ML20127G191
| ML20127G191 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 01/14/1993 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Pollard R UNION OF CONCERNED SCIENTISTS |
| References | |
| REF-GTECI-163, REF-GTECI-NI, TASK-163, TASK-OR NUDOCS 9301210178 | |
| Download: ML20127G191 (6) | |
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January 14, 1993 Mr. Robert D. Pollard Nuclear Safety Engineer Union of Concerned Scientists 1616 P Street, NW, Suite 310 Washington, DC 20036
Dear Mr. Pollard:
I am responding to your letter of December 4,1992, wherein you expressed concern that the staff's responses concerning Generic Issue.(GI) 163,
" Multiple Steam Generator Tube Leakage," had deficiencies in regard to nuclear power plant operation.
In particular you expressed concern about the operation of the steam generators at Trojan Nuclear Plant and at other facilities that had conservative voltage-based repair criteria.
The present letter augments the earlier response to you of December 10, 1992.
In this letter, I address each of the major concerns you raised.
Union of Concerned S ientists (UCS) Statement
... Leakage of-many flawed steam generator tubes could have the same-result as the rupture of-fewer tubes. The. reason is that flawed steam -
generator tubes which are not leaking during normal operation, could begin leaking under the higher differential pressure caused by a main steam line break (MSLB). Similarly, the leakage through. flawed-tubes during normal operation at a rate of less~ than 130 gpd could increase significantly as a result of the higher differential pressure-of-a MSLB...The fact that degraded. tubes neither leak at normal pressures, nor burst under SLB [ steam line break] pressures-is not an indication that they will not leak following a SLB incident....It makes no difference whether the leak origin was from one ruptured tube-or many pin hole leaks."
Resoonse The cumulative leakage from a large number of tubes leaking at small-individual rates under steam line break pressure conditions:was specifically addressed in the review of the Trojan Interim Plugging Criteria.
Section 2.5.2.4 of the NRC staff's safety evaluation describes the licensee's assessment of the leakage _that could occur as the result of a steam itne break postulated to occur at the end of the operating cycle, and Section 2.8.4.2 presents the-NRC staf_f's evaluation of the issue.
The licensee's assessment of the potential SLB leakage at the endiof cycle 14 was calculated using correlations derived from laboratory. tube-specimens and tube specimens-removed from the Trojan steam generators.
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Mr. Robert D. Pollard,
The calculation was a Monte Carlo analysis based on the distribution of voltage signals at tube support plate (ISP) intersections from the indications left in service at Trojan. The analysis included probability distributions to account for uncertainties in voltage measurement, voltage growth, and leak rate versus voltage. Based on the cycle 14 analysis done by Westinghouse for Trojan, applying the predicted end of cycle distribution of indications left in service, a total leakage of 0.16 gallon per minute (gpm) was calculated.
This leak rate was significantly below the allowable leak rate based on acceptable site boundary dose considerations.
VCS Statement "The staff's response asserts that the Office of Nuclear Regulatory Research made assumptions that ignored the information obtained from tests and inspections of the Trojan steam generator tubes [which were) described in the staff's safety evaluation report (SER) (for Trojan]...."
Response
As indicated in your December 4 letter, the Office of Nuclear Regulatory Research (RES) prioritization analysis did reference the Trojan-specific evaluation and data.
Nonetheless, the RES analysis did not reconcile the differences between their assumptions and the experimental data that were provided by the licensee.
UCS Statement
....there is a low probability that flaws will be detected and that even if detected, it is difficult to determine the length and depth of the cracks."
Response
Although the probability of detecting short cracks at the support plates may be relatively low under certain circumstances, on the basis of considerable field evidence, cracks will be detectable by eddy current inspection before they become sufficiently large (in terms of both length and depth) to potentially impair tube integrity.
On the basis of our review of the inspection program implemented at Trojan, we believe there is assurance that all significant cracks that could have potentially caused tube rupture resulting in large leak rates were plugged or repaired.
VCS Statement
"....There also appears to be insufficient data to be confident that the estimates of crack growth during operation are conservative or that leak rate monitoring during operation can provide an adequate basis for
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Mr. Robert D. Pollard,
evaluating crack growth during operation and during accidents....there is insufficient data to make a compelling case that an adequate safety margin remains."
IbdPRER The development of the interim plugging criteria for Trojan is based on estimates of the change in voltage response during cycle 14 for indications left in service at the beginning of cycle 14, in lieu of being based directly on estimates of crack growth in terms of crack length and/or crack depth.
The estimated changes in voltage response during cycle 14 are based on observed voltage changes for 806 indications during cycle 13. The staff believes the extensive cycle 13 data provides an adequate basis for estimating voltage changes to be expected during cycle 14, particularly in view of actions taken by the licensee to reduce the molar ratio imbalance in the tube support plate crevices. The value of the interim plugging limit at Trojan has been set to provide assurance that projected end-of-cycle 14 voltages will be limited to values ensuring adequate structural and leakage integrity, based on empirical relationships for burst pressure and SLB induced leakage as a function of voltage amplitude response.
These empirical relationships are based on data from both tube specimens from the field (including Trojan) and laboratory tube specimens, as described in the staff's SER.
Steam generator tube structural and leakage integrity during cycle 14 has been assured through the extensive tube inspections and repairs performed prior to cycle 14 operation.
For the reasons cited above, the interim plugging limits at Trojan are believed to make conservative allowance for potential crack growth during cycle 14. These interim limits have been developed independently of the leak rate limits and monitoring procedures in place at Trojan.
These leak rate limits and monitoring procedures are intended to provide added assurance of tube structural and leakage integrity during cycle 14.
In the unexpected event that the stress corrosion cracking at the tube support plates should lead to significant operational leakage, the leak rate limits and monitoring procedures ensure timely shutdown of the unit for the appropriate diagnostic and corrective actions before rupture can occur.
As stated in my December 10, 1992, letter to you, the NRC has a formal internal review process, including peer and management review of all analyses and opinions, to ensure that all technical and safety issues are identified and resolved before requiring industry action. This is required in the generic issue process.
Some of the issues that have been raised are important relative to the review of the alternate plugging criteria that is being proposed by the industry. The NRC staff will continue to research and assess those issues prior to making a final decision on the industry proposals.
However, the NRC staff believes the interim criteria that have been implemented at Trojan, and other plants, are conservative and provide adequate assurance of the public health and safety.
1 Mr. Robert D. Pollard -
Finally. with regard to the question of whether there is a fundamental disagreement between two offices within the NRC, I am enclosing for your information a memorandum describing the Office of Nuclear Regulatory Research's position on interim plugging criteria for the Trojan Nuclear Plant.
Sincerely, w at [
"t Thomas E. Murley, Director Office of Nuclear Reactor Regulation
Enclosure:
Memorandum dated January 5, 1993 (Beckjord to Murley) d 1
Mr. Robert D. Pollard January 14, 1993 Finally, with regard to the question of whether there is a fundamental disagreement between two offices within the NRC, I am enclosing for your information a memorandum describing the Office of Nuclear Regulatory Research's position on interim plugging criteria for the Trojan Nuclear Plant.
Sincerely, Originni signed by 7nocus E. Marl 07 Thomas E Murley, Director Office of Nuclear Reactor Regulation
Enclosure:
Memorandum dated January 5,1993 (Beckjord to Murley)
DISTRIBUTION:
See next page See previous concurrence OFC LA/PDV*
PM/PDV D/PDVk TECH ED*
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D/DE*
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NAME DFoster
,LKokajko TQuay RSanders JStrosnider JRichardson DATE 01/0 /93
/ /f /93
/ /4/93 12/16/92 12/17/92 12/18/92 0FC A045/DRPW*
D/DRPW*
ADT/NRR*
ADP/NRR*
DD/NRR*
Q/t)ILR_
NAME MVirgilio JRoe WRussell JPartlow FHiraglia IMurley
/ / [793
/
DATE O1/07/93 01]08]93 01/11/93 01/12/93 01/12/93 0FFICIAL RECORD COPY DOCUMENT NAME: GT8325.II
e-January 14, 1993
.t DISTRIBUTION:
Docket File (50-344 w/ incoming)
NRC & Local PDRs (w/ incoming)
JTaylor, 17G21 JSniezek, 17G21 l
JBlaha, 17G21 HThompson, 17G21 TMurley/FMiraglia,12G18 JPartlow,12G18 DCrutchfield,11H21 FGillespie, 12G18 EBeckjord, NLS007 PDV Reading File (w/ incoming)
JRoe MVirgilio E00 #0008325, 17G21 TQuay l
KKnubel,17G21 DFoster LKokajko OGC (15B18)
ACRS (10)(P315)
PDV Action File (w/ incoming)
KPerkins, RV KKarwoski, 7D4 HConrad, 704 GJohnson, 704 JStrosnider, 704 JHeltemes, NLS007 GBurdick, NLS314 j
-RBarr, RV i
JRichardson, 7026 SRichards, RV PJohnson, RV-SHoffman,-14B20 EAdensam, 14B20 NRR Mailroom*8325 (w/ incoming), 12G18 TGibbons CHawes BClayton, 12G18 SEC) 92-0980,-16G15 1900M
ENCLOSURE ucu UNITED STATES n
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JAN 0 51993 MEMORANDUM FOR:
Thomas E. Murley, Director Office of Nuclear Reactor Regulation FROM:
Eric S. Beckjord, Director Office of Nuclear Regulatory Research
SUBJECT:
INTERIM PLUGGING CRITERIA FOR TROJAN NUCLEAR PLANT The purpose of this letter is to state the position of this Office on the matter of the interim plugging criteria (IPC) for the Trojan nuclear plant.
In Enclosure (1), the Division of Engineering (DE), RES, provided a discussion of the key technical aspects of the rationale used to support steam generator tube interim plugging criteria (IPC) and reached the conclusion that operation of Trojan for the remainder of cycle 14 (approximately 4 months) is justified.
I concur in this position.
Calculations of leak rate across the steam generator tubes for a main steam line break (MSLB), and of the core damage frequency for MSLB support this conclusion. These calculations are based on the available information on steam generator tube inspection prior to December 1992, We do not have the results of PGE eddy current tests of a number of tubes taken recently.
The best estimate of the leak rate expected following a MSLB at the end of cycle 14, for a differential pressure of 2600 psi across the tubes is 145 gallons per minute (gpm). The best estimate of core damage frequency for the sequence initiated by MSLB or a stuck open safety valve is 1.4 E-6 per I will provide descriptions of these calculations as soon as we have
- year, completed the uncertainty analysis for them.
I note also that Trojan procedures for MSLB call for stopping high head safety injection pumps at pressure differences substantially below 2600 psi. We will complete an analysis of this in several days. The point of this analysis is that the steam generator tube differential pressure would be less than the value assumed, and accordingly the steam generator tube leakage would be less than 145 gpm.
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JAN 0 51993 In Enclosure (2), J. Hopenfeld, Division of Safety Issue Resolution (DSIR),-
RES, provided several comments on a draft of Enclosure.(1). Enclosure (3) provides DE responses to those comments that apply to final DE memorandum.
Eric S. Beckjord, )irector Office of Nuclea Regulatory Research
Enclosures:
(1)
Memorandum, L.C. Shao to E.S. Beckjord, INTERIM PLUGGING CRITERIA FOR TROJAN NUCLEAR PLANT, December 9, 1992.
(2)
Note to G. Burdick from J. Hopenfeld, COMMENTS ON DRAFT RES POSITION ON STEAM GENERATOR TUBE INTEGRITY, December 9, 1992.
(3)
Division of Engineering Responses to Comments of J. Hopenfeld cc:
J. Taylor J. Sniezek J. Roe W. Minners F. Miraglia W. Russell J. Richardson J. Strosnider L. Kokajko J. Fouchard
ENCLOSURE 1 DEC 0 91992 MEMORANDUM 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 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) Ind intergranular attack (IGA) at tube support plate (TSP) intersections in the steam generators.
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 on the discussion presented in the enclosure, the Division of Engineering 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, i
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(4)
The proba*cility of a main steam line break, the key initiating event for a steam generator tube rupture is very low for one fuel CyCie.
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Lawrence C. Shao, Director Division of Engineering Office of Nuclear Regulatory Research j
[nclosure:
As stated cc:
J. Taylor J. Sniezek T. Spets J. Heltemes W. Minners T. Murley F. Miraglia W. Russell J. Richardson
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1 ENCLO5tlRE l
Discussion of Technical Rationale for Steam Gene /ator Tube Interim Plugging Criteria (IPC) at The Trojan huclear Plant The purpose of this report is to provide a discussion of the key technical aspects of the rationale used to support steam generator 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 (lGA) 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 reflWs staff technical experience and opinions. The Office of Huclear Reattor 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.
lhe rationale presented in this report are based on technical consideritions 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 ci 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 NRt research which is being coordinated with NRR as part of an overall tteam 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 nenerator tubes at several plants has revealed numerous short cracks at TSP intersections which are greater than 40% through-wall and yet can withstand prestures 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 leest for the case of short axial ODSCC/lGA confined to TSP intersections.
Burst testing of cracked 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.8 NRC research results on tubes with machined and chemically induced flaws support the contention that the tubes retain significant structural _
3 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 HSLB 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 hy 1
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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 prewnted 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
- record ~i during eddy current tube inspections with a bobbin coil and subsequ.nt measurements of the tube burst pressures and leak rates. The APC are also currently restricted to ODSCC at TSP intersections. A modified version of steam generator tube APC has been accepted by HRR for several licensees.
These criteria have been termed interim plugging criteria (IPC).
(2) Trojan Service Experiences Examination of steam generator tubes removed from service at the Trojan plant has revealed cracks which were generally confined to the 15P intersections.
Evidence from the Trojan pulled tube examinations' has shown that the outer diameter (00) lengths of the cracks ranged almost up to the 15P thickness.
Subsequent evaluation has revetled that 2 of the 21 TSP intersections examined had cracks which extended beyond the ISP thickness; these cracks extended 0.025 and 0.110 inches beyond the ISP.
(3) Steam Generator Tube Burst Test Results: Burst test results on tubes removed from the Trojan steam generators showed no leakage under normal operating or MSLB pressures.' When pressurized to failure, the burst pressures measured 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 HSLB 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 failure for tubes with stress corrosion cracks.' An empirical equation fitted to data from burst tests of unitarmly thinned steam generator tubes has also been developed.' This equation is contrasted with the EDM equation in rigure 1.
It can be seen that the two equations are of similar form but that 'he uniform thinning equation provides more conservative estimates of tube burst pressures for flaw depths greater than a/t of 0.8, where a - flaw depth and t tube wall thickness. Use of either equation to bound degradation up to a/t of 0.8 should yleid similar results in terms of burst pressure. However, the EDM equation provides a more accurate representation of stress corrosion cracking and should be used ft,r flaw depths greater than a/t - 0.8.
(4) Stress Corrosion Crack Growth Rate: Growth of the 00500 tube cracks at the 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-2
wall crack extending on the order of 0.5 inches beyond the ISP intersection.
As described previously in (2), the Trojan cracks were generally confined to the ISP 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 00500 grow *h rate data' indicate that ctack growth of this reagnitude would not be expected to occur during one fuel cycle. While a tF-thukness, full TSP length crack would be expected to fail at HSt t sure, the opening or rupture would be constrained by the tube support pla True rupture for the portion outside of the TSP would be expected to oci.ur at HSLB pressure only if the crack had grown on the order of 0.5 inches beyond the TSP intersection, further, little or no movementoftheTSPwhichcouldpotentially* uncover"thecracksispredicted to occur for the HSLB condition.
(5) Probability of Main Steam Line Break: The probability of a HSLB, the key initiating event for a steam generator tube rupture, is very low.
The HSLB would cause approximately a 2600 psi pressure differential across the steam generator tubes.
A HSLB has never occurred in a U.S. nuclear plant.
Quoting from reference 5, "Under the Evaluation and Irorovement of NDE for Inservin Inspection of licht Water Reactors Proaram sponsored by the NRC, a teart of experts estimated the median frequency of a HSLB to be 1.7 x 10" per reactor year for a volume of 50 gallons per minute. This extraslates to a frequency estimate 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 frequency, 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 af ter completion of one cycle and would include consideration of infomation 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 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 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 steem line break, the key initiating event for a steam generator tube rupture is very low for one fuel cycle.
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References
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' Trojan Nuclear Plant Steam Generator Tube Repair Criteria for Indications at Tube Support Plates,
- Westinghouse Energy Systems WCAP-13129. Revision 1. December, 1991, WESTINGHOUSE PROPRICTARY CLAS$2.
2-NUREG/CR-0718, Steam Generator Tube Integrity Program, Phase 1 Report, USNRC, September, 1979.
3-NUREG CR/2336, Steam Generator Tube Integrity Program, Phase 11 Final Report, USNRC, August, 1988.
4-HUREG CR/Sil7, Steam Generator Tube Integrity / Steam Generator Group Project, Final Project Sumary Report, U$NRC. Hay,1989.
5-Memorandum, C.J. Heltemes to F.P. Gillespie, GI-163, Multiple Steam Generator Tube Leakage, September 28, 1992.
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NOTE TO: G. Burdick FROM :
J. Hopenfeld
SUBJECT:
Reply to your request for comments'on Draft *RES POSITION ON STEAM GENERATOR TUBE INTEGRTTY by LC. Shao Based on certain data, dismssed in items 10, 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 dooumont further suggests that ' subsequent operation will require additional review after completion of one cyde and willinclude consideration of information developed at that time'.
GENERAL COMMENT
The information provided in items 18, of the subject document does not address the main issue conceming steam generator tube integrity which arose from recent operating 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 accidents.
The fact that cracks within the TSP can withstand the MSLB pressure and that their length will not become critical during one fuel cycle is not an indication that they also will not leak. The Trojan burst test results show that three out of the 21 test specimen developed leaks at pressures, of 3300 psi, 7500 psi, and SS00 psi,. with an average depth of penetrations of 38%,58% and 72% respectively.
Item (7) points out that the above 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.?. Considering that the 21 specimeh represent a sample of a population of 13,000, the conclusions in (7) 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 N!YWW'
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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 14 could be considered valid only if one ignores the available data which indicate that higher than normal leakage will occcur 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.
SCO is a semi empirical art, in the absence of applicable database other routes of approaching the problem should be considered.
The justification for operating with cracked 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 demonstrate compliance with 10CFR100.
I beleive that the staff can more properly judge operator action than predict localized corrosion behavior.
SPECIFIC COMMENTS:
flem 1.
The EDM initiated grooves studies provide some measure of the ability of the tube to resist rupture given certain known wall imperfections. It bears little relation to how ODSCC form, propagate and leak in steam 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. !t 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....
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, G.8x 104/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 fdomo. The above number is 3
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 Ahhough this item is correct, as stated, it presents only part of the data. As already discussed, three tubes leaked at Trojan. Three tubes from other plants also leaked at MSLB pressures. Rudimentary consideration dictate that leakage increases when delta p across the wallis increased.
Item 8 The lengthy discussion of uniform thinning 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 UNKNOWN ATTACK RATE.
Item 9 A discussion should be added of the type of new information which is required for the
' additional review" to justify subsequent operations.
ATTACHMENT 1 i:
Second item : Dr. Instead of Mr. _ or just Hopenfeld The following is missing:
On Sept 11,1992 J. Hopenfeld filed an addendum to the March 27,1992 concluding that
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t On Sept 11,1992 J. Hopenfeld filed an addendum to the March 27,1992 concluding that ' strong coupling 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 RUPTURE BEFORE THE RCS IS DEPRESSURIZED BY FAILURE OF THE SURGE LINE.
THE RESULTANT CONTAINMENT BYPASS WILL INCREASE T1 E S URCE TERM.*
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J.H P. Norlan, Gm. Mazetis, W. Minner,. Beckjor 1
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4 ENCLOSURE 3 DIVISION Of ENGINEERING RESPONSES TO COMMENTS OF J. HOPENFELD The Division of engineering has compiled responses to the coments of J.
Hopenfeld on steam generator tube interim plugging criteria (IPC).
The responses are restricted to the Division's areas of expertise and refer to the note from J. Hopenfeld to G. Burdick of December 9, 1992 (attached).
The comments by J. Hopenfeld, in turn, refer to a previous draft of the Division of Engineering memorandum which was issued in final form on December 9,1992.
(1) COMMENT: 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 4
significant or not would depend on whether the operator can stop the leak before the refueling water storage tank (RWST) is depleted.
Degraded tubes may also cause a significant increase on risk from severe accidents.
RESPONSE
Plants operating with the interim plugging criteria (IPC) will have an increased probability of leakage under MSLB or FLB. As a result these plants have more restrictive limits on leak rate, and operators have received additional training in the handling of steam generator leakage problems. The Division of Engineering has estimated the overall leakage under Main Steam Line Break (MSLB) to be 145 gallons per minute (gpm) for the Trojan plant.
With consideration of a 428,000 gallon RWST, this provides the operators with significant time in which to respond to steam generator tube leakage or rupture.
(2) COMMENT: The fact that cracks within the TSP can withstand the HSLB pressure and that their length will not bt;ome critical during one fuel cycle is not an indication that they will not leak.
RESPONSE
As stated previously, plants operating with IPC will have an increased probability of leakage under MSLB or FLB. The Division of Engineering has. attempted to estimate this leakage as described in the response to comment (1).
(3) COMMENT:
... There is no data which would allow one to relate the above leakage with the observed degree of degradation.
RESPONSE
A leak rate model (reference 1) has been used to relate crack size with the amount of leakage under normal operating conditions and MSLB. The leak rate model has been used to determine the estimate presented in the response to comment (1). The model has been benchmarked to a previous EPRI leak rate model (reference 2) and to leak rate data-f rom both U.S. and foreign sou rc e s.-
The leak rate data were for cracks in laboratory-degraded tubes that were intended to simulate tubes with actual service-induced cracks.
While the leak rate data exhibit significant scatter, the overall trends of leak rate versus crack length were consistent with the leak rates predicted by the model.
(4) COMMENT:
The document ignores two other tubes which were pulled out of two U.S. plants and developed leaks under SLB pressures.
The leakage was at least an order of magnitude higher than that developed under normal delta p's.
A third tube from a Belgian plant indicates a factor of eight increase in leakage under SLB conditions (see Har. 23 memo).
Theoretical considerations also indicate a factor of 1000 increase in leakage under SLB conditions.
RESPONSE
The Division of Engineering memorandum considered Trojan plant-specific data (reference 3).
Pulled tube data from other U.S. plants, also documented in reference 3, showed four cases where leakage rates increased by a range of from 7X to 2,040X in going from normal operating differential pressure to MSLB.
However, in all of these cases the leakage rate went from completely insignificant to a barely measurable leakage, e.g 0.0041/hr (0.000017 gpm) to 8.161/hr (0.035 gpm), and not to a large leak or rupture.
The leak rate data and modeling from reference 1 indicate an expected increase of 10 times in going from normal operating differential pressure to MSLB for a 0.7 inch long flaw.
Trojan pulled tube burst tests revealed no leakage under normal operating or MSLB conditions. A deterministic model for leakage from an axial crack in a pressurized tube, described in reference 4, indicates a possibility of up to a factor of 1000 increase in leak rate in going from normal operating conditions to MSLB. However, to obtain such a large increase, the model assumes tight (10 m wide at the center) cracks which would have completely insignificant leakage under normal operating conditions.
Belgian test data provided in reference 3, show approximately a factor of 2 increase in leakage for MSLB versus normal operating conditions. Our conclusions from the examination of this data indicate that while a factor of 1000 increase in leak rate is possible in going from normal operating differential pressure to MSLB, such an increase is not likely.
Further, on an absolute basis, the resulting amount of leakage for a given tube would likely be insignificant.
(5) COMMENT:
In conclusion, the absence of 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 (draft memorandum) are supported by items 1-B (draft memorandum) could be considered valid only if one ignores the available data which indicate that higher than normal leakage will occur at SLB pressures even if the tubes do not rupture.
RESPONSE
The conclusions in the memorandum were reached primarily on the basis of the limited incremental crack growth expected for one fuel cycle and the low probability of MSLB. The emphasis in the memorandum was also on steam generator tube rupture as opposed to leakage.
Subsequent work by the Division of Engineering has provided a best estimate of tube leakage under MSLB based on a probabilistic flaw distribution (using Trojan pulled tube data as the basis) and a leak rate model (reference 1). This analysis indicates a best estimate leak rate of 145 gpm under MSLB. This analysis, coupled with the Trojan pulled tube data and the discussion in the response to comment (4) above, do not support the claim for a drastic increase in leakage under MSLB.
(6) COMMENT:
Finally, the justifications for any plant operation should not be based on staff opinions or published data on SCC.
SCC is an empirical art, in the absence of an applicable database, other routes of approaching the
problem should be considered.
I believe that the staff can more properly judge operator action than predict localized corrosion behavior.
RESPONSE: Justifications for plant operation are commonly based on consensus staff opinion regarding the state-of-the-art in a specific technical area.
SCC growth rate data specific to the Trojan plant steam generator tubing and environment were not available for this analysis. However, SCC growth rate data for mill-annealed Inconel 600 in a NaOH environment (4g/1 concentration) at 350'C (662'f) were available from reference 5.
Utilizing this data, several growth rates were applied to a probabilistic flaw distribution based on the Trojan pulled tube examinations to yield a final leak rate determination. A "best estimate" growth rate of 0.1 m/hr was used which yields total growth over a twelve month cycle that would be conservative base'd on comparisons with Trojan operating experience.
(7) COMMEN1:
Item 1, draft memorandum - The EDM initiated grooves study provides some measure of the ability of the tube to resist rupture given certain known wall imperfections.
It bears little relation to how ODSCC fonn, propagate and leak in steam generator environments.
RESPON5C: The EDM data were utilized to ascertain the length of crack that would be expected to " burst" at various pressures, including MSLB.
Durst is defined as through wall penetration of the tube wall with pressure from the primary side.
No claims were made for EDM slots simulating initiation and propagation of SCC cracks.
Equations fitted to the EDM slotted tubes data were found to predict burst pressures for tubes with stress corrosion cracks of similar lengths and depths (reference (1)).
(8) COMMENT:
Item 4, draft memorandum - The definition of significant is questionabic.
It makes no difference whether the cracks extend beyond the TSP if they leak at the gap.
It 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.
RESPONSE
It does make a significant difference if cracks extend beyond the TSP both from the standpoint of potential tube rupture and increased leakage, if the cracks extend far enough beyond the TSP (on the order of 0.5 inches),
full rupture could occur under MSLB. While a through-thickness, full TSP length flaw would be expected to fail at MSLB pressure, the opening would be constrained by the TSP. This constraint would serve to limit a rupture and any consequent leakage.
(9) COMMENT: Item 4, draft memorandum - The statement that " upper bound laboratory ODSCC... 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.
RESP 0tl5E: The statement referred to above is specifically based on data that was referenced in the memorandum (reference (5)). The discussion in response to comment (6) is applicable here.
(10) COMMEllT: Item 7, draft memorandum - Although this item is correct as stated, it provides only part of the data. As already discussed, three tubes leaked at Trojan. Three tubes from other plants also leaked at MSLB pressures. Rudimentary considerations dictate that leakage increases when delta p across the wall increases.
RESPONSE: The aulled tube data froo Trojan showed that none of the tubes that were tested lea (ed at MSLB pressure. Three of the pulled Trojan tubes eventually leaked beyond MSLB pressures (3300 psi, 5500 psi and 7500 psi) before rupturing (reference 6). The tube that leaked at 3300 psi was also noted to have been damaged during removal.
Pulled tube data from other plants have shown leakage for MSLB pressure as described previously in the response to comment (4).
For a through-wall flaw, leakage should generally increase with increasing differential pressure across the tube wall.
However, as pointed out previously in the response to comment (4), the leakage may be completely insignificant.
(11) COMMEl4T: Item 8, draft memorandum - the lengthy discussion of uniform thinning only confuses the main issue. There are several ways that the reduction in load bearing capabilities of a component due to corrosion can be accounted for (and) there is nothing special about these equations.
The ASME Code takes this into account.
The main problem here is localized corrosion with an unknown attack rate.
RESP 0tiSE:
The discussion of uniform thinning was included at the request of the Division of Safety issue Resolution (DSIR) to clarify differences with the EDM slot equation.
While the growth rate of the localized corrosion (ODSCC and Intergranular Attack (IGA)) is uncertain, it can be reasonably estimated with data availabic from reference 5.
(12) COMMEt4T: A discussion should be added of the type of information which is required for the " additional review" to justify subsequent operations.
RESP 0tiSE:
Information for additional review would include, but not be limited to, eddy current inspection data at the end of the cycle, projected versus f
5 actual crack growth based on destructive analyses of pulled tubes, metallographic examination of pulled tubes, leak and rupture tests of pulled tubes and additional secondary side inspections.-
References:
1-NUREG CR/2336, Steam Generator Tube integrity Program, Phase !! final Report, USNRC, August, 1988.
2-Griesbach, T., R. Cipolla and J. Lang, ' Analysis Methods for Evaluating Leak-Before-Break in U-Tube Steam Generators " Proceedings of the ASME Winter Meeting, Miami Beach, FL, 1985.
3-
" Trojan Nuclear Plant Steam Generator Tube Re) air Criteria for Indications at Tube Support Plates,
- Westingiouse Energy Systems, WCAP-13129, Revision 1, December,1991, WESTINGHOUSE PROPRIETARY CLASS 2.
4 EPRI NP-6864-L. LICENSABLE MATERIAL, 'PWR Steam Generator Tube Repair Limits: Technical Support Document for Expansion Zone PWSCC in Roll Transitions (Revision 1)," Electric Power Research Institute, Palo Alto, CA, December, 1991.
5-NUREG CR/5117, Steam Generator Tube Integrity / Steam Generator Group Project, Final Project Summary Report, USNRC, May,1989.
Magee, T.P, J.F. Hall and R.S. Maurer, " Trojan Steam Generator Tubing 6
Destructive Examination Interim Report Update," TR-MCC-186, ABB Combustion Engineering, Windsor, CT. NOTE: THIS REPORT IS NOT DATED AND MAY BE CONSIDERED PROPRIETARY.
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MULTIPLE STEAM GENERATOR TUBE RUPTURE ACCIDENTS TAYLOR SNIEZEK DATE: 11/23/92 BLAHA KNUBEL
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4 December 4,1992 Ivan Sclin, Chairman James R. Curtiss E. Oafl de Planque Forrest J. Remick Kenneth C Rogers U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
01 103, Multipic Steam Generator Tube 12akage I am writing to express my concerns about the adequacy of the NRC's responses to the requesta mnde in my letter to you dated November 23.1992.
Thank you for the prompt uction in placing the documents I requested in the public document ronm and for advance notification of the agenda for the meeting held at the Trojan plant on December 1,1992. I did not attend because of the location, the short notice and the fact that the ogenda was not focused on the stuffs assessment of the risks posed by multiple steam generator tube leakage.
As far as I am aware, the stnff has not yet provided an explanation of why Westinghouse plants, other than Trojan, are being allowed to enntinue operation with flawed stenm generator tubes before the subject generic issue is resolved.
As for the staffs explnnation, thus far, concerning the Trojan plant (l.c., a November 24,1992 memo to CJ. Heltemes, Jr. from F. Gilles pie and a November 30,1992 memo to F. Gillespie from CJ. Heltemes, Jr.), I consider ; t to be wholly inadequate.
Two major defielencies in the staffs response, which are discussed in more detail below, are:
1)
Addressing only the September 28,1992 memo from C.J. Heltemes, Jr. and generally ignoring the well documented technical issues raised in J. Hopenfeld's memos dated March 27,1992 and september 11,1992 and in J. Muscara's memo dated March 16,1992.
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Falling to crplain why the Trojan plant may be permitted to operate prior to the resolution of a generic safety issue that was opened as o direct result of the Trojan lleense amendment permitting operation with hundreds of flawed steam generatur tubes.
The following technlealissues are among those that the staff has generally ignored in its response.
- An accident involving core melting and radioactive releases outside the containment can occur even without the rupture of any steam generator tubes. leakage of many flawed steam generntor tubes could have the same result as the rupture of fewer tubes.
The reason is that finwed steam generator tubes which are not leaking during normal operation, could begin leaking under the higher differential pressure caused by a main steum line break (MSLD). Similarly, the leakage through flawed tubes during normal operatinn at a rate less thnn 130 gpd could increase signifienntly as the result of the higher differential pressure of a MSLB.
The statements supporting the Differing Professional Opinion (DPO) are more succinct: 'The fact that degraded tubes neither leak, at normal pressures, not burst under SLB pressures is not an indication that they will not lenk following a SLB neeldent. * *
- It makes no difference whether the leak origin was from one ruptured tube or many pin hole leaks." [ Memo from J. Hopenfeld, March 27,1992, Enclnsure,
- p. S.)
' The staffs response asserts that the Office of Nuc: car Regulatory Research made assumpilons that ignored the information obtained from testr, and inspectiom of the Trojan steam generator tubes described in the staffs safety evaluation report (SER) for the Trojan license amendment no.178. Even a cursory examination of the documents cited above would have shown that this assertion is false. The September 28,1992 memo from C Heltemes, Jr. und the March 27,1992 memo from J. Hopenfeld explicitly took into consideration the Trojan SER and the tests and inspections conducted on the Trojan steam generator tubes.
- The staffs response fails to even acknowledge the existence uf, much less address, the large number of uncertainties involved in assessing the risk of plant operation with hundreds or even thousands of cracked tubes. For example, there is a low probability that finws will be detected and that, even if detected,it is difficult to determine the length and depth of the cracks. There also appears to be insufficient data to be confident that the estimates of crack growth during operation are conservative or that leak rate monitoring during operation can provide an adequate basis for evaluating crack growth during operation and during accidents such as MSLH. Although the staff may have attempted to be conservative, there is insufficient data to make a compelling case that an ndequnte safety margin remains.
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' The staffs response claims that the Offlee of Nuclent Regulatory Research erred in stating that a leak.before brenk concept is inherent in the staffs safety evaluation report for Trojan. [ Memo from F. Gillespie, Nmtmber 24,1992, Enclosure, p. 3.]
If that were a correct claim, the stnft would have to acknowledge that the reductic.n in the allowable leak rate to 130 gallons per day provides no improvement in safety over the tube leak rate previously applienble to the Trojan steam generators. This la so because, unten the tube leak rate increases slowly (f.e., unless the leak before break concept is inherent in the staffs safety evaluation re, ort for Trojan), the tube leakage rate will go from negligible to a rate in execu of boti the new 130 gpd limit and the previous higher limit. In fact, this is cuentially what occurred on with the tube leak at Trojan on November 9,1992, although I recognize that the licensee believes the cause was Imptoper heat treatment of a sleeve wcld rather than a Dawed tube. Neverthelen, the same behavior of a rapidly increasing leak rate or tube rupture could occur during a steam line break accident.
As a final example, it should be nnted that the staffs response does not nddress the point that relaxation of the steam generator tube repair criteria provides no safety benefit to the public. The sole benefit is to the economics of continued operation of the Trojan plant because, absent NRC npprnval of the more lenient repnir criteria, the plant probably could not operate at 100 percent of rated power. Furthermore, the approval of the Trojan license amendment violated the much touted principle of defense.in depth. What remains in terms of protection for public may be httle more than a Maginot Une.
With regard the second major deficiency in the staffs response to date, the staffimplies thut the new generic safety issue established ni directly result of the Trnjan license amendment need not be resolved in order to permit operation of the Trojan plant.
This ignores the criticism of NRC's handling of generic safety issues expressed by the President's Commission on the Accident at Three Mlle Island (the Kcmeny Commission) and the U.S. General Accounting Office (OAO).
The staff of the Kemeny Commission found that defining an issue as generic was a mechanism the NRC used to " insure the granting of an operating license for an already constructed plant." [Steff Report to the President's Commission on the Accident at Three Mile Island, The Nuclear Regulatory Commlulon, October 1979, p 43.)
The Kemeny Commission itself pointed to NRCa handling of generic safety issues as "an important exumple" of how "NRC's primary focus is on licensing and insufficient attention has been paid to the ongoing proecss of assuring nuclear safety." The Kemeny Commission concluded that "the evidence indicates that labeling of a problem as ' generic' may provide a convenient way of postponing decision on a difficult question." (Report of the President's Commission on the Accident at Three Mile Island, October 1979, p. 20.)
W IW.8/De TO I3015H 2275 s992es2.M i1,e apg c),pfg More recently, the OAO performed an assessment of NRC's safety standards, enforcement activitics and inspection efforts. [U.S. General Accounting Ofuce, " Efforts to Ensure Nuclear Plant Safety Can Be Strengthened," OAO/RCED47141, August 1987.) The following are some of the issues raised by the OAO:
- The lack of guidelines to identify safety violations severe enough to require a nuclear plant to shut down;
- The slow corrective actions on the part of the NRC to shut down nuclear plants with records of chronic safety violations. It was documented that the NRC has taken from several months to up to ten years to resolve generic safety issues. Including thoso that the Commission believes pose the highest safety risk.
- The backlog of 163 unresolved generic safety issues as oflast December (1986],
including 32 considered to pose a significant risk to public health and safety.
I am inclined to conclude that the same problems are affecting the NRC staffs handling of the issues concerning the Dawed steam generator tubes at the Trojan plant, at least based on the staffs response to date.
I remain open to the possibility that there may be an adequate technien) basis for concluding that operation of the Trojan plant g!ven the current condition of the steam generator tubes is acceptable. However,I do not believe that the staff has yet provided such a basis. Dus, any requests for an adjudicatory hearing before Trojnn resumes operation whfeh have been made by the pubtle are understandable and should be given careful attention.
The problem fucing Commission is that there is a fundumental disagreement between two of your staff of0ces about the risks posed by Dawed steam generator tubes. One of0cc bases its conclusions and recommendation solely on technient considerations.
The other office feels obligated tojustify its decision to nilow plant operation despite the unresolved te:hnicalissues. How the Commission resolves this 3roblem will depend,in part, on whether higher priority is given to protecting pu alic hen!th and safety or the financial int: rests of the nuclear industry.
Sincerely, Robert D. Pollard Nuclear Safety Engineer
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Docket No S0-29g U**ary 14, 3993 Mr Nuc. Guy R. Horn liebraska Public Plear Power Group M Post Off anager Columbus, ice Box 499ower District
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Dear Mr. Horn:
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SUBJECT:
DELAY IN By letter datedCOOPER NUCLEARSUBMll1At 0F 5 TAT 10N (TAC NOREACTOR VESSEL r quested e
that October 28 for thCooper Nuclear Stthe NRC gran,t a1992, the N SUR
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1991. eAppendix Hsurveillanceation ;CliS) r schedularRESUL TS FOR material surveilla extension for theublic Power Distr capsule withdrawn fCFR Part 50 requi eactor to 10 the v
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vessel extenston to the 1er 31, 1992withdr wn results withdrawal. cimens s
granted in the past f on O of tests Thus must be t owever,ctober31, 1991 shconducted on,the In your October 28 or good cause requir year under Appendix H allows the ould have been 1, 1993withdra,wn1992 letter, yot shown.ement, and such ex capsule to March the NRC also ons have beento grant a on October 31 requested stated In capsule to the testithat, due to state that support of you,r 1991, bethat the you during early 1991ng facility was delacircumsta 1991 ashipping tasks (er)e 1
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certifiedsingle GE-Nuclear E er 31 removed from previously-cert ifiyed by nearly 6 mo,t control extens ion,,1992, for nuclear industry f serv, ice due to sh n hs.ipment of use by the NRC, gy cask of a ner you the GE-Nuclear E the ed GE uclear EnergSpecifically, or capsule on ner what was and (3) theresafety concerns N
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schedular not submittal ofextensiondispute the ented shipment ofnd by the to summary report. enabove points, we b l ng cask could have be review the the CNS in hat you and act on your This would havesubmitted well befoe ieve that your hanges toreview your request before the d ircumstances jeopardize allow early identificationsystem for tracking re the deadline forequest for afforded us the eadlin a
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Mr. Guy R. Horn.
With regard to the safety aspects of the requested schedular extension, the CNS reactor vessel material, configuration, and irradiation history are such that we do not believe that a safety problem would result from the delay of the summary report until March 1993.
Specifically, the current pressure-temperature (PT) limit curves in the CNS technical specifications are valid up through 21 effective full-power years (EFPY) of operation, whereas the CNS vessel now has experienced approximately 12 EFPY of operation.
Based on the above considerations, we hereby grant a schedular extension until March 1, 1993, for submittal of the report on the reactor vessel material capsule specimens removed from the CNS vessel on October 31, 1991.
Sincerely, ORIGINAL SIGNED BY.
George T. Hubbard, Acting Director Project Directorate IV-1 Division of Reactor Projects - Ill/lV/V Office of Nuclear Reactor Regulation cc:
See next page pl11RIJll)llM:
Docket file NRC & Local PDRs PD4-1 Reading M. Virgilio G. Hubbard P. Noonan ACRS (10) (P-315)
OGC (15818)
A. B. Beach, RIV PD4-1 Plant File H. Rood J. Gagliardo, RIV J. Strosnider J. Tsao E. Collins, RIV K. Wichman
- 6. Elliot R. Kopriva, RIV J. Roe
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NAME o an HRoo k JStrosnider CBarth GHubb'ard M
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\\ /6/93l 12/23/92 12/29/92 / /# /93l / //Lf/93 I /l@93 OfflCIAl RECORD COPY Document Name:
C0085167.ltr
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Mr. Guy R. Horn,
With re ord to the safety aspects of the requested schedular extension, the CNS reactor vessel material, configuration, and irradiation history are such that we do not believe that a safety problem would result from the delay of the summary report until March 1993.
Specifically, the current pressure-temperature (PT) limit curves in the CNS technical specifications are valid up through 21 effective full-power years (EfPY) of operation, whereas the CNS vessel now has experienced approximately 12 EfPY of operation.
Based on the i
above considerations, we hereby grant a schedular extension until March 1, 1993, for submittal of the report on the reactor vessel material capsule specimens removed from the CNS vessel on October 31, 1991.
Sincerely,
,Y'<'ef Y o,
George T. Hubbard, Acting Director Project Directorate IV-1 Division of Reactor Projects - I!!/IV/V Office of Nuclear Reactor Regulation P
cc:
See next page
~
Mr. Guy R. Horn Nuclear Power Group Manager Cooper Nuclear Station cc:
Mr. G. D. Watson, General Counsel Nebraska Public Power District P. O. Box 499 Colui.ibus, Nebraska 68602-0499 Cooper Nuclear Station AT1N: Mr. John H. Meacham Site Manager P. O. Box 98 Brownville, Nebraska 68321 Randolph Wood, Director Nebraska Department of Environmental Control P. O. Box 98922 Lincoln, Nebraska 68509-8922 Mr. Richard Moody, Chairman Nemaha County Board of Commissioners Nemaha County Courthouse 11124 N Street Aeburn, Nebraska 68305 Senior Resident inspec. ir U.S. Nuclear Regulator;, Commission P. O. Box 218 Brownville, Nebraska 68321 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Mr. Harold Borchert, Director Division of Radiological Health-Nebraska Department of. Health 301 Centennial Mall, South P. O. Box 95007 Lincoln, Nebraska 68509-5007 LL..