SLB leakage as a function of end-of-cycle voltage.
By letter dated October 29, 1991 (Reference 9), as supplemented on July 1, 1992 (Reference 10), the licensee submitted a Technical Specification amendment request to reduce the total allowable primary-to-secondary operational leakage as a result of the discovery of circumferential flaws in the tube sheet expansion re~gion of Farley, Unit 1, SG tubes.
The requested amendment was issued on October 1, 1992.
This amendment reduced leakage from any one steam generator from 500 gallons per day (gpd) to 140 gpd.
The total allowable primary-to-secondary operational leakage through all SGs was reduced from 1 gpm (1440 gpd) to 420 gpd. This permanent leakage limit for Unit 1 is more restrictive than the 150 gpd limit associated with the requested interim plugging criteria. Therefore, no temporary changes to primary-to-secondary leakage limits are required as part of this amendment request.
By letter dated February 20, 1992 (Reference 6), as supplemented on March 27, 1992 (Reference 11), a similar interim plugging criteria amendment was rer;uested for the ninth operating cycle of Farley, Unit 2.
The requested amendment was subsequently issued as Amendment No. 87 to Facility Operating License No. NPF-8, by NRC letters dated April 1 and 22, 1992 (References 12 and 13).
3.0 PROPOSED INTERIM PLUGGING CRITERIA Farley, Unit 1, Technical Specification 4.4.6.4.a.6, Plugging or Repair Limit, and Bases 3/4.4.6, Steam Generators, are revised to specify that the repair limit at the TSP intersections for the twelfth operating cycle is based on the analysis in WCAP-12871, Revision 2, as supported by WCAP-13464, to maintain SG tube serviceability as described below:
a.
An eddy current inspection using a bobbin for inspection of 100%
of the hot and cold leg SG TSP intersections will be performed for tubes in service.
b.
Degradation within the bounds of the TSP with a bobbin voltage a
less than or equal to 1.0 volt will be allowed to remain in service.
c.
Degradation within the bounds of the TSP with a bobbin voltage greater than 1.0 volt will be repaired or plugged except as noted in d. below, d.
Indications of potential degradation within the bounds of the TSP with a bobbin voltage greater than 1.0 volt, but less than or equal to 3.6 volts, may remain in service if a rotating pancake coil probe (RPC) inspection does not detect or confirm the degradation.
Indications of degradation with a bobbin vo~.tage greater than 3.6 volts will be plugged or repaired.
in addition to the above Technical Specification changes, the licensee also made the following commitments:
a.
A sample RPC inspection of at least 100 TSP intersections will be performed.
All intersections with a bobbin dent voltage exceeding five volts will be inspected.
Other intersections in the sample
.e 4 population will be based on inspecting intersections with artifact indications and intersections with unusual phase angles, Expansion of the sample plan, if required, will be based on the nature and number of 'he indications, b.
RPC flaw indications not found by the bobbin due to masking effects, (due to denting, artifact indications, and noise), will be plugged or repaired.
c.
By letter dated September 29, 1992, the licensee committed to removing and analyzing SG tubes in accordance with the following priorities and guidelines:
Priority ': ODSCC within the bounds of the TSP, with a bobbin voltage greater than 3 volts.
Priority 2: ODSCC within the bounds of the TSP.
If two pullable tubes are found in Priority 1, two tubes will be-pulled to provide the maximum number of indications.
If two Priority 1 pullable tubes are not found, one tube will be pulled to provide the maximum number of indications.. This tube will be from Priority 1, _if possible, or if no Priority.1 tubes are found, then a Priority 2 tube will be pulled. Additional guidelines are contained in the September 29, 1992, letter.
Results of the analysis of the tube pull will be provided to the NRC staff within_
two months of the Unit 1 startup.
4.0 EVALUATION 4.1.
Inspection Issues In support of the proposed interim repair limit, the licensee proposes to utilize the eddy current te'st guidelines provided in Attachment 4 of the August 24, 1992, letter to ensure that the field bobbin indication voltage measurements-are obtained in a manner consistent with how the voltage limit was developed.
These: guidelines define the bobbin specifications, _ calibration requirements, specific: acquisition and analyses criteria, and flaw recording guidelines to be used in the inspection of the SGs. All flaw indict.tions, regardless of voltage amplitude, will be recorded. With these commitments, the staff-finds the licensee's eddy current test guidelines to be acceptable.
The staff finds that the proposed bobbin inspection program is consistent with the development of voltage-based repair limits; namely, the establishment of the relationship between burst pressure and bobbin voltage.
In addition, the
~. licensee states in its August 24, 1992, submittal that it will perform an RPC sample inspection of tubes at TSP intersections. The program will include RPC inspection of locations _where the dent voltage is greater than five valts as measured by the bobbin and_RPC inspection of TSP intersections with artifact indications or_ indications with unusual phase angles.
This sample program-will be performed on at least 100-TSP intersections.
The RPC can provide improved resolution of flaw indications as compared to the bobbin probe in the presence of dents and artifacts and is sensitive to both axial and '
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circumferential flaws.
The licensee states that the sampling program will be expanded as necessary, based on the nature and number of flaws discovered, in addition, RPC flaw indications not foo d by the bobbin due to masking effects (due to denting, artif act indications, or noise), will be plugged or repaired.
The Farley, Unit 1, eddy current test guidelines in Attachment 4 of the August 24, 1992, letter require the RPC inspection of TSP intersections exhibiting bobbin indications exceeding 1.0 volt.
The RPC inspections will permit better characterization of the indications found by the bobbin to confirm or deny the existence of any actual.ube degradation.
The proposed repair limit is based 1
on axially oriented ODSCC as the dominant degradation mechanism with some IGA involvement.
The proposed limit is alto based on the premise that any significant degradation is confined to the TSP.
The licensee has agreed to inform the staff prior to Cycle 12 operation of any unforeseen RPC findings relative to the characteristics of the flaws at the TSPs.
This includes any detectable circumferential indication or detectable indications extending outside the thickness of the TSP.
A safety analysis of these unexpected findings will also be provided.
4.2 lub.e Integrity i ssuq[
i The purpose of the Technical Specification tube repair limits is to ensure that tubes accepted for continued service will retain adequate structural and leakage integrity during normal operating, traninnt, and postulated accident conditions, consistent with General Design criteria (GDC) 14, 15, 31 and 32 of 10 CFR Part 50, Appendix A.
Structural integrity refers to maintaining adequate margins against gross failure, rupture, and collapse of the SG tubing.
Leakage integrity r$fers to limiting primary-to-secondary leaka " to within acceptable limits.
The traditional strategy for accomplishing these objectives has been to establish a minimum wall thickness requirement in accortance with the structural criteria of Regulatory Guide 1.121, " Basis for Plugging Degraded PWR Steam Generator Tubes." Allowance for eddy current measurement error and flaw growth between inspections hac been added to the minimum wall thickness requirements (consistent with the Regulatory Guide) to arrive at a depth-based repair limit.
Enforcement of a minimum wall thickness requirement would impliciti> orve to ensure leakage integrity (during normal operation and accidents), a1. wil as structural integrity.
It has been recognized, however, that defects, especially cracks, will occasionally grow entirely through-wall and develop small leaks.
For this reason, tight limits on allowable primary-to-secondary leakage have been established in the Technical Specifications to. ensure timely plant shutdown before adequate structural and leakage integrity of the affected tube is impaired.
The proposed tube repair limitt for Farley Unit.1, consist of voltage amplitude criteria rather than the traditional depth-based criteria.
- Thus, the proposed repair criterion represtats a departure from the past p' actice of explicitly enforcing a minimum wall thickness requirement.
The pulled tube examina w show that for bobbin indications at or near 1 volt (i.e., the propos
- limit) maximum crack depths range between 20% and 98%
through.ial)-
.ne likelihood of thrmh-wall or near through-wall crack pen trations appears to increase with increasing voltage amplitude.
For indications at or near 3.6 volts (i.e., the APC limit), the maximum crack depths have been found to range between 90% and 100% through-wall.
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l 1 i found to contain "non-repairable" in61 cat ons h ugh-ay develop through-wall and near t roming il many of the tubes which will be i
The staf f's under the proposed interim criter a m tulated SLB accidents.
wall crack penetrations during the upcofor leakage 'uring norm iteria from a structural and leakage tions 4.3 and 4.4 of this Safety evaluation of the proposed repair crintegrity standpoint is provid Evaluation, respectively.
51rgG1. ural inRHrjlY 4.3 4.3.1 Ent. Inteqrily h/ voltage correlation to demonstrate posed 1.0 volt interim repair The licensee has developed a burst strengtthat bobbin iterion will retain adequate structural sistent with the criteria of Regulatory criterion or the 3.6 volt APC cr includes the burst margins during Cycle 12 operation, conThe burst strength / voltage correlatio These pulled tube data are C
laboratory tube specimens containing ODSC Guide 1.121.
pressure versus field bobbin voltage 1SPs. including one TSP it 1.
The i ulated field conditions.
supplemented by 30 data points from burst pressure / voltage correlation flaws produced in model boiler tests under s r>
tion standard voltage set-ups andt w bobbin voltage data used to construct the have been normalized to reflect calibra voltage measurement procedures consisten 2, Anpendix A.
y among the voltage data in the i
Analysis Guidelines in WCAP-12871, Revis on in addition, ensures consisttmey that this normalization ensuru consistenc l tion and the field voltage measurements burst pressure / voltage correlation and,between the vo at f arley, Unit 1 of Regulatory Guide 1.121 is that three against burst at normalfor f arley, Unit The most lidting burst pressure criterion degraded tubes shall retain a margin of From the berst the tube, criterion of 4380 psi.
operatthg dif ferential pressure across i um voltage which will satisfy this translates to a limiting burst pressure The 3.6 interval is 6,2 volts.
pressure / voltage correlation, the max m i ion 2, is intended to support, includes d
burst pressure criterion at a 95% confi ence tainty and for a 50% increase in volt APC limit, which WC AP-12871, Rev s lhe NDE measurement uncertainty an allowance for 20% NDE mesurement uncer voltage during the next operating cyc e,tainties stemming from l
estimate considered measurement uncer f fects centering), tion standards, and characteristics, bubbin wear (which a (ASME) calibra The staff tion of the signal voltage.
American Society of Mechanical Engineers 7,
nd Analysis Cuidelines (Reference variability in the analysts' interpreta d into the f arley, Unit 1, eddyin mini concurs that the NDE Data Acquisition a Appendix A), which have been incorporatecurrent tes Based on implementation of th mdist i
they apply to the interim criteria.
The bin indication) has been developed.
guidelines, a cumulative pr.obability ltage measurements is comervative with16% a uncertainty (applicable to each bob b bility value assumed 20% uncertainty in the vo respect to the upper 90% cumulative pro af rom the c in
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l many of the tubes which will be found to contain "non-repairable" indications under the proposed interim criteria may develop through-wall and near through-wall crack penetrations during the upcoming cycle, thus creating the potential for leakage during normal operation and postulated SLB accidents.
The staff's evaluation of the proposed repair criteria from a structural and leakage integrity standpoint is provided in Sections 4.3 and 4.4 Of this Safety Evaluation respectively.
4.3 Structural Intearit y 4.3.1 [Lutst Intearit.y The licensee has developed a burst strength / voltage correlr
~ to demenstrate that bobbin indications satisfying the proposed 1.0 volt in, repttr e
criterion or the 3.6 volt APC criterion will retain adequats
- <uctural margins during Cycle 12 operation, consistent with the critar'a of Regulatory Guide 1.121.
The burst strength / voltage correlation includes tM burst pressure versus field bobbin voltage data (pre-pull values) for 29 pMled tube ISPs, including one tsp from f arley, Unit 1.
These pulled tube data are supplemented by 30 data points from laboratory tube specimens containing ODSCC flaws produ m in model boiler tests under simulated field conditions.
The bobuin voltage data used to construct the burst pressure / voltage correlation have been normalized to reflect calibration standard voltage set-ups and voltage measurement procedures consistent with the NDE Data Acquisition and Analysis Guidelines in WCAP-12871, Revision 2, Appendix A.
The staff finds that this normalization ensures consistency among the voltage data in the burst pressure / voltage correlation and, in addition, ensures consistency between the voltage data in the correlation and the field voltage measurements at farley, Unit 1.
The most limiting burst pressure criterion of Regulatory Guide 1.121 is that degraded tubes shall retain a margin of three against burst at normal operating differential pressure across the tube.
For Farley, Unit 1, this translates to a limiting burst pressure criterion of 4380 psi.
From the burst pressure / voltage correlation, the maximum voltage which will satisfy this burst pressure criterion at a 95% confidence interval is 6.2 volts.
The 3.6 vu.t APC limit, which WCAP-12871, Revision 2, is intended to support, includes an allowance for 20% NDE measurement uncertainty and for a 50% increase in voltage during the next operating cycle.
The NDE measurement uncertainty estimate considered measurement uncertainties stemming from bobbin design characteristics, bobbin wear (which affects centering), variability among American Society of Mechanical Engineers (ASME) calibration standards, and variability in the analysts' interpretation of the signal voltage.
The staff concurs that the NDE Data Acquisition and Analysis Guidelines (Reference 7, Appendix A), which have been incorporated into the Farley, Unit 1, eddy current test gutdelines, will be effective in minimizing th) uncertainties as they apply to the interim criteria.
Based on implementation of these guidelines, a cumulative pr.obability distribution of the residual measurement uncertainty (applicable to each bobbin indicM ion) has been developed.
The assumed 20% uncertainty in the voltage measurements is conservative with respect to the upper 90% cumulative probability salue of 16% as determined from the cumulative probability distribution.
, Potential flaw growth between inspections has been evaluated based on cbserved voltage amplitude changi during Cycles 8 and 9 at farley, Unit 1.
Specifically, the eddy current data from the 1988 and 1989 inspections were re-examined for each indication reported during the inspection in 1991 using a consistent data analysis procedure.
This examination showed that nearly all of the 1991 indications were traceable back to the 1988 and 1989 inspections.
The average percent changes in voltage, considering the entire data set, were 43X between 1988 and 1989 and 37% between 1989 and 1991.
These averages conservatively treat negative voltage changes as zero changes, if the data set i s restricted to voltage changus where the initial indication exceeded 0.75 volts, the average voltage changes are smaller, e.g.,
17% between 1989 and 1991.
The 50% average voltage growth allowance used to support the 3.6 volt APC limit is intended to provide margins for variation in future growth rates at Farley, Unit 1.
For any specific individual tube, NDE measurement uncertainty and/or voltage growth may exceed the values assumed in thr above deterministic basis for the 3.6 volt APC repair limit, since the deteruinistic basis does not considor the tails of the voltage measurement uncertainty and voltage growth distributions.
In addition, the burst pressure for some tubes may be less than the 95% and/or the 99% lower prediction interval values in the burst pressure / voltage correlation.
The staff notes that Regulatory Guide 1.121 provides no guidance on the appropriate cumulative probability and prediction interval values to be employed in these types of analyses.
The staff is continuing to evaluate this issue as part of the ongoing lead plant review of the alternate plugging criteria.
The licensee proposes that these uncertainties be directly accounted for by use of Monte Carlo methods to demonstrate that the probability of burst during St.B accidents is acceptably low for the distribution of voltage indications being left in service. Under this approach, the beginning-of-cycle (B00) indications left in service are projected to the end-of-cycle (EOC) by randomly sampling the probability distributions for MDE uncertainties and voltage growth per cycle.
For each MC Monte Carlo sample of b.obbin voltage, the burst pressure / voltage correlation is randomly sampled to obtain a burst pressure.
The 100,000 Monte Carlo samples are performed for the entire BOC distribution.
The probability of tube burst at SLB is obtained as the sum of the samples resulting in burst pressures less than the SLB pressure differential of 2650 psi divided by the number of times the distribution of indications left in service is sampled.
This kind of Monte Carlo analysis was performed for the distribution of indications found during a previous (i.e., 1990) inspection at Farley, Unit 2.
l This analysis indicated that implementation o'.
3.6 volt repair criterion at 4
l of approximately 3x10',ielded a conditional probability of burst, given an SLB, that time would have y A similar estimate was not presented for Unit 1 in -
L WCAP-12871, Revision 2.
However, the staff review indicates that the comparable estimate for Unit 1 is anticipated to be of the_same order of magnitude as that for Unit 2.
The staff concurs that this is an extremely low probability, three orders of magnitude less than the value considered in a statf generic risk assessmeno for SGs (NUREG-0844).
The staff concludes that the proposed 1.0 volt interim criterion will provide adequate assurance that tubes with indications whicn are accepted for continued service will meet the burst pressure criteria of Regulatory Guide
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l.121-for this cycle.
The staff notes that for the proposed 1 volt interim plugging criteria, the projected F0C voltage is 1.76 volts assuming the 90%
cumulative 1.121 probability values of voltage measurement ancertainty and voltage growth, and is 2.41 volts assuming the 99% cumulative probability values.
The staff further notes that the bounding value of voltage l
growth / cycle at f arley, Units 1 and 2, since 1987, has not exceeded 2.6 volts.
i The staff estimates this 2.6 volts to represent a bounding value, assuming no increase in corro lon rates over what has been observed previously at farley, Units 1 and 2.
Assuming a 20% voltage measurement uncertainty (upper 95%
confidence value determined by the,itensee) for a 1.0 volt indication left in 3
service, the EOC voltage is expected by the staff to be bounded by 3.8 volts.
This is substantially below the 6.2 voltage limit evaluated by the licensee as the lower 95% confidence limit for meeting the most limiting burst pressure criterion (i.e., three times normal operating pressure differential).
The staff notes, however, that this allowable voltage (i.e., 6.2 volts) reduces to approximetely 3.3 volts (staff estimate) using the lower 99% predictinn interval curve.
Finally, the !,icensee is proposing as part of the interim repair ctiteria that indications with bobbin voltages greater than 1.0 volt, but less than or equal to 3.6 volts, remain in service if RPC inspection does not confirm the indication.
The staff notes that short and/or relatively shallow cracks that cre detectable by the bobbin may sometimes not be detectable by RPC, although IN RpC is considered by the staff to be more sensitive to longer, deeper flaws which are of structural significance.
The staff further notes that burst strength is not a unic,ue function of voltage, rather for a given voltage there is a statistical distribution of possible burst strengths as indicated in the burst pressure / voltage correlation.
The staff concludes that burst pressures for bobbin indications which were not confirmed by RPC will tend to be at the upper end of the burst pressure distribution.
The 3.6 volt cutoff, such that all bobbin indications would be plugged or repaired (with or without confirming RPC indications), provides additional assurance that all excessively degraded tubes will be r& moved from service.
Thus, the staff finds the proposed exception to the 1.0 volt criterion to be acceptable.
4.3.2 Combined Accident loadinas The licensee has evaluated the effects of combined safe shutdown earthquake (SSE) and loss-of-coolant accident (LOCA) loads and SSE plus SLB loads on tube integrity, consistent with GDC 2 of 10 CFR Part 50, Appendix A.
A combined LOCA plus SSE must be evaluated for potential yielding of the TSPs which could result in subsequent dcformation of the tubes.
If significant tube deformation should occur, primary flow area could be reduced and postulated cracks in tubes could open up which might create the potential for in-leakage 1
(i.e., secondary-to-primary under LOCA conditions.
In-leakage during LOCA would pose a potential conce)rn since it may cause an increase in the core pe clad temperature (PCT).
I The most limiting accident conditions for tube deformation considerations result from the combination of SSE and LOCA loads.
The seismic excitation defined for SGs is in the form of t.cceleration response spectra at the S_G supports.. In the seismic analysis, the licensee has used generic response spectra which envelop the farley-specific response spectra.
A finite element lU-
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model of the Series 51 SG was developed and the analysis was performed using the WECAN computer program.
The mathematical model consisted of three dimensional lumped mass, beam, and pipe elements as well as general matrix input to represent the piping and support stiffness.
Interactions at the TSP shell and wrapper /shell connections were represented by concentric spring-gap dynamic elements.
Impact damping was used to account for energy dissipation at these locations.
LOCA loads, developed as a result of transient flow following a postulated primary coolant pipe break, were calculated for five different pipe break locations.
These included three large and two minor pipe breaks.
The large pipe break locations evaluated were the SG inlet and outlet lin a and the reacter coolant pump outlet line, wnile the minor pipe breaks analyzed were two branch line breaks.
Prior qualification of the farley, Unit 1, primary piping for leak-before-break requirements resulted in the limiting LOCA event being the break of a minor branch line.
The licensee, however, has used the loads for the primary piping break as a conservative approximation.
The principal tube loading during a LOCA is caused by the rarefaction wave in the primary fluid.
This wave initiates at the postulated break location and travels around the tube U-bends. A differential pressure is created across the two legs of the tube which causes an in-plane horizontal motion of the U-bends and induces significant lateral loads on the tubes.
The pressure time histories needed for creating the differential pre.ssure across the tube are obtained from transient thermal-hydraulic analyses using the MULTIFLEX computer code.
For the rar.efaction wave induced Icadings, the predominant motion of the U-bends is in the plane of the U-bend.
Thus, the individual tube motions are not coupled by the anti-vibration bars and the structural analysis is performed using single tube models limited to the U-bend and the straight leg region over the top two TSPs.
In addition to the rarefaction wave loading discussed above, the tube bundle is subjected to bending loads during a LOCA.
These loads at due to the shaking of the SG caused by the break hydraulics and reactor coolant loop motion.
However, the resulting TSP loads from this motion are small compared to those due to the rarefaction wave induced motion.
To obtain the LOCA induced hydraulic forcing functions, a dynamic blowdown ana'ysis is performed to obtain the system hydraulic forcing functions assuming an instantaneous (1.0 msec break opening time), double-ended guillotine break.
The hydraulic forcing functions are then applied, along with the displacement time-history of the reactor pressure vessel (obtained from a separate reactor vessel blowdown analysis) to a system structural model that includes the SG, the reactor coolant pump, and the primary piping.
This analysii yields the time-hisry displacements of the SG at its upper lateral and lower support nodes.
The>e time-history displacements formulate the forcing functions for obtaining the tube stresses due to LOCA shaking of the SG.
In calculating a combined TSP load, the licensee combined the LOG rarefaction and LOCA shaking laads directly, while the LOCA and SSE loads were combined using the square root of the sum of the squares.
The staff found this combination methodology acceptable.
The overall TSP load was transferred to the SG shell through wedge groups located at discrete locations around the plate circumference.
! l The radial loads due to combined LOCA and SSE could potentially result in yielding of the TSP at the wedge supports, causing some tubes in the vicinity of the wedge supports to be deformed. Utilizing results from recent tests and analysis programs, the licensee has shown that tubes will undergo permanent deformation if the change in diameter exceeds a minimum threshold value.
This threshold for tube deformation is related to the cont.ern for tubes with pre-i existing tight cracks that could potentially open during a combined LOCA plus SSE event.
For Farley, Unit 1, the LOCA plus SSE loads were determined to bc of such magnitude that none of the tubes (which are assumed to contain pre-existing tight cracks) are predicted to exceed this deformation threshold value and, therefore, will not lead to significant tube leakage.
The licensee has assessed the effect of SSE bending stresses on the burst strength of tubes with axial cracks.
Tensile stress in the tube wall would tend to close the cracks while compressive stress would tend to open the cracks. On the basis of previously performed tests, the licensee has concluded that the burst strength of tubes with through-wall cracking'is not affected by an SSE event.
i Based on a review of the information provided by the licensee, the staff concludes that at farley, Unit 1 no significant tube deformation or leakage is likely to occur during an SSE plus LOCA event, in addition, burst strength of tubing with through-wall cracks is not.ffected by an SSE event, 4.4 Leakane Inteority As discussed earlier, a number of the indications satisfying the proposed interim 1.0 volt repair limit can be expected to have or to develop through-wall and/or near through-wall crack penetratiuns during the next cycle, thus creating the potential for primary-to-secondary leakage during normal operation, transients, or postulated accidents.
The staff finds that adequate leakage integrity during normal operating conditions is assured by the restrictive Technical Speci'fication limits on allowable primary-to-secondary leakage as discussed in Section 4.5 of this Safety Evaluation.
Adequate leakage integrity during transients and postulated accidents is demonstrated by showing that for the most limiting accident assumed to occur at the end of the next cycle, the resulting leakage will not exceed the rate assumed in the farley, Unit 1, design basis accident.
Th! licensee has identified in the Final Safety Analysis Report (FSAR),
Chrpter 15, accidents that result in secondary steam release, and thus whose co, sequences could be affected by the extent of primary-to-secondary leakage, O' these accidents, the SLB, with respect to the interim plugging criteria, nas determined to be the most limiting.
In this case, since the SG in the faulted loop is subject to dryout, the activity release path is conservatively assumed _to be direct to the environment, without any mitigation resulting from mixing with secondary liquid coolant in the SG.
In support of the 3.6 volt APC proposal, the licensee has submitted an SLB leak rate / radiation dose analysis in WCAP-12871, Revision 2, that was revised by letter dated June _4, 1992 (Reference 14).
This analysis. which is still under staff review, is intended to demonstrate an allowable SLB primary-to-secondary leak rate of 5.7 gpm.
For the purpose of supporting the interim
- repair limit proposal, tne licensee has proposed that the maximum allowable primary-to-secondary leak rate during SLB be 1.0 gpm, which is consistent with the assumed leak rate in the FSAR design basis analysis.
Therefore, there is no change in the offsite dose as a result of the use of the interim repair limit.
The staff concurs that 1.0 gpm is an acceptable primary-to-secondary leak rate limit for the interim plunging criteria.
The SLB leakage calculation model in WCAP-12871, Revision 2, utilizes a correlation between leakage test data obtained under simulated SLB conditions (at a given 1SP location), and the corresponding normalized bobbin voltage (SLB leakage / voltage correlation).
The SLB leakage data includes 27 drte points from the model boiler :pecimens described earlier and 7 data points from pulled tube specimens.
The calculation method involves establishing the voltage distribution of the indications being accepted for continued service.
Probability distributions of voltage measurement uncertainty, voltage growth / cycle, and SLB leak rate versus voltage are accounted for by Monte Carlo techniques in predict.ing the distribution of E0C voltages and the associated SLB leakage for the BOC distribution of indications.
SLB leakage is evaluated at the 90% cumulative probability level.
Based or the voltage distributicns found during a previous inspection (1990) at Farley, Unit 2, and assuming implementation of the 3.6 volt repair criterion, the estimated leakage during a postulated SLB at EOC 8 was 0.4 gpm at the 90X cumulative probability level, well within the current licensing basis of 1.0 gpm.
The estimate is almost entirely made up of leakage from indications whose BOC voltages were greater than 2 volts.
Utilizing the same model, the licensee has datermined that about 4000 B0C indications at 2 volts would be necessary to produce an SLB leak rate of 1.0 gpm (at 90% cumulative probability) at EOC.
In support of tt-t 1 volt interim repair criterion for Farley, Unit 1, the licensee will update the above analyses to consider the distribution of voltages for indications satisfying the 1 volt criterion during t'e eleventh refueling outage inspection.
The analysis will also reflect the distribution of voltage changes observed during Cycle 11 (i.e. 1991 to 1992).
In addition to the above analysis, the licensee committed to also verify the 1.0 gpm at E0C for SLB using deterministic calculation method.
The method will consist of the following:
Determine the EOC voltage distribution in terms of the number of indications falling into each of the following E0C voltage ranges:
s 2.5 volts
> 2.5 to 4 volts
> 4 volts Acceptable methods for determining the EOC voltage distribution include:
The methodology described in WCAP-12871, Revision 2.
(This involves sampling of the cumulative probability distributions of f;DE measurement error and of voltage growth during the most recent operating cycle using Monte Carlo techniques and applying the results to the beginning-of-cycle (B0C) voltage distribution.)
1
4 A simplified approach may be used as an alternative (to the WCAP-12871, Revision 2 approach) provided it provides for a conservative treatment of the tails of the cumulative probability t
distributions of NDE measurement error and of voltage growth to the 100% cumulative probability values.
SLB leakage as a function of EOC voltage shall be determined as follows:
EOC Voltaae ELB Leakage s 2.5 volts 0
> 2.5 to 4 volts 1 liter / hour
> 4 volts 10 liters / hour The staff's approval of the proposed interim repair limit is based on the licensee's being able to demonstrate that acceptance of all bobbin indications satisfying the 1.0 volt criterion will not create the potential for leakage in excess of the 1.0 gpm licensing basis for a postulated SLB.
Should the calt.ulated SLB leakage exceed the 1.0 gpm limit, tubes with the highest voltages will be plagged or repaired until the 1.0 gpm leakage rate is met.
The licensee has committed to submit a report with the results of the SLB leakage analysis prior to restart from the Unit 1 eleventh refueling outage.
4.5 Jnterim leakane limits 4.5.1 Description Implementation of the interim plugging criteria requires a reduction in the Technical Specification reactor coolant system leakage limits that would usually be applicable for the or.e operating cycle to which the interim criteria apply.
Specifically, for the interim criteria, the normal 500 gpd limit for primary-to-secondary leakage through any one SG would be reduced to 150 gpd.
In addition, the limit on total leakage through all SGs would be reduced from 1.0 gpm (1440 gpd) to 450 gpd.
Leakage during an SLB would not exceed the current design basis of 1.0 gpm.
However, by Amendment No. 94 to Facility Operating License No. NPF-2, reactor coolant system leakage limits for Farley, Unit 1, were permanently reduced in Technical Specification 3.4.7.2 to 140 gpd primary-to-secondary leakage through any one SG, and 420 gpd total through all of the SGs.
These leakage limits were imposed for reasons unrelated to the interim plugging criteria and are more restrictive than those required by the interim plugging criteria.
4.5.2 Discussion The normal 500 gpd limit per SG is intended to ensure that through-wall cracks which leak at rates up to this limit during normal operation will not propagate and result in tube rupture under postulated accident conditions consistent with the criteria of Regulatory Guide 1.121.
The current 1.0 gpm limit for total primary-to-secondary leakage is consistent with the assumptions used in the FSAR design basis accident analyses.
L Development of the 150 gpd per SG interim leakage limit has utilized the l
extensive industry data base regarding burst pressure as a function of crack length and leakage during normal operation.
Based on leakage evaluated at the lower 95% confidence interval for a given crack size, the proposed 150 gpd limit would be exceeded before the crack length reaches the critical crack e
length for SLB pressures.
Based on nominal, best estimate leakage rates, the 150 gpd limit would be exceeded before the crack length reaches the critical crack length corresponding to a burst pressure of three times normal operating pressure.
The interim leakage limits are more restrictive than normal operating limits in order to provide a margin of safety against rupture.
The interim leakage limits are also intended to provide an additional margin to accommodate a rngue crack which might grow at much greater than expected rates, or unexpectedly extend outside the thickness of the TSP, and thc; provide additional protection against exceeding SLB leakage limits.
However, the current farley, Unit 1, leakage limits contained in lechnical Specification
' 4.7.2 are more restrictive than the required interim leakage limits.
Therefore, the staff finds the existing leakage limits in Technical Specification 3 ~ 7.2 to be acceptable for implementation of the interim plugging criteria.
4.6 henty Based on the above evaluation, the staff concludes that the proposed interit tube repair limits will ensure adequate structural and leakage integrity of the SG tubing at f arley, Unit 1, consistent with applicable regulatory requirements.
The staff's approval of the proposed interim repair limit is based on the licensee's being able to demonstrate that acceptance of all indications satisfying the 1.0 volt criterion will not create the potential for leakage in excess of the 1.0 gpm licensing basis for a postulated SLB occurring at E0C 12.
STATE CON 5tATAT10N in accordance with the Commission's regulations, the State of Alabama official was notified of the proposed issuance of the amencment.
The State official had no comments.
t 6.0 ENVIRONMENTAL CONSjDERATION The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements.
The NRC staff.has determined that the amendment involves no significant increase in the amounts, and no significant change _ in the types, of _ any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (
),
Accordingly, the amendment meets the eligibility criteria for categorical exclusion-set forth in 10 CFR St.22(c)(9).
Pursuant to 10 CFR Sl.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
_.__.____.a,__
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, 7.0 CONCLUS10B The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will n,t be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the is3uance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
8.0 REFERENCES
1.
Southern Nuclear Operating Company letter dated August 24, 1992, "Josepn M. Farley Nuclear Plant - Unit 1 Steam Generator Tube Support Plate Interim Plugging Criteria."
2.
Alabama Power Company' letter dated February 26,1991, " Joseph M. Farley Nuclear Plant - Units I and 2, Steam Generator (SG) Tube Support Plate (TSP) Alternate Plugging Criteria (APC)."
3.
Alabama Power Company letter dated November 13, 1991-, " Joseph M. Farley Nuclear Plant, Steam Generator (SG) Tube Support Plate (TSP) Alternate Plugging Criteria (APC)."
4.
MRC letter to Alabama Power Company dated August 8, 1991, " Request for Additional Information Concerning Steam Generator Tube Support Plate Alternate Tube Piugging Criteria for Joseph M. Farley Nuclear Plant, Units 1 and 2."
5.
NRC letter to Southern Nuclear Operating Company dated January 29, 1992,
" Steam Generator Tube Support Plate Alternate Plugging Criteria for
(
Joseph Farley Nuclear P1, int, Units 1 and 2."
6.
Southern Nuclear Operating Company letter dated February 20, 1992, v
" Joseph M. Farley Unit 2, Steam Generator Tube Support Plate Interim Plugging Criteria."
7.
Westingho'ise Reports WCAP-12871, Revision 2 (Proprietary Version) and WCf.P-12872, Revision 2 (Non-Proprietary Version), "J.M. Farley Units '.
and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates."
8.
Westi% house Reports WCAP-13464 (Proprietary Version) and WCAP-13465 (Non-Proprietary Version), " Response to NRC Questions on f arley Steam Generator Tube Alternate Plugging Criteria Presentation Materials,"
August 1992.
9.
Alabama Power Company letter datid October 29,1991, " Joseph M. Farley Nuclear Plant - Unit 1, Steam Generator Tube Leakage Technical Specification "
10.
Southern Nuclear Operating Company letter dated July 1, 1992, "J.
M.
Farley Nuclear Plant - Unit 1, Steam Generator Tube Leakage Technical Specification."
l
' 11.
Southern Nuclear Operating Company letter dated March 27, 1992, " Joseph M. farley Nuclear Plant - Unit 2, Interim Plugging Criteria."
12.
NRC letter to Southern Nuclear Operating Company dated April 1, 1992,
" Issuance of Amendment No. 87 to facility Operating License No. NPF-2 Regarding S+eam Generator Tube Interim Plugging Criteria - Joseph M.
)
Farley Nuclear Plant, Unit 2 (TAC ho. M82810)."
13.
f4RC letter to Sou+hern Nuclear Operating Company dated April 22, 1992,
" Correction to Technical Specifications for Amendment No. 87 for Joseph M. Farley Nuclear Plant, Unit 2 (TAC No. M82810)."
14.
Southern Nuclear Operating Company letter dated June 4, 1992, " Joseph M.
Farley Nuclear Plant, Roquest for Additional Information Concerning the Steam Generator Alternate Plugging Criteria."
15.
Southern Nuclear Operating Company letter dated September 29, 1992.
" Interim Plugging Criteria."
Principal contributor:
S. Hoffman E. Murphy K. Karwoski Date: October 8, 1992 I
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