ML20126C752
| ML20126C752 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/10/1992 |
| From: | Kokajko L Office of Nuclear Reactor Regulation |
| To: | Danni Smith OREGON, STATE OF |
| References | |
| NUDOCS 9212230200 | |
| Download: ML20126C752 (2) | |
Text
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December 10, 1992-e
' Docket "o 50-344 Mr. David-Stewart-Smith Oregon Department of: Energy _
Salem, Oregon 97310
Dear Mr. Stewart-Smith:
SUBJECT:
INTERIM PLUGGING CRITERIA FOR TROJAN NUCLEAR PLANT--
Enclosed are two documents for your review. The first document (Enclosure 1) is a memorandum from L. C. Shao to E. S. Beckjord dated December 9, 1992, entitled, " Interim Plugging Criteria for Trojan Nuclear Plant " The-second document (Enclosure 2) is a memorandum from J. Hopenfeld tc G. Burdick dated December 9,1992, entitled, " Reply to your request for comments on Draft 'RES Position on Steam Generator Tube Integrity' by L. C. Shao."
Both of these documents have been placed on the docket file for Trojan Nuclear.
Pl ant.
As such, both documents have been placed in the Public Document Room.
Sincerely, Original signed by Lawrence E. Kokajko, Senior Project Manager Project Directorate V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
As stated DISTRIBUTION:
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December 10, 1992 Docket No. 50-344 Mr. David Stewart-Smith Oregon Department of Energy Salem, Oregon 97310
Dear Mr. Stewart-Smith:
SUBJECT:
INTERIM PLUGGING CRITERIA FOR TROJAN NUCLEAR PLANT Enclosed are two documents for your review. The first document (Enclosure 1) is a memorandum from L. C. Shao to E. S. Beckjord dated December 9, 1992, entitled, " Interim Plugging Criteria for Trojan Nuclear Plant." The second document (Enclosure 2) is a memorandum from J. Hopenfeld to G. Burdick dated December 9, 1992, entitled, " Reply to your request for comments on Draft 'RES Position on Steam Generator Tube Integrity' by L. C. Shao."
Both of these documents have been placed on the docket file for Trojan Nuclear Plant. As such, both documents have been placed in the Public Document Room.
Sincerely,
{
lawrence E. Kokajko, Senior Project Manager Project Directorate V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
As stated
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umTED STATES ENCLOSUREl
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'MEHORANDUM FOR:
Eric S. Beckjord, Director Office of Nuclear Regulatory Research FROM:
Lawrence C. Shao, Director Division of Engineering Office of Nuclear Regulatory Research
SUBJECT:
INTERIM PLUGGING CRITERIA FOR TROJAN NUCLEAR PLANT The Division of Engineering has provided a discussion of the key technical aspects of the rationale used to support steam generator tube int _erim plugging criteria (IPC) for the Trojan nuclear plant and to provide independent conclusions on the viability of IPC for one fuel cycle. The IPC apply'only to the specific case of outer diameter stress corrosion cracking (ODSCC) and intergranular attack (IGA) at tube support plate (TSP) intersections in the steam ger.erators.
The technical rationale presented in the enclosure are based on data and analyses available from NRC research, Trojan plant operating experience, and the technical literature.
The _ enclosure also reflects staff technical experience and opinions. - The Office of Nuclear Reactor Regulation (NRR) has been consulted on technical details regarding IPC during the preparation of this document.
The report endeavors to maintain a distinction between staff opinion and published data.
Based an the discussion presented in the enclosure, the Division of Enginearing concludes that continued operation of the Trojan plant for one fuel cycle is justified.
This justification is based on:
(1)
Examination of steam generator tubes removed from service at the Trojan plant which has revealed cracks that are generally confined to the tube support-plate intersections.
(2)
Burst test results from cracked tubes removed from service at the Trojan plant-which showed burst pressures.well in. excess of main steam line break (MSLB) pressure.
(3)
Stress corrosion crack growth rate results which indicate-that incremental growth of the cracks to a critical length beyond the tube support plate during one fuel-cycle is unlikely, fWf
Eric S. Beckjord 2
DEC 0 s ;s;2 (4)
-The probability of a main steam _line break, the key initiating event for a steam generator tube rupture is very low for one fuel cycle.
Hw Lawre ce C. Shao, Director Division of Engineering Office of Nuclear Regulatory Reser.rch
Enclosure:
As stated cc:
J. Taylor J. Sniezek T. Speis J. Heltemes W. Minners T. Murley F. Miraglia W. Russell J. Richardson J. Strosnider L. Kokajko J, Fouchard e
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ENCLOSURE I
Discussion of Technical Rationale for Steam Generator Tube Interim Plugging Criteria (IPC) at The Trojan Nuclear Plant The purpose of this report is to provide a discussion of the key technical aspects of the rationale used to support steam ger.erator tube interim plugging criteria (IPC) for the Trojan nuclear plant and to provide independent conclusions on the viability of IPC for one fuel cycle. The IPC apply only to the specific case of outer diameter stress corrosion cracking (ODSCC) and intergranular attack (IGA) at tube support plate (TSP) intersections in the steam generators.
The technical rationale presented in this report are based on data and. analyses available from NRC research, Trojan plant operating experience, and the technical literature.
The report also reflects staff technical experience and opinions. The Office of Nuclear Reactor Regulation (NRR) has been consulted on technical details regarding IPC during the preparation of this document. The report endeavors to maintain a distinction between staff opinion and published data.
The rationale presented in this report are based on technical considerations which we believe are adequate to justify IPC for one fuel cycle. Subsequent operation with IPC would require additional review after completion of one cycle and would require consideration of additional information developed at that time.
Longer term technical considerations, such as reliability and sensitivity of NDE techniques for steam generator tube inspection, are the subjects of on-going and new NRC research which is being coordinated with NRR as part of an overall steam generator tube alternate plugging criteria (APC) action plan.
(1)
Background:
Steam generator tube structural integrity guidance provided in Regulatory Guide 1.121 has generally translated into a 40% through-wall
" plugging limit" for flaws in steam generator tubes as part of the plant technical specifications.
However, evidence from pulled steam generator tubes at several plants has revealed numerous short cracks at TSP intersections which are greater than 40% through-wall and yet can withstand pressures in excess of three times operating as required by Regulatory Guide 1.121.
It has therefore been argued by the industry that the 40% plugging limit is conservative, at least for the case of short axial ODSCC/lGA confined to TSP intersections.
Burst testing of crackti tubes removed from service at the Trojan plant has resulted in burst pressures of at least a factor of two in excess of main steam line break (MSLB) pressure, even for through-wall cracks.' NRC research results on tubes with machined and chemically-induced flaws support the contention that the tubes retain significant structural integrity even for up to through-wall cracks, provided that the cracks are short.
From this research "short" can be defined as less than 0.5 inches, which is the length of a near through-wall crack needed to burst for 7/8-inch diameter, 0.050 inch wall thickness tubing under MSLB differential pressure' (see Figure 1).
The burst pressure is defined as the pressure required to penetrate the tube wall. Tube burst then, can result in either small or large leakage.
Tube burst results when the differential pressure acts from the primary side.
Tube rupture relates to a significant opening under burst 1
4 pressure with a consequent increase in leak rate and potential ductile crack advance. Burst failure is differentiated from collapse failure where the differential pressure acts from the secondary side.
Based on the arguments presented previously and supporting analyses, the industry has proposed an alternative to the traditional 40% depth-based guidance, the so-called alternate plugging criteria (APC), for steam generator tubes.
The APC are based on correlations between the voltage amplitude recorded during eddy current tube inspections with a bobbin coil and subsequent measurements of the tube burst pressures and leak rates.
The APC are also currently restricted to 005CC at TSP intersections. A modified version of steam generator tube APC has been accepted by NRR for several licensees. These criteria have been termed interim plugging criteria (IPC).
(2) Trojan Service Experience: Examination of steam generator tubes removed from service at the Trojan plant has revealed cracks which were generally confined to the TSP interrections.
Evidence from the Trojan pulled tube examinations
Subsequent evaluation has revealed that 2 of the 21 ISP intersections examined had cracks which extended beyond the TSP thickness; these cracks extended 0.025 and 0.110 inches beyond the TSP.
(3) Steam Generator Tube Burst Test Results:
Burst test results on tubes removed from the Trojan steam operating or MSLB pressures.' generators showed no leakage under normal When pressurized to failure, the burst pressures mosured for the tubes were in excess of the MSLB pressure by at least a factor of two.
NRC research results from burst tests of tubes with machined, chemically-induced and service-produced defects have also provided a significant body of data on tube integrity.
Equations which have been fitted to the burst test data for electric discharge machined (EDM) slots in steam generator tubes have shown that 0.5 inches would be the length of through-wall crack that would be expected to burst at MSLB pressures for 7/8-inch diameter, 0.050 inch wall thickness tubing' (see Figure 1).
The equation for the through-wall EDM slot represents an extrapolation from data measured on up to 90% through-wall slots. NRC research has shown that the empirical equation developed from EDM slots provides a realistic estimate of remaining margin to f ailure for tubes with stress corrosion cracks.' An empirical equation fitted to data from burst tests of uniformly thinned steam generator tubes has also been developed.' This equation is contrasted with the EDM equation in Figure 1.
It can be seen that the two equations are of similar form but that the uniform thinning equation provides more conservative estimates of tube burst pressures for flaw depths greater than a/t of 0.E. where a flaw depth and t tube wall thickness. Use of either equation to bound degradation up to a/t of 0.8 should yield similar results in terms of burst pressure. However, the EDM equation provides a more accurate representation of stress corrosion cracking and should be used for flaw depths greater than a/t - 0.8.
(4) Stress Corrosion Crack Growth Rate: Growth of the ODSCC tube cracks at the i
Trojan TSP intersections is not expected to be significant during one fuel cycle.
For purposes of this report, significant can be defined as a through-t l
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wall crack extending on the order of 0.5 inches beyond the TSP intersection.
As described previously in (2), the Trojan cracks were generally confined to the TSP thickness; hence, growth beyond the TSP on the order of 0.5 inches would be required for these cracks to be considered critical from a MSLB pressure standpoint.
Upper bound laboratory 005CC growth rate data' indicate that crack growth of this magnitude would not be expected to occur during one fuel cycle. While a through-thickness, full TSP length crack would be expected to fail at MSLB pressure, the opening or rupture would be constrained by the tube support plate. True rupture for the portion outside of the TSP would be expected to occur at MSLB pressure only if the crack had grown on the order of 0.5 inches beyond the TSP intersection.
Further, little or no movementoftheTSPwhichcouldpotentially" uncover"thecracksis-predicted to occur for the MSLB condition.
(5) Probability of Main Steam Line Break: The probability of a MSLB, the key initiating event for a steam generator tube rupture, is very low.
The MSLB would cause approximately a 2600 psi pressure. differential across the steam generator tubes. A MSLB has never occurred in a U.S. nuclear plant.
Quoting from reference 5, "Under the Evaluation and Ifrorovement of NDE for Inservice Inspection of Licht Water Rejttto% Proaram sponsored by the NRC, a team of experts estimated the median frequency of a MSLB to be 1.7 x 10" per reactor year for a volume of 50 gMlons per minute.
This extrapolates to a frequency es+imate of 6.8 x 10" per reactor year for a four loop plant such as Trojan.
(6) Sumary and
Conclusions:
Based on a review of Trojan steam generator tube operating experience, on destructive examinations of tubes removed from the Trojan plant, stress corrosion crack growth rates and expert opinion concerning MSLB frequen:y, it is concludeo that operation of the Trojan plant with steam generator tube IPC for one fuel cycle does not constitute a significant threat to public health and safety.
Subsequent operation with IPC would require additional review after completion of one cycle and would include consideration of information developed at that time.
In summary, the above conclusion is based on:
(1)
Examination of steam generator tubes removed from service at the Trojan plant which has revealed cracks that are generally confined to the tube support plate intersections.
(2)
Burst test results frnm cracked tubes removed from service at the Trojan plant which showed burst pressures well in excess of main steam line break (MSLB) pressure.
(3)
Stress corrosion crack growth rate results indicate that incremental growth of the cracks to a critical length beyond the tube support plate during one fuel cycle is unlikely.
(4)
The probability of a main steam line break, the key initiating event for a steam generator tube rupture is very low for one fuel i
cycle.
3 1
i References 1-
' Trojan. Nuclear Plant Steam Generator Tube Repair Criteria for
-Indications at Tube. Support Plates,
- Westinghouse Energy Systems, WCAP-13129, Revision 1, December,1991, WESTINGHOUSE PROPRIETARY. CLASS 2.
2-NUREG/CR-0718, Steam Generator Tube _ Integrity Program, Phase-I Repart, USNRC,. September, 1979.
3-NUREG CR/2336, Steam Generator Tube Integrity Program Phase !!_ Final-Report, USNRC, August,-1988.
4-HUREG CR/5117, Steam Generator Tube Integrity / Steam Ge'nerator Group Project, Final Project Summary Report, USNRC.-May,1989.
5-Memorandum, C.J. Heltemes to F.P. Gillespie, GI-163, Multiple Steam Generator Tube Leakage, September 28, 1992.
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ENCLOSURE 2 QIC 9 1997 NOTE TO: G. Burdick FROM :
J. Hopenfeld
SUBJECT:
Reply to your request for comments on Draft 'RES POSIDON ON STEAM GENERATOR TUBE INTEGRITY' by LC. Shao Based on certain data, discussed in items 1-8, the document concludes that *it is reasonable to continue operation for one fuel cycle with flaws greater than 40% through-wall at TSP intersections.* The document further suggests that 'subsee: lent operation will require additional review after completion of one cycle and willinclude consideration of information developed at that time'.
GENERAL COMMENT
The information provided in ftems 1-8, of the subject document does not address the main issue conceming steam generator tube integrity which arose from recent cg irating experience with ODSCC, The issue is as follows:
Is it safe to operate plants where an accident such as steam or feed line break may open existing but previously undetected cracks, which will result in a significant primary-to-secondary leakage. Whether the leakage is significant or not would depend on whether the operator can stop the leak before the RWST is depleted.
Degraded tubes also may cause a significant increase on risk from severe acddents.
The fact that cracks within the TSP can withstand the MSLB pressure and that their length will not become critical during one fuel cyde is not an indication that they also will not leak. The Trojan burst test resutts show that three out of the 21 test specimen developed leaks at pressures, of 3300 psi, 7500 psi, and 5500 psi,. with an average depth of penetrations of 38%,58% and 72% respectively.
Item 9) points out that the ebnve specimens "have shown no leakage under normal operating or MSLB pressure conditions'. IT FAILS TO POINT OUT, HOWEVER, THAT THERE IS NO DATA WHICH WOULD ALLOW ONE TO RELATE THE ABOVE LEAKAGE WITH THE OBSERVED DEGREE OF DEGRADATION In other words, if these specimen had undergone a more severe wall penetration would these specimens have leaked at 2600 psi.7. Considering that the 21 specimeh represent a sample of a population of 13,000, the conclusions in p) above are questionable.
The document ignores two other tubes which were pulled out of two US plants and developed leaks at SLB pressures. The leakage was at least an order of magnitude higher than under normal detta ps'. A third tube from a Belgian plant indicates a factor of eight-increase in leakage under SLB conditions, (see Mar. 23 memo). Theoretical
y considerations also indicate a factor of 1000 increase in leakage under SLB conditions, in conclusion, the absence of a deterministic and empirical models for these newly observed cracks precludes the conclusions reached in the subject document. The claim that the conclusions in item 9 are supported by items 1-8 could be considered valid only if one ignores the available data wtiich indicate that higher than normal leakage will occour at SLB pressures even if the tubes do not rupture.
Finally the justifications for any plant operation should not be based on staff opinions or published data on SCC.
SCC is a semi empirical art, in the absence of applicable database other routes of approaching the problem should be considered.
The justification for operating wi+h oracked tubes should be based on what procedures would the operator follow given certain primary to secondary leak and a MSLB between the containment and the MSIV.
These justifications should clearly demene'mte compliance with 10CFR100.
I beleive that the staff can more property judge operator action then predict localized corrosion behavior.
SPECIFIC COM 1ENTS:
Item 1.
The EDM initiated grooves studies provide some measure of the ability of the tube to resist rupture given certain known wall impe.iections. It bears little relation to how ODSCC form, propagate and leak in stcam generator environments.
Item 4 This definition of "significant* is questionable. It makes no difference whether the cracks extend beyond the TSP if they leak at the gap. -lt appears that operators rely on such leakage because they lowered the leakage requirements during normal operations.
Unless one can show that the TSP will cause cracks to plug and they will remain plugged under MSLB pressures the above definition may lead to confusion.
The statement that " upper bound laboratory ODSCC....
.d would not be expected to occur during one fuel cycle
- is not supported by data. The document should compare and present plant and laboratory data with regard to stress intensities and environments before making such claims.
Item 5 The high frequency quoted,6.8x 10 4/RY contradicts the statement that
- it is reasonable *
, item 9, because this frequency would result in a core melt probability of 6.8x 10-3/RY with containment bypass as discussed in the March 27 Memo. The above number is considerably higher than present safety goals, The statement that the key initiating event for SGTR is MSLB is incorrect when taken in the context of the entire document. Item 6 contradicts this statement.
Item 7 Although this item is correct, as stated, it presents only part of the data. As already discussed, three tubes leaked at Tro;an. Three tubes from other plants also leaked at MSLB pressures. Rudimentary consideration dictate that leakage increasos when delta p across the wall is increased.
Item 8 The lengthy discussion of uniform thiraing only confuses the main issue. There are several ways that the reduction load bearing capabilities of a component due to corrosion can be accounted for there is nothing special about these equations. The ASME code takes this into account. The main problem here is LOCALIZED corrosion with an UNK.NOWN ATTACK RATE.
Item 9 A discussion should be ad6ed of the type of new information which is required for the
" additional review' to justify subsequent operations.
ATTACHMENT 1 Second item : Dr. Instead of Mr. or just Hopenfeld The fouowing is missing:
On Sept 11,1902 J. Hopenfeld filed an addendum to the March 27,1992 concluding that " strong coupUng exists between hot leg mass flow,- SG tube leakage and crack l
l l
a On Sept 11,1992 - J. Hopenfeld filed an addendum to the March 27,- 1992 concluding that ' strong coucling exists between hot leg mass flow, SG tube leakage and crack propagation, if confirmed, such a relation between system behavior and undetected tube defects may cause smallleaks to quickly enlarge and results in a MULTIPLE TUBE i
RUPTURE BEFORE THE RCS IS DEPRESSURIZED BY FAILURE OF THE SURGE LINE.
THE RESULTANT COlJTAINMENT BYPASS WILL INCREASE T E S URCE TERM.*
cc; P. Norlan, Gm. Mazetis, W. Minne,. Beckjor i'
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