ML20072G341
| ML20072G341 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/17/1983 |
| From: | Jeffrey Riley CAROLINA ENVIRONMENTAL STUDY GROUP |
| To: | NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| References | |
| NUDOCS 8303240584 | |
| Download: ML20072G341 (13) | |
Text
.RELATED (USWEMIERW UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION-. -- -
BEFORE THE ATOMIC SAFETY 1ND LICENSING BOARD
'53 r.
23 #1L'2 l
In the Matter of
)
)
)
Docket Nos. 50-413
)
50-414 (Catawba Nuclear Station,
)
Units 1 and 2)
)
CESG'S RESPONSE TO NRC STAFF'S SECOND SET OF INTERROGATORIES AND DOCUMENT PRODUCTION REQUESTS CESG responds herewith to NRC Staff's interrogatories and document production requests of Dec. 15, 1982.
This is filed after the 14 day period for a sequence of reasons including the press of other obligations and an illness on the part of the responder.
The matter has been discussed with Mr. Johnson, Staff counsel, who indicated that it would be satisfactory to respond as soon as able.
As appropriate, the responses to interrogatories will be made in six parts per the Staff request.
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By " reactor degradation" CESG means a reduction in the capacity of the reactor vessel to withstand the Pange of internal pressures to which it will be subject in routine and faulted operation.
Crack formation in welds is an early step in degradation.
Crack growth is driven by stresses.
Internal pressure, which increases the reactor periphery, results in stress concentration at the reg. ion of minimum radius, the leading edge, of the crack.
A temperature gradient in which the interior of the reactor is cooler
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than the exterior, contributes stress due to a differential in thermal expansion, further increasing stress concentration in the leading edge of the crack.
Crack propagation occurs more readily in brittle than in ductile materials.
A high nil ductility crack propagation can occur under conditions temn no+ue. ----
+w 8303240504 830317 M
gDRADOCK05000
-2 t
which would be encountered of infection of cooling water into cn appreciably pressurized reactor.
The combination of cracks and incipient cracks wi,th an increased nil ductility (brittle) reference temperature constitutes reactor degradation.
It-would be pointless and burdensome to list texts, articles, news items, and 10CFR items well known to the parties.
J. L. Riley provided the foregoing response and will provide the subsequent responses.
If CESG introduces a witness in this area it will probably be Riley.
If another witness is secured the parties will be promptly advised, Part (5) does not apply.
See response to part (1) for (6).
For the remaining interrogatories response to these items will be made only if it differs from the foregoing.
2.
What are the physical or chemical mechanisms through which such
" reactor degradation" is contended to occur?
See the response to 1 foregoing for the mechanical model by which the physical capability of the reactor degrades.
The writer cannot speak to the mechanism by which neutron fluence alters the morphology of steel so as to increase the nil ductility reference temperature other than to note that ductile structures tend to be, for crystallizing materials, associated with small crystallite Fatigue, which is also a factor in embrittlement, is size.
associated with growth in crystallite size.
3.
Do you contend that such " reactor degradation" will occur at Catawba?
CESG contends that reactor degradation will occur at Catawba.
4.
If your answer to the previous interrogatory is affinnative, explain fully the basis for drawing conclusions about Catawba
" reactor degradation" based on the experience of other reactors?
I doubt that there is any contest that the Catawba reactors will degrade.
The key question is--will the degradation be of a magnitude that has safety significance.
While it is advanced by Staff that the reduced copper content of the steel in the Catawba reactors and weld materials will reduce the rate of degradation, in my view there has not been sufficient experience to make this a reliable conclusion.
Just as the s~ tate-of-the-art at the time of fabricating the Oconee generation reactors was not adequate to provide an acceptably low rate of degradation, I am unaware of an evidentiary basis that the present state of the art can reliably predict the low rates of embrittlement which Staff posits.
5.
What do you mean by "much more rapid increase in reference temperature than had been anticipated," as used in CESG Contention 18 (Palmetto 44)?
I have not seen the predictions made for the Oconee generation of reactors of the rate of nil ductility reference increase.
However news itens I have read indicated that the rate was greater than expected.
It is also a matter of common sense that it would be unlikely that a large investment would be made in a plant which would, in a fraction of its lifetime, degrade so as to fail to meet specific safety criteria.
6.
Please explain the conditions, timing, events, properties, through or by which this "much more rapid increase in reference temperature", etc., has occurred?
I am only able, as in the foregoing response, to indicate that 2
the changes in embrittlement rate occurred more rapidly than had been
. 4 expected.
I reject the burden of providing an explanation for a phenomenon which, to my knowledge, remains, in a fundamental, theoretical sense,' unexplained but which is present as fact.
.7.
What do you contend to be the " anticipated" increase in reference temperature at Catawba?
I do not contend a specific increase in reference temperature at Catawba.
I do contend that knowledge about rate of embrittlement was deficient at the time of the Oconee generation of reactors and that the validity of present knowledge has yet to be demonstrated.
8.
What is the basis for your answer to the previous interrogatory?
See the preceding response.
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9.
De you contend that the "much more rapid increase in reference temperature than anticipated" which you state has occurred at "a number of PWR's" will occur at Catawba?
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I do not contend that the magnitude of the change will be the same as for high copper steels.
I contend that the rate is u'certain; n
that the' kinetics of change are not established; that the rate is not necessarily linearly time dependent.
10.
Please explain fully the basis for drawing conclusions about increase in Catawba reference temperature based upon the experience of other reactors.
The experience with other reactors indicates that some years ago the experts. in the embrittlement field were no,t able to make accurate predictions of embrittlement rate.
I have seen no evidence that current expert predictions are reliable.
In view of the fact that without neutron' fluence there is fatigue and thermal stress embrittlement, there is every reason to expect an
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increase in nil ductility reference temperature, but at a rate
5-which will be revealed by experience but which is not presently known.,
- 11. Please identify the "PWR's" referred to'in CESG Contention 18 (Palmetto 44).
' Fort Calhoun, Robinson-2, San Onofre-1, Maine Yankee, Palisades, Yankee-Rowe, Oconee-l.
- 12. Doyoubaseyourstatementsthat"[t]helicenseshouldnotissue" and " Catawba should not be permitted to operate" upon failure of Catawba to meet any regulatory requirement?
No.
Rather on the exemptions granted by the NRC for plants not meeting regulatory requirements.
13.
If the answer to the previous interrogatory is affirmative, please identify each such regulatory requirement and explain your basis for applying it to Catawba.
Response not required.
14.
If the answer to Interrnaatory 12 is negative, please explain fully the authority and the basis for your position that " Catawba should not be permitted to operate" and that [t]he license should not issue".
The granting of exemptions from Appendix H, 10CFR 50, as for Oconee-1,
-2, and -3 presages the grant of similar exemptions for Catawba if they will permit operation after the plant falls outside the scope provided by regulation.
- 15. Do you contend that the Oconee plant was shut down for half a year because of "much more rapid increase in reference temperature than had been anticipated"?
Among other reaso'ns, yes.
16.
If the answer to (a) above is affirmative, please explain fully the basis for your contention.
It took time to remove and reload the full core in order to make an ultrasonic inspection, as well as to make the inspection.
However I have not contended anywhere that' the reactor was shut down
half a year for this cause.
17.
Please clarify the meaning'of your statement that "there was a series of cracks and welds shown...".
Typographical. error.
The reporter should have written ".
cracks in welds
- 18. What is the relevance of the quoted statement in Interrogatory 17 to your contention that there has been "a much more rapid increase in reference temperature than had been anticipated."
None.
See response to 17
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- 19. What is the relationship between your statement concerning " cracks and welds shown..." and the adjusted reference temperature of the Catawba reactor vessels?
None.
See respons to 17 20.
Do you contend that the Catawba reactor vessels were manufactured or otherwise "provided" by Babcock & Wilcox?"
No.
2'1. 'Do' you contend any unit of Oconee was shut down because of
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" reactor degradation" or " reactor embrittlement"?
Yes.
Temporarily to examine the coupons for change in nil ductility
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reference temperature and to examine welds for the number and dimensions of cracks.
22.
If your answer to the previous interrogatory is affirmative, explain fully the relationship, if any, between the Ocenee shutdown and such " reactor degradation" or " reactor embrittlement".
The shutdown was made-to monitor the magnitude of reactor degradation and embrittlement.
23.
Do you contend that the Catawba reactors will be shut down because of "a much more rapid increase in reference tempe ature than anticipated"?
l l
The Catawba reactor, under the regulations, will receive periodic shutdowns for monit,oring purposes.
The duration of the shutdown
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will depend on a variety of factors' including the degree of embrittlement, the number, location, and size of cracks, the regulat' ions in force at the time, the willingness of the NRC to provide an exemption, if that seems to be required, and the representations of Duke to the NBC re need for power, need for earnings, etc.
24.
If your answer to the previous question is affinnative, explain ecch and every basis for your position.
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It is generous to suggest that I can " explain each and every basis for my position".
I freely admit making no claim to omni-prescience.
The considerations given foregoing should suffice to indicate my line of thought.
~2 5'. If your answer to Interrogatory 23 is affinnative, how often and how long do you contend the Catawba reactor will be shut down?
I cannot make any such explicit prediction.
I will resort to the standard probabilistic refuge of the staff and say that there are finite probabilities that the reactor may be shut down for varying periods of time.
Absent an empirical basis for assigning such probabilities I will answer, au Judge McMillan, I'm n'ot a bookie.
- 26. Do you contend that a " change in the capacity factor" should be reflected in the environmental cost-benefit analysis for Catawba?
It would be reasonable to alter the capacity factor, but I doubt ths.t the Staff is in any better position than I to.be quantitative.
It should be born in mind that Duke initially anticipated a capacity factor of 80% for the sister McGuire units.
Experience has shown that 50% is a better reflection for 1000 MW plus units.
Considering its present steam generator problems at McGuire, and prospectively
'at Catawba 1, Duke would be delighted..to reach 50% without embrittlement considerations.
_8_
27.
If your answer.to the previous question is affirmative, explain in detail what cost, if any, should be " entered" and the basis for such costs.
See preceding response re quantifiability.
- 28. Do you contend that the environmental cost-benefit evaluation for Catawba's operating license is inadequate or incomplete unless it reflects a cost for shut down time attributable to eith.er "much more rapid increase in reference temperature", or " reactor embittlement", or both?
Yes.
29.
If your answer to tile previous question is af'innative, explain f
fully each and every factual and legal basis for your answer thereto.
Factual:
a conservative position would provide a lower limit of electrical production based on the experience indicated in response to interrogatory 11.
An upper limit would be no effect.
The lower limit would be e.stimated excluding exemption from the 100 F terminal value for nil ductility reference temperature.
Legal:
compliance with the NEPA requires a defensible and realistic definition of the alleged benefit.
30.
Describe fully any " reactor degradation" which you contend has occurred at Oconee, including the sources for the description provided.
A report on an NRC draft letter, dated July E8, 1981, reported in INSIDE NRC. - August 10, 1961, p. 5 referring to "signiricantly reduced fracture toughness";
letter from an NRC employee who witnessed the Oconee UT; IE Bulletin No. 83-02 indicating serious unanticipated stress corrosion cracking in large diameter stainless steel recirculation piping at BWR plants; paragraph II.C.3.a of Appendix H, 100FR part 50 vs. Staff response of Dec. 15, 1982 to PA interrogatories 3- "... an RT as of December 31,
- NDT, 1981, at the vessel ID greater than 150 F," and 35; and Duke's
-9 responsc to same interrogatory 35
- 31. Do you contend that Applicants will not be in compliance with any provision of 10 CFR, Part 50, Section 50.55a, and Appendices G and H?
It has happened at Oconee, see response to interrogatory 35 (b).
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If present forecasts of embrittlement rate are in error it may happen again.
32.
If your answer to the previous interrogatory is affinnative, what specific paragraph do you contend will not be met, and what provision thereof? When do you contend it will not.be met?
Appendix H, III C 3 a re adjusted reference temperature not exceeding 100 F.
Foregoing makes clear there is no present empirical basis for projecting a date.
33.
If your answer to Interrogatory 31 is affinnative describe fully the factual basis for your position.
NA.
- 34. Do you challenge the sufficiency of any standard contained or referenced in Appendix G and H of Part 50?
Yes.
35.
If your answer to the previous interrogatory is affirmative, please identify each standard which you contend to be insufficient, and
' explain the basis therefor..
I am dubious about the entire construct.
If the standards were really adequate there would be no need for a surveillance ~ program.
If they had been adequate the reactor degradation problem would not have occurred; IE would not be issuing bulletins on stress corrosion cracking or the failure of fasteners.
It would not be necessary to further school persons in UT crack detection.
The specification of res.ctor safety is an uncompleted art.
O
36.
If your answer to Interrogatory 37 is affinnative, what do you contend would be sufficient standards for purposes of Appendices G
['
and H?
[
I would obviate the need for such standards by denying licenses to all nuclear plants, in view of the ~ cos t; of a major failure.
37.
Explain the basis and reasons for your answer to the previous interrogatory.
l See response to 36.
- 38. Do you contend that the method of calculating adjusted reference temperature specified in Reg Guide 1.99.is in any way insufficient?
The sample size is inadequate, the variance excessive.
39.
If your. answer to the previous interrogatory is affirmative, j
identify each way you contend such method is insufficient?
See response to 38.
- 40. Do you challenge the end of life adjusted reference temperature for the Catawba reactor pressure vessels, calculated pursuant to the methodology specified in Reg Guide 1.99?
Yes.
41.
If your answer to the previous interrogatory is affinnative, identify each and every basis for your answer.
See response to 36.
A linear time function is assumed.
The method is nonconservative in that no effort is made to find the weakest sample.
Failure will, of course, occur at the weakest point.
It is optimistic to assume the complete homogeneity of weldments and, for that matter, of base stock for alloys.
42.
If you challenge the end of life adjusted reference temperature for either of the Catawba reactor pressure vessels, please specify the values you believe to be correct, and explain your method of arriving at the values so reached, including the reasons for using such methods.
The question incorrectly assumes that I think a correct value can be arrived at based on available knowledge and information.
- 43. Do you contend that either of Cstawba's reactor pressure vessels can exceed the end of life adjusted reference temperature stated in your answer to the previous interrogatory during the operating life of the reactors.
See responses to 38, 41 and 42.
44 If your answer to the previous interrogatory is affirmative, please give the basis in full for your answer.
NA.
- 45. 'At what point in the. life of the Catawba reactor pressure vessels do you contend the adjusted reference temperature calculated pursuant to Reg Guide 1.99 will be reached.
At a different point than the Staff thinks it will.
Consider the cases of Oconee et.al.
46.
Explain the reasons for your answer to Interrogatory 48.
"48"?
Typo?
- 47. Do you contend that all. reactor pressure vessels have the same
. characteristics with respect to the increase during the operating life of the vessel in reference temperature?
No.
48.
Explain the basis for your answer to the previous interrogatory.
There apparently is a scatter pattern of embrittlement rates for operating reactors.
- 49. Upon what variables do you contend the rise in reference temperature depends?
Neutron fluence, composition, stress history for routine heatup and cooldown, crack nucleation, stress due to inverted radial temperature gradients in ECCS operation.
- 50. How do each of the variables enumerated in your answer to the previous interrogatory affect the rise in reference temperature.of the Catawba reactor press ~ure vessels? Explain the basis for your answer.
e
. o It is clear from the finding of cracks in welds as at Oconee that initiation and growth occurs under conditions in which the ET sa ep a e
e wu seem reas nable to think NDT that cracks would not start nor propagate under conditions of ductility and design levels of stress.
But cracks have formed and have propagated.
Once initiated, the stress for crack propagation is a small fraction of that for crack initiation and propagation in a whole piece.
The Charpy V notch test utilizes stress concentration at a small radius to induce controlled cracking in the brittle regime.
Nowhere in the Appendix H regulation do I see the ma.tter of such crack initiation and growth dealt with.
To read the material, an operation living up to technical specifications would not permit crack development.
Having made l't clear that there are unrecognized variables in the situation it is simple enough to remark that nil ductility reference temperature will rise with neutron fluence,. rise more rapidly as trace elements including copper, p osphorus and nickel are present in increasing concentrations, rise with the fatigue associated with routine stress history, particularly the temperature gradients present during warmup and cooldown, rise with the increasing concentration of morphological crack initiation sites, and particularly with inverted temperature gradients due to ECCS cooling under pressurized conditions.
I affirm that the foregoing responses are true, correct, and complete insofar as I am able to provide them..
,s
March 17, 1983 y
__ _[
Jesse L. Riley for CESG'
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C,0MMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
DUKE POWER COMPANY, et al.,
)
Docket Nos. 50-413
)
50-414 (Catawba Nuclear Station,
)
Units 1 and 2)
)
AFFIRMATION OF SERVICE I hereby affirm that copies o
'CESG's Response to NRC Staff's Second Set of Interrogatories and Document Production Requests"'
in the above captioned matter has been served on the following in the U.S. mail, first class, this 21st day of March, 1983 James L. Kelley, Chairman Michael McGarry, III, Esq.
Atomic Safety & Licensing Board Debevoise and Liberman U.S.
Nuclear Regulatory Commission 1200 17th Street, N.W.
Washington, D.C.
20555 Washington, D.C.
20036 Dr. Dixon Callihan William L. Porter, Esq.
Union Carbide Corporation Albert V. Carr, Esq.
P.O. Box Y Duke Power Company Oak Ridge, Tennessee P.O. Box 33189 Charlotte, NC 28242 Dr. Richard F. Foster P.O. Box 4263 Ge orge E. Johnson, Esq.
Sunriver, Oregon 97701 Counsel for NRC Staff U.S. Nuclear Regulatory Commission Atomic Safety & Licensing : Appeal Panel Washington, D.C.
20555 U.S.
Nuclear Regulatory Commission Washington, D.C.
20555
. Robert Guild, Esq.
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< Attorney for the Palmetto A111ance Atomic Safety & Licensing Board P. O. Box 12097 Panel
' Charleston, South Carolina 29412 U.S. Nuclear Regulatory Commission
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Washington, D.C.
20555 Palmetto Alliance 21359 Devine Street Docketing and Service Section.
Columbia, SC 29205-U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Henry Presler, Chairman Charlotte Meck. Env't'1. Coalition Richard P. Wilson, Esq.
942 Henley Place Assistant Attorne'y General Charlotte, NC 28207 2600 Bull Street Columbia, SC 29201 WM 0. l1 ui Jes'se L. Riley for yrESG