ML20052D503
| ML20052D503 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/26/1982 |
| From: | Knotts J DEBEVOISE & LIBERMAN, SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML19277A831 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8205060516 | |
| Download: ML20052D503 (27) | |
Text
r ,
e April 26, 1982 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATCMIC SAFETY AND LICENSING BOARD In the Matter of: )
SOUTH CAROLINA ) Docket No. 50-395 OL ELECTRIC & GAS COMPANY, )
et al. (Virgil C. Summer )
Nuclear Station, Unit 1) )
APPLICANTS' RESPONSE IN OPPOSITION TO " PETITION TO INTERVENE AND REQUEST FOR HEARINGS" OF FAIRFIELD UNITED ACTION On April 9, 1982, Fairfield United Action (FUA) filed a petition for leave to intervene as a party of record in this proceeding and requested that evidentiary hearings be conducted on the issues raised in its petition.
Almost a year ago, on June 3, 1981, the Licensing Board dismissed FUA as a party pursuant to ALAB-642 dated 1981. 1/
June 1, The Appeal Boarsi decision was based on petitioner's unexcused lateness and lack of ability to contribute, and on the impact on,the rights of the existing parties. The Appeal Board stressed the impact on the rights of the existing parties in its decision denying FUA's earlier petition.
-1/ South Carolina Electric & Gas Co., et al. (Virgil C. Summer Nuclear Station, Unit 1), ALAB-642, 13 NRC 881 (1981),
reversing South Carolina Electric & Gas Co., et al. (Virgil C. Summer Nuclear Station, Unit 1), LEP-81-11, 13 NRC 420 t l (1981), appeal pending sub nom. Fairfield United Action v.
NRC, et al., No. 81-2042 (D.C. Cir.).
8205060516 820426 ,
PDR ADOCK 05000395 .
l G PDR
4
" Simple fairness to [the existing parties] --
to say nothing of the public interest require-ment that NRC licensing proceedings be conducted in an orderly fashion -- demand that the Board be very chary in allowing one who had slept on its rights to inject itself and new claims into the case as last-minute trial ALAB-642, 13 NRC preparations were underway."
ct 886. "Further, a delay in the progress of the proceeding is not merely a theoretical possibility but rather a very likely proximate result of the belated intervention [ footnote omitted]." Id. at 888.
The Appeal Board recognized that granting FUA's petition at such a late date would have seriously impaired the existing parties' right to pretrial discoverj and to invoke summary disposition procedures. I d,. at 888-89. The Appeal Board concluded:
"By ... remaining on the sidelines while the proceeding moved closer and closer to trial, it [FUA] voluntarily assumed the precise risk which has now materialized: that its partici-pation in the proceeding could no longer be sanctioned without destructive damage to both the rights of other parties and the integrity of the adjudicatory process itself." Id. at 895.
Since that decision was rendered, 22 days of evidentiary hearings have been held, 6,137 pages of transcript compiled, 70 exhibits admitted, and on January 20, 1982, the record l
was closed as to all issues (Tr. 6137; see also Tr. 3871, l
I I 3872, 4677, 6014). We are now truly at the eleventh hour.
In a briefing to the Commissioners on April 20, 1982, the Staff reported that the plant will be ready in May and that Board's initial decision is expected at the end of May.
(Priefing on Status and Assessment of Near-Term Operating Licenses, April 20, 1982, Tr. 29-30) The Applicants have an investment in the V.C. Summer Nuclear Station in excess of
$1.0 billion. It cannot reasonably be controverted that substantial carrying costs would be incurred by both owners if the plant were ready to load fuel and commence operation but was without a license; that both the State of South Carolina and the Federal Energy Regulatory Commission (FERC) j provide for recovery of such costs from customers; and that South Carolina Public Service Authority would incur further additional costs for purchasing capacity and energy to replace Summer generation. Finally, it is reasonably ascertainable that energy from Summer would be cheaper from an incremental cost standpoint for SCE&G than oil. We 1
explain later why there is no reason for a hearing on this 1 1
matter; but focusing for the moment on the practicalities, l no prompt hearing could be conducted without sacrificing trial preparation as discussed by the Appeal Board in Virginia Electric & Power Co. (North Anna Station, Units 1 &
2), ALAB-289, 2 NRC 395, 400 (1975). Delay would be measured in terms of months required for trial preparation, hearings, proposed findings, and decision.
Thus, the adverse impact on the rights of the parties and the very real consequences of delay are even more severe at this' juncture than when the Appeal Board wrote on the matter last year.
Undaunted, FUA has filed an even more untimely petition. In its " Petition to Intervene and Request for Hearings" dated April 9, 1982, FUA again seeks to inter-vene and to have the hearings reopened to take up the matter l
of accelerated steam generator tube wear (FUA Petition, at 2). FUA urges that 1) despite the Applicants' commitment to implement preventive measures involving monitoring and inspection during initial limited power operation, the license should be either denied or unspecified conditions imposed to protect the health and safety of the public and
- 2) a new cost-benefit analysis should be prepared which assumes an indefinite 35% capacity factor and 57% of the projected kwh listed as benefits at page 9-1 of the Final Environmental Statement (FES). FUA also purports to incorpo-rate by reference contention 14 on steam generators from its previously denied petition. Finally, FUA asks the Board take up these matters sua sponte should its petition be denied (FUA Petition, at 14).
l At some point, there must be an end to litigation. 2/
As the Appeal Board pointed out last year in this case, 2/ For a complete discussion see Cleveland Electric Illumi-nating Co., et al. (Perry Nuclear Power Plant, Units 1 l and 2), ALAB-443, 6 NRC 741, 750 (1977); Houston Lighting and Power Co., et al. (South Texas Project Unit Nos. 1 and 2), ALAB-381, 5 NRC 582, 591 (1977); Duke Power Co.
(Catawba Nuclear Station, Units 1 and 2), ALAB-359 4 NRC 619, 620 (1976); Northern Indiana Public Service Co. (Bailly Generating Station, Nuclear-1), ALAB-227, 8 f
whether a subject should be treated as a hearing issue is not the same question as whether it deserves agency attention and action. ALAB-642, 13 NRC at 895-96. The NRC Staff and the Commission are well apprised of the accelerated tube wear experienced in preheater type steam generators at foreign units after periods of full power operation and are overseeing industry's program to remedy the problem. The Applicants are committed to operate the'V.C. Summer facility initially at reduced power levels, to monitor and inspect for tube degradation, and to make modifications as indicated.
FUA, on the other hand, has given no indication as to what evidence it would offer, has shown no technical ability to contribute to resolution of this matter, and has (footnote cont'd from previous page)
AEC 416, 418 n.4 (1974); see Vermont Yankee Nuclear Power Corp. v. NRDC, 435 U.S. 519, 554-55 (1978); ICC v.
Jersey City, 322 U.S. 503, 514, (1944); see also United States v. ICC, 396 U.S. 491, 521 (1970). In ICC v.
Jersey City, supra, 322 U.S. at 514 the Supreme Court recognized:
" Administrative consideration of evidence ... always creates a gap between the time the record is closed an'd the time the administrative decision is promulgated.
This is especially true if the issues are difficult, the evidence intricate, and the consideration of the case deliberate and careful. If upon the coming down of the order litigants might demand rehearings as a matter of law'because scme new circumstance has arisen, some new trend has been observed, or some new fact discovered, there would be little hope that the administrative process could ever be consummated in an order that would not be subject to reopening."
f .
not specified what if any, additional ane:/or alternative corrective actions should be taken or what license conditions it would impose. Given these circumstances, the key questions are: 1) Should the Licensing Board (as distinct from the NRC Staff and the Commission) take up this particular issue?
- 2) What purpose would be served by holding a hearing since the Applicants and Staff have matters well in hand? 3) What would be the impact on the rights of other parties?
See discussion at pages 1-3, supra. The answers to these questions emerge from the familiar analyses of late intervention standards and are buttressed by consideration of the authorities regarding reopening.
Legal Standard Governing Petitions for Intervention The standards governing late intervention are set forth at 10 C.F.R. $ 2.714(a)(1). 3/ Five factors are balanced in deciding whether to grant or deny a late-filed petition: (1) good cause, if any, for failure to file on time; (2) availability of other means whereby the petitioner's interests will be protected; (3) extent to which the peti-tioner's participation may reasonably be expected to assist 3/ Houston Lighting & Power Co. (Allens Creek Nuclear Generating Station, Unit 1), ALAB-671, slip op. (March 31, 1982); Project Management Corp., et al. (Clinch River Breeder Reactor Plant), ALAB-354, 4 NRC 383, 388-94 (1976). The Allens Creek case involves strikingly similar facts to the instant case, in that it involved a second petition by a person previously denied ,
intervention.
9 in developing a sound record; (4) extent to which the l
petitioner's interest will be represented by existing parties; and, (5) the extent to which the petitioner's participation will broaden the issues or delay the proceeding.
10 C.F.R. $ 2.714(a)(1)(i)-(v). FUA's petition refers to 10 C.F.R. $ 2.714(a)(1) and touches upon the five factors but the overall showing is weak.
Good Cause. First, as to good cause, FUA argues that it only recently became aware of accelerated tube wear and leaks caused by flow-induced vibrations in Westinghouse-Model D steam generators. It contends that prior to the Board Notification, BN-82-02, dated January 20, 1982, subsequent correspondence between the Applicants and the agency, NRC memoranda, and newspaper articles in the Columbia Record and The State it had no way of knowing about potential accelerated steam generator tube wear (FUA Petition, at 8-10).
In its petition to intervene filed March 23, 1981, FUA sought to raise a contention on steam generators (FUA Supplement to Petition to Intervene, Contention 14, March 23, 1981). 4/ Although that contention focused on occupational i
4/ In its latest petition to intervene, FUA attempts to I incorporate by reference matters set forth in its l original petition to intervene and supplement dated March 23, 1981. Such matters are not before the Board as Contention 14 was one of the group of 17 contentions initially denied by the Licensing Board (See Partial (footnote cont'd on next page)
I
exposure to workers, 5/ it stated that " Westinghouse steam generators have demonstrated a generic tendency to denting, cracking, leaking and rupturing" and notes that extensive.
l repairs, including replacement of the steam generators, have been required at some plants. (Id.) FUA's representative i
said the following about its contention 14 at the April 8, 1981 prehearing conference:
"DR. RUOFF: given the history of Westinghouse steam generators, even recognizing that this D-3 Model is a model which was only in place at this plant and one in Sweden ... it is my understanding that the differences in the steam generator are not significant from other models. And given the proportion of Westinghouse steam generator nadels, which have had difficulty, this is something that can reasonably be expected to occur during the design life of the plant.
And we would strive to get a witness to discuss those issues.
I (footnote cont'd from previous page)
Order Following Prehearing Conference, LBF-81-11, 13 NRC 420, 428 (1981)). FUA's inability to contribute to development of a sound record was weighed most heavily against it by the Board in rejecting these contentions.
That inability to contribute has not been cured in this petition.
-5/ The issue of occupational radiation exposure to nuclear plant workers has already been addressed by the Staff in the FES. The Staff determined that the Applicants are committed to design features and operating practices that will assure that individual occupational exposure is maintained within the limits of 10 C.F.R. Part 20 and ALARA. (FES, May 1981, at 4-23). Based on actual operating experience, average annual reactor doses of up to 1300 persons-rems over the life of the plant may be expected at Summer, but the actual doses vary greatly depending on factors such as the amount of required routine and special maintenance and the degree of plant operations (Id. at 4-23 to -24). Thus, the FES has already accounted for activities such as steam generator maintenance and modification in determining occupational exposure levels.
_9-CHAIRMAN GROSSMAN: But yoQ do not have one available now?
DR. RUOFF: No."
(Tr. 623).
As to the group of 17 contentions not admitted, including contention 14 on steam generators, the Licensing i
Board in its order following the prehearing conference stated, "we conclude that the good cause, delay, and ability-to-contribute to-a-sound-record factors weigh heavily against admission." LBP-81-ll, 13 NRC 420, 427. As stressed by the Appeal Board in denying FUA's original petition to intervene, a petitioner who files at the lith hour, has a very heavy burden to carry. ALAB-642, 13 NRC at 886,888.
In this instance, as before, the good cause factor should be weighed against FUA.
Availability of Other Means and Representation by Existing Parties. Turning to the second factor, the avail- i ability of other means whereby the petitioner's interests would be protected, we will combine our discussion of that factor with the fourth factor, the extent to which the petitioner's interest will be represented by existing parties. FUA states at pages 10 and 12 of its petition that no existing parties will represent its interest and that no other forum exists to protect such interest. With regard to f
other available means or forums to protect its interest, FUA seems to assume that any and all incompletely resolved I
matters regarding licensing of a nuclear plant should be resolved in adjudicatory hearings. This is not the practice of the Commission, nor is it the purpose of the hearing process. Many matters in the licensing process are handled routinely without hearing as part of the Staff's review prior to licensing or thereafter. The Appeal Board discussed this point in its decision denying FUA's earlier petition to intervene: l "It does not follow from FUA's exclusion from the proceeding that its concerns I perforce will be ignored in the licensing i of this reactor.... To the extent that they go beyond the bounds of the hearings as fixed prior to the belated FUA intervention attempt, under the long-prevailing regulatory scheme these concerns fall within the province of the Staff." ALAB-642, 13 NRC at 895.
"As to those aspects of reactor operation not considered in an adjudicatory proceeding (if one is conducted), it is the Staff's duty to insure the existence of an adequate basis for each of the requisite Section 50.57 determinations."
(Id. at 896).
This is precisely the situation we have here.
We do not claim that it is the function of the Applicants or the Staff in this proceeding to protect FUA's litigative interests. But those parties are protecting the public health and safety outside the hearing process.
To the extent that FUA is interested in the public health and safety (as distinct from opposing the licensing of this plant), the NRC as a whole protects that interest outside the hearing process and FUA is free to make known its
views to the Staff or the Commissioners. As is evident from the affidavit of W.D. Fletcher, Westinghouse Corporation, the affidavit of Michael D. Quinton, SCE&G, and correspondence with the NRC Staff attached hereto, SCE&G has been diligently working with Westinghouse and the Staff to correct the problem with full appreciation of its significance. We believe that the Staff has been on top of the matter and, as evidenced by the documents attached hereto, is engaged in a program involving cooperative efforts of the industry and the utilities to guard against and remedy the tube wear problem. (See Attachment A, Steam Generator Status Report, February 1982, SECY-82-72, February 18, 1982). For these reasons, in the present circumstances, and especially given FUA's lack of demonstrated ability make a technical contribution to this matter, there is no reason why the Staff cannot protect FUA's interest outside the hearing process.
FUA argues that it cannot rely on Mr. Bursey to represent its interest in this matter. Nonetheless, FUA initially took the risk of relying on the existing intervenor.
(FUA Petition to Intevene and Request for Hearings, March 23, 1981, at 3-4) On this factor, the Appeal Board in Puget Sound Power & Light Co. (Skagit Nuclear Power Project, Units 1 and 2), ALAB-559, 10 NRC 162, 172-73 (1979) wrote:
4
"the promiscuous grant of intervention petitions inexcusably filed long after the prescribed deadline would pose a clear and unacceptable threat to the integrity of the entire adjudicatory process. [ citations omitted]. More specifically, persons potentially affected by the licensing action under scrutiny would be encouraged simply to sit back and observe the course of the proceed-ing from the sidelines unless and untl they became persuaded that their interest was not being adequately represented by the existing parties..."
Ability to Contribute. As to the third factor, the extent to which the petitioner's participation may reasonably be expected to assist in developing a sound record, FUA contends that only through its participLtion will the Board have a complete and sound record on this matter (FUA Petition, at 12 ) . The question of a petitioner's ability to contribute to the development of a sound evidentiary record was most recently addressed by the Appeal Board in Allens Creek, supra March 31, 1982 slip op. at 10. The Appeal Board in this case weig'aed the ability to contribute factor heavily against FUA in denying its earlier petition to intervene.
ALAB-642, 13 NRC at 891-93. As previously discussed, the Board, in ruling on the 17 contentions not admitted in FUA's first petition to intervene, found the ability to contribute factor " weighed most heavily against the petitioner." LBP-81-ll, 1
I 13 NRC at 426. With regard to those contentions including steam generators) originally denied, the Board stated, "we can only contrast petitioner's familiarity with the substance of these issues [those originally admitted] with its lack of prior (footnote cont'd on next page) l l
h involvement or expertise in the other areas it raised. On those other issues, it named few or no witnesses committed to testifying on its behalf, but sought mainly the opportunity to search for such witnesses. In view of the late date, we see no reason to afford that opportunity." Id.
In its most recent petition, FUA makes no showing as to any ability to contribute sound evidence; and gives no indication as to what expert witnesses or other evidence, if any, it would offer on this matter. 6/ Further, FUA does not specify what actions, if any, it believes should be taken in lieu of or in addition to those to which the Applicants are already committed. Based on FUA's petition, there is no reason to believe that it can contribute meaningfully and substantially to resolution of this matter.
When a petitioner has no ability to contribute to development of a sound record, the fourth factor (as to protection of its interest by existing parties) should not be weighed very heavily and the second and third factors, i.e., other means by which its interest may be protected and its ability to contribute, should be weighed quite heavily.
See ALAB-642, 13 NRC at 895. We can well understand that the Board would want assurance that the Applicant and the Staff are aware of the need for corrective action, and that it is being implemented by the Applicant with appropriate
~6/ See Detroit Edison Co. (Greenwood Energy Center, Units 2 and 3), ALAB-476, 7 NRC 759, 764 (1978).
?
I oversight by the regulatory staff. If that assurance is forthcoming, as is shown by the attached affidavits and l
documents, and we would expect by the Staff response, there is nothing to be gained by further hearings. If the matter will be resolved so as to protect the public health and safety without the petitioner's intervention, then the Board need not make any findings with regard to steam generators but should leave appropriate resolution of this matter to l the Staff. This is consistent with the role of the Board in an operating license proceeding under 10 C.F.R. I 2.760a vis-a-vis that of the Staff.
Delay. The last of the five factors in 10 C.F.R. $
2.714(a)(1) is the extent to which the petitioner's pa: _.ic ip a-tion will broaden the issues or delay the proceeding. FUA admits at page 12 of its petition that admission of these new contentions will expand the issues now before the Board, yet argues that because of their gravity, any delay is warranted. The Appeal Board, denying FUA's petition to intervene last year, stated that a late petitioner bears a heavy burden and the delay factor is extremely important.
ALAB-642, 13 NRC at 888-89; see Allens Creek, supra, ALAB-671, March 31, 1982 slip op. at 6, 11. As discussed at pages 1-3 of this response, the consequences of delay and the impact on the rights of the parties is extremely crucial at this point in the proceeding. FUA urges that any delay is 9
justified by the gravity of the issue. We address the significance of the steam generator issue in the following section by reference to the standards for re-opening. 7/
The Legal Standard for Reopening The Appeal Board reserved the question of whether the standards for reopening the record are applicable to a non-party in Allens Creek, supra, March 31, 1982 slip op. at 5 n.5. The reopening analysis is appropriate, however, because a non-party should not be held to a lesser standard than a party to a proceeding. This is a case in which assessment of the five factors for admitting a petition for late intervention would be ait.ed by the "significant" issue analysis applied to reopening. FUA itself suggests the " significance" analysis is appropriate in its argument on the fifth factor for intervention when it says expanding the issues and delay are warranted by the gravity of the matters it seeks to raise.
Briefly, reopening the record is based on appraisal of three factors: (1) Is the motion timely? (2) Does it address 7/ We briefly respond to the petitioners allegations regarding completion of the plant. FUA argues the Board "should not credit any estimates" of a fuel load readiness date (FUA Petition, at 12). Completion dates are estimates made in good faith of when the physical completion, checkout, and all of the numerous other items which go to make up readiness for fuel loading, will be completed assuming, generally, only a minor allowance for things going wrong. The most recent estimate of a fuel load readiness date is May 1982 (Briefing on Status and Assessment of Near Term Operating Licenses, April 20, 1982, Tr. 29-30).
significant safety (or environemntal) issues? (3) Might a different result have been reached or would the outcome have been affected had the newly proferred material been considered initially? G/ Further, to justify the granting of a motion to reopen, the moving papers must be strong enough, in light of any opposing filings, to avoid summary disposition. Vermont Yankee Nuclear Power Corp. (Vermont Yankee Nuclear Power Station), ALAB-138, 6 AEC 520, 523 (1973). Thus, even though a matter might be timely raised and involve a significant safety issue, no reopening of the evidentiary hearing will be required if the affidavits submitted in response to the motion demonstrate there is no genuine unresolved issue of fact, i.e., if the undisputed facts establish that the allegedly significant safety issue does not exist, has been resolved, or for some other reason will have no effect upon the outcome of the proceeding. (Id.)
The affidavits and other attachments hereto provide the basis for summary disposition on the pleadings. As is evident from review of the Commission and Staff papers, the Applicants' affidavit, and the Westinghouse affidavit, resolution of the problem of accelerated tube wear in Model 8/
For a complete discussion of the standards for reopening see Pacific Gas & Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2), ALAB-598, 11 NRC 876, 879 (1980) and Public Service Co. of Oklahoma (Black Fox Station, Units 1 and 2), ALAB-573, 10 NRC 775, 804 (1979) Kansas Gas & Electric Co., et al. (Wolf Creek Generating Station, Unit No. 1), ALAB-462, 7 NRC 320, 328 (1978).
i D3 steam generators is being capably pursued. We discuss these various documents in chronological order.
On January 20, 1982, in a letter from Mr. Eisenhut to Mr. Nichols, the agency requested that SCE&G provide information concerning its reliance on Westinghouse test data, testing at operational plants, its plans for instrumentation to detect flow-induced vibrations, and the testing and start-up procedures proposed for Summer (See Attachment B).
9 Also on January 20, the Staff issued a Board Notification, BN-82-02, on Preheater Type Steam Generators which included information on the results of testing and inspection of steam generators at Duke Power Company plants as well as two foreign plants (See Attachment C). The McGuire testing program discussed was conducted primarily at power levels of 50% or less. No significant tube wear as indicated (Id.).
An NRC memorandum from Mr. Chesnut to Mr. Youngblcod, dated February 19, summarized a meeting in Bethesda between the NRC, Westinghouse, and representatives of various utilities, including SCE&G, at which the problems with Model D steam generators were addressed (See Attachment D).
SCE&G, in a letter from Mr. Nichols to Mr. Denton dated February 19, 1982, outlined its plans to inspect for, monitor and correct problems associated with the Model D3 steam generators (See Attachment E). The letter describes
the Westinghouse testing program for Model D steam generators and proposes the following program for Summer: 1) normal low power testing, 2) complete startup testLig with power escalation up to 50% power, 3) continue operation at 50% power for approximately two months, or at power levels above 50% based on information available to preclude tube damage, 4) shut down and eddy current test rows 49, 48, and 47 of one steam generator, and 5) reevaluate available data i
to confirm continued limited power operation until a modifica-tion can be made to resolve the matter (Id.). Following that phase of the test program, SCE&G and Westinghouse are committed to jointly establish an operating power level for -
Summer considering: 1) results of the eddy current inspection,
- 2) experience and data from other operating plants, 3) status of the Westinghouse program, and 4) status of any proposed Westinghouse modification to resolve the problem (Id.).
Further, on April 14, SCE&G informed the Commission, in a letter from Mr. Nichols to Mr. Denton, that internal diagnostic instrumentation will be installed in one of the steam generators at the Summer plant as part of its monitoring program outlined on February 19 (See Attachment F).
At the request of the Commission, the Staff is engaged in an overall review of steam generators as a generic matter. In a memorandum from Mr. Dircks to Mr.
Minogue dated February 17, transmitting its Status Report on
on Steam Generators, the NRC outlined its current overall steam generator program (See Attachment A). The report defined the problem associated with tube degradation and its safety significance (which includes problems of tube wear arising from flow-induced vibration in preheater-type units); the NRC regulatory approach; current corrective actions; the NRC, industry, and foreign research and develop-ment activities; and, the long term approach for resolving !
the matter of steam generator tube integrity and the NRC/ industry program (See Attachment A). The NRC Status Report and an agency's outline of ongoing work were transmitted to the Commissioners as SECY-82-72 on February 18, 1982.
+
(See Attachment A).
In a memorandum dated March 25, 1982, SECY-82-72A, !
Mr. Dircks advised the Commissioners of the Staff's coordinated program to address steam generator problems. (Attachment G).
The program will involve the efforts of the Atomic Industrial Forum, Electric Power Research Institute, Steam Generator Owners Group, the ACRS, and the Staff to coordinate and manage research, review needs, and develop short-term and long-term solutions. The program will pursue the areas of materials, water chemistry and control, design and technical considerations (which includes problems arising from excessive vibration), secondary system components, primary and secondary side inspection, repair procedures and personnel exposure,
systems interactions, quality assurance, and operating experience.
As a result of these combined efforts, both the Applicants and the Staff, with the cooperation of the industry, have been fully aware the further work which needs to be done with respect to resolving the problems with tube
_ degradation in the Model D3 steam generators. The Applicants are committed to the Staff to correct the problem. (See Quinton Affidavit at 4).
_The-parties-have satisfied their obligation to report new information which is relevant to the proceeding.
See Georgia Power Co. (Alvin W. Vogtle Nuclear Plant, Units 1 and 2), ALAB- 291, 2 NRC 404, 411 (1975). In this instance, as in numerous others, a board notification was issued informing the Board of a potentially significant issue.
That alone, of course, does not determine whether a late petition should be granted or the hearing reopened. 8/ As we stated in an earlier response to Mr. Bursey's motion for admission of a new contention, while the Board might reasonably want assurance that the Staff is "on top of" the matter, it need not, in an operating license proceeding, take up every 8
~/ On April 14, 1982, Mr. Bursey filed a motion for admission of new contention and request for hearings on this and other issues to which we will be responding separately.
matter that crops up during ongoing Staff review, but may -
leave matters outside the hearing process to resolution by the Staff. (10 C.F.R. $ 2.760a; ALAB-642, 13 NRC at 895-96; see Grossman, Tr. 6136) (See Applicants' Response in Opposition to Intervenors Motion for Admissica of New Contention, March 11, 1982, at 6). The Applicants are confident that this matter can and should be so resolved.
It should be understood that the tube wear issue which FUA uses as a basis for its belated petition is only one part of the overall steam generator " picture" discussed in these NRC documents. Mr. Fletcher, of Westinghouse, states in his affidavit that significant tube wear will be precluded during interim operation and alleviated by a l
permanent modification (Fletcher Affidavit, at 2) . FUA raises issues concerning tangent point cracking of Row 1 f tubes, tube rupture events, PORV malfunction, and accident sequences. These issues are not concerns during the V.C.
Summer interim program. (Fletcher Affidavit, at 5).
Operation during this period is designed to minimize tube wear so that tube rupture, multiple tube rupture, actions of the PORV, and LOCA events are not relevant (Id.) All the evidence currently available shows that it is amply conserva-tive to operate plant employing Model D preheater-type steam generators at redu,ra power (Fletcher Affidavit, at 2)
(Quinton Affidavit, at 4). Thus, the various consequences of steam generators tube failure postulated by FUA as a
l l
i result of tube wear never arise because significant tube wear will be precluded.
For these reasons, as detailed in the attached f l
affidavits, the "significant" safety conccin alluded to in FUA's petition does not arise. (Fletcher Affidavit, at 2)
(Quinton Affidavit, at 6) Therefore, FUA's petition does not meet the standards of Vermont Yankee, supra 6 AEC at 523, for avoiding summary disposition on these pleadings.
Cost-Benefit Analysis. FUA in its "ew contention B2 questions the favorable cost-benefit analysis reached at the construction. permit phase. FUA assumes that the V.C. Summer plant will never operate at greater than 50% power and therefore "a more realistic 35% capacity factor" should be used in a new cost-benefit analysis (FUA Petition, at 6).
The Commission, on March 26, 1982, issued its final i rule on the need for power and alternative energy issues in operating license proceedings. 47 Fed. Reg. 12,940 (March 26, 1982). The rule effectively eliminates consideration of cost-benefit analyses based on power production at the operating license stage. The Commission has not diminished the importance of these issues at the construction permit stage, but has recognized that at the operating license stage the plant would be needed either to meet increased energy needs or replace older, less economical, generating
capacity and that no viable alternatives to the completed nuclear plant are likely to tip the NEPA cost-benefit balance against issuance of the operating license. Id.
Experience shows that completed nuclear power plants are used to their maximum availability and there has never been a finding in an NRC operating license proceeding that a viable environmentally superior alternative to operation of the nuclear facility exists. I d,. at 12,942. The purpose of the amendment is to avoid unnecessary consideration of icsues at the operating license stage that are not likely to tilt the cost-benefit balance. (Id.) Hence, 10 C.F.R. $
51.53(c) is amended to provide " Presiding officers shall not admit contentions proferred by any party concerning need for power or alternative energy sources for the prcposed plant in operating license hearings." Id. at 12, 943. We necessarily defer to the Staff for further development of the impact of the amended rule.
On the merits of FUA's proposed contention, we believe there is no change in the favorable cost-benefit analysis reached in the construction permit proceeding.
There is no basis to assume that the plant will operate at reduced power for its entire lifetime. Although we would not argue that operation at reduced power for as much as a !
is beyond possibility or that a period such as a year could be assumed, for purposes of discussion of the contention, it
I seems obvious that even one year at fifty percent power would not result in more than a minimal adjustment in total cutput over the lifetime of the facility. .
Sua Sponte Authority. As to the alternative relief l r
sought in the event intervention is denied, FUA would have the Board hear these matters under its sua sponte .
authority pursuant to 10.C.F.R. $ 2.760(a). As we have l stressed earlier in this response, there is no need for the ;
Board to consider these matters for the Commission and the NRC Staff are moving to resolve the problem. There is nothing for the Board to add or require which is not i
already being done.
The issue of the Licensing Board's sua sponte 1 authority was thoroughly addressed by the Commission in Texas Utilities Generating Co., et al. (Comanche Peak Steam L Electric Station, Units 1 and 2), CLI-81-36, slip op.
j (Dec. 30, 1981). The standard for exercise of sua sponte authority is analogous to that for reopening, i.e., that a !
significant safety, environmental, or common defense and i i
security matter remains. 10 C.F.R. $ 2.760a; see Consolidated l Edison Co. of New York (Indian Point Station, Unit 3 ), j CLI-74-28, 8 AEC 7, 9 (1974). In Comanche Peak, the Commission '
stated that "the apparent need to . . . monitor the Staff's {
progress in identifying and/or evaluating potential safety l e
i l i
l
I l
\
o or environmental issues are not factors which authorize a board to exercise its sua sponte authority." Comanche Peak, supra, December 30, 1981 slip op. at 3. Based on the authorities cited, we do not believe this to be a case in which the exercise of the Board's sua sponte authority is warranted or justified.
Conclusion For all the foregoing reasons, the Petitioner has failed to satisfy the five requirements for late intervention and has similarly failed to meet the burden required for reopening the record.
Accordingly, the Board should deny the Petitioner's motion in all respects.
Respectfully submitted, M
V iose B. Knotts, Jr.
C. Sanford Debevoise & Liberman 1200 Seventeenth Street, N.W.
Washington, D.C. 20036 Attorneys for Applicants Of Counsel:
Randolph R. Mahan General Attorney South Carolina. Electric &
Gas Company P.O. Box 764 Columbia, South Carolina 29218
. ~ .
E UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ,,,,
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of SOUTH CAROLINA ELECTRIC & GAS COMPANY Docket No. 50-395 (Virgil C. Summer Nuclear Station, Unit 1)
COUNTY OF ALLEGHENY STATE OF PENNSYLVANIA ss AFFIDAVIT OF W. D. FLETCHER My name is W. D. Fletcher. I am Manager, Steam Generator Development and Performance Engineering in the Nuclear Technology Division of the Westinghouse Electric Corporation. My business address is P.O. Box 855, Pittsburgh, Pennsylvania 15230. A statement of my educational, pro-fessional qualifications and experience is attached and forms a part of my affidavit. l I have reviewed the " Petition to Intervene and Request for Hearing" filed by Fairfield United Action ("FUA") dated April 9,1982 in this
]
proceeding. The purpose of this affidavit is to address the safety issues raised in the contention identified as "Bl" by FUA. I have reviewed the results of the data gathered from operating plants, 1
M. ~ ,,
_ _ _ _ _ . _ =__
5 I
utilizing Model D steam generators. In my opinion, the interim operation program developed by Westinghouse and South Carolina Electric & Gas Company ("SCE&G") for the V. C. Summer plant wi.11 preclude any significant steam generator tube wear until modi-fication to the steam generators can be implemented. Contention B1 does not present, in my opinion, significant safety issues. ;
I am familiar with data from the operating plants utilizing Model l D3 steam generators. Wear has been observed at plants utilizing :
Model D3 steam generators which have operated for extended i periods of time at high main feed flow. Significant tube motion i has been determined from instrumented steam generators and model ;
tests to begin at about 60-65% main feed flow. It is believed I that tube wear is related to significant tube motion. Significant [
wear is not expected below about 60-65% power level. As explained
{
in greater detail below, I have concluded that the interim ;
operating program for the Summer plant within the specified ;
parameters will minimize the potential for significant tube wear. l An interim operating program for the Summer Plant has been developed by Westinghouse and South Carolina Electric and Gas Company with the objective of precluding any significant steam !
generator tube wear until the permanent modification to the steam generators can be implemented. This interim program includes:
Instrumentation installed inside steam generator L Low power testing i Power escalation testing to main feed flow up to 50%
power Power operation with main feed flow at 50% power for specified time intervals (including operation above 50%
power based on infonnation available at that time)
Tubing eddy current inspections 2
l l
For present purposes, power level and main feed flow can be <
taken to be equal. The 50% main feed flow power conditions identified for the interim period have been based on instru ; .
mentation data obtained from three operating plants (having Model l D steam generators of similar design as installed at Summer) and hydraulic, vibration test model results, analysis of operating plant eddy current inspection data and tube inspections made on operating plants before and after 50 percent power operation.
During the interim operating period, it is understood that data from installed instrumentation will be examined and inspection of the steam generator tubing will be perfomed at appropriate intervals as a means of verifying that significant tubing degrada-tion has not occurred.
In D3 operating plants where tube wear was observed, instrumenta-tion was installed to measure tube motion, which has provided data relating motion to power level. This data supports the power level which can be used during the interim operating 1 period. Characterization of tube motion in the Model D-3 steam generators at operating conditions showed that significant motion begins to occur at about 60% to 65% feed flow through the main nozzle. It is concluded from these data that operation at 50%
feed flow through the main nozzle is an acceptable powe- level for the interim period to minimize the potential for significant tube wear.
A pertinent comparison of operating data which follows shows the operating history of three plants with similar steam generators, in tems of hours and power levels. The data are based upon the time of operation prior to tube inspection in which wear indications were measured.
COMPARATIVE POWER HISTORIES Number of Hours of Operation ;..
Plant 1 Plant 2 Plant 3
% Power (f;on-domestic) (Non-domestic) (Domestic)_
as of Oct. 1981 as of Nov. 1981 as of Feb. 1982
>90 to 100 1640 1456 72
>75 to 90 540 95 252
>50 to 75 1580 537 11761
>20 to 50 1390 738 (negligible)
Wear Observed Wear Observed No significant eddy current indications 1Most of this time period was at 50% power At Plant 1 and Plant 2, where wear indications were observed in the tubes, significant periods of operation were at high power levels, i.e., greater than 90%. Plant 3 has operated for about 55% of the time since initial startup in December,1981. At Plant 3, where the power level has been primarily at 50%, no significant wear was indicated. The judgment derived from analysis of this data is that no significant we3r is to be expected from interim time periods of operation at t'ie lower power levels.
Additionally, the results of other plant operating experience at 50% power, equivalent to the program for the Summer D-3 steam generators is relevant. AtPlant2(D-3steamgenerators)a program of 50% power operation for 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> (with short intervals of higher power operation to obtain tube instrumentation data) '
was completed in March of 1982. Comparing the tubing inspection data in the preheater region prior to (i.e.,in Nov. 1981) and after (i.e., in March 1982) this operating interval showed no significant change, including tubes which previously exhibited indications of wear.
4
Experimental testing has provided a data base which also supports the results observed from the tube instrumentation and plant i inspection data. The tube instrumentation data indicate that .the onset of tube motion is a result of turbulent feed flow at the preheater inlet at the higher equivalent power levels. The testing results have been obtained from several hydraulic models which demonstrate levels of turbulence and associated flow fluctuation at various feed flow rates. Data from a scale model (s4/10 scale) of the D-3 preheater demonstrates tube motion which increases with the feed flow rate in a manner similar to the data from the operating plant instrumentation. These test results ;
support the conclusions derived from the field data.
Westinghouse currently has an engineering design program to modify Model D steam generators to alleviate tube wear. The anticipated design modification can be implemented in the field.
An extensive testing program is in progress to verify that the anticipated modification to the steam generator preheater section addresses and corrects tube wear. The testing program incorporates a variety of testing facilities such as 0.417, 2/3 and full scale model testing. These test results, when combined with actual plant performance data, are expected to provide verification that the design modification will perform as predicted by engineering design analysis.
The other issues raised in FUA Contention B1 are not concerns during the V. C. Summer interim program as described above.
Operation during this period is designed to minimize tube wear so that tube rupture, multiple tube rupture, actions of the PORY and LOCA events are not relevant.
Notwithstanding this, a brief status of these issues is as follows:
5
Row 1 The staff analysis of tangent point cracking of Row 1 tubes U-bGnds concludes that while there remains a potential that this issue may occur in Model D Row 1, it is not a safety concern. ;:.
Westinghouse supports the technical basis for this conclusion as detailed in Section 5.4.2 of the V. C. Summer SER, Supplement 3, January, 1982.
Tube Westinghouse has performed analyses for the postulated double-Rupture ended steam generator tube rupture event for all Westinghouse-Events designed NSSS plants. Systems installed in these plants are designed to accommodate these events. The results of these -
analyses and of other analyses have been used to formulate Emergency Response Guidelines (ERG) for use by utilities in writing plant-specific Emergency Operating Procedures. West-inghouse has also perfonned evaluations of those significant tube leakage events that have occurred and has used the results of these evaluations to further improve the guidance provided in earlier versions of the Westinghouse ERG's.
Additional evaluations of these significant tube leakage events and the manner in which those leaks developed, leads Westinghouse to believe that multiple tube rupture is very unlikely. This belief is further reinforced by the results of steam generator tube inspections that are routinely performed in operating steam generators and also as described above for the Sumer interim operation program. These inspections are specifically aimed at early detection of any condition that would have the potential for tube rupture. In addition, tube conditions are continuously monitored for tube leakage during normal operation.
As noted above, Westinghouse believes that concurrent multiple tube ruptures or concurrent tube ruptures in multiple steam i 6
generators are very unlikely. Notwithstanding this belief, Westinghouse has performed, for the purposes of developing additional ERG's, evaluations of multiple tube ruptures and of, tube ruptures in multiple steam generators. These ERG's take into consideration past operating experience, post-TMI lessons learned and multiple equipment failure contingencies. These procedures have been written to minimize radioactive release from ,
the plant. This effort was sponsored by the Westinghouse Owners Group (WOG) of which South Carolina Electric and Gas'is a member.
The latest revision of these guidelines were submitted by the WOG k to the NRC in November 1981 and are currently under review. In a meeting with the ACRS on 3/24/82, NRC staff expressed basic agreement with the Westinghouse guidelines.
PORV With respect to a tube rupture, the PORV provides a means to depressurize the primary side of the steam generators. If the PORV were to stay in the open position during use, the operator could isolate the PORV with the PORV isolation (block) valves, such as used at both Ginna and Three Mile Island, to effectively isolate the PORVs. The presence of three PORVs and associated block valves in the Summer plant allows the operator sufficient flexibility to enable him to decrease and control RCS pressure.
The NRC has requested certain operational data for pressurizer PORV's through NUREG-0737, Item II.K.3.2. In responding to this request, Westinghouse conducted a survey of domestic Westinghouse NSSS operating plants, and the results of such survey were ,
provided to the NRC. Notwithstanding the experience in the past that some PORVs have in a few instances remained in the open position and did not close on demand, the results of the Westing-house survey and operating experience subsequent to the survey, still indicates that the PORV relhbility is acceptable.
7
l Moreover, the industry has implemented certain testing. For example, the Electric Power Research Institute (EPRI) has data from such tests that are directly applicable to the PORVs at ;:.
Virgil Summer, since EPRI tested the same exact model PORV as installed in the Virgil Summer nuclear facility [ Copes-Vulcan PORY Model D-100-160 (316 w/ Stellite plug and 17-4Ph cage)]. As reported in the "EPRI PWR Safety and Relief Valve Test Program, Safety and Relief Valve Test Report," (April 7, 1982), twenty tests were conducted on this valve during which the valve fully opened and fully closed on demand. .
Accident Postulated accident sequences, such as the loss-of-coolant Stqu:nces accident (LOCA) and the feedline break (FLB) accident, impose increased loads on the steam generator tubes. Such postulated increased loads have been evaluated by Westinghouse in an analy-i tical program. The results of these analyses show that tubes which exhibit degradation less than that specified in the tube plugging criteria will maintain their integrity for all postulated design basis accident sequences.
These studies have further shown that for a LOCA, the maximum steam generator tube stresses, caused by rarefaction waves, blowdown and vibrational forces, occur in the U-bend region of the steam generator. The tube stresses near the tubesheet and in the preheater region are lower. In addition, the pressure force mechanism during the postulated LOCA is in the direction of the potential for tube collapse rather than tube rupture since the primary side has depressurized. A postulated tube collapse has small potential for creating a leak path between the secondary side and primary side of the steam generator. ;
I f
8 l
i Westinghouse studies have shown that the postulated feedline break accident produces the highest steam generator tube stresses of any of the design basis accident sequences. During the ;-
postulated feedline break accident, the primary steam generator tube stresses result from the pressure differential across the tube. The calculated stresses resulting from the postulated feedline break accident are within acceptance limits for a tube with wear even more than that limit specified in the tube plugging criteria.
Postulating a steam generator tube rupture in conjunction with another design basis accident sequence constitutes a double failure. As such, the ECCS Acceptance Criteria in 10CFR50.46 and the requirements of 10CFR50 Appendix K do not require evaluation of such an event. Accordingly the evaluation models do not apply to such postulated double failures and evaluation of such a postulated double failure has been performed using modified codes and assumptions.
Westinghouse has performed calculations of a postulated loss-of-coolant accident ("LOCA"), with and without secondary to primary steam generator tube leakage, for a typical Westinghouse plant.
The calculation was performed for a double-ended cold leg break of the reactor coolant pipe using codes and assumptions modified to more realistically represent the system than is performed for design basis calculations. Contrary to FUA contention (page 4),
these calculations showed that adequate core cooling was main-tained. The results of these calculations were that less than a 10*F increase in peak fuel cladding temperature was calculated when a 250 gpm secondary to primary tube leak was modelled. This leak rate is consistent with a postulated double-ended break of a steam generator tube. Since the pressure force mechanism during 9
i a LOCA is postulated to occur in the direction of the potential !
for tube collapse and since wear patterns on tubes removed from two operating plants exist on primarily one side of the tube,sa.
I postulated double-ended tube break resulting from a LOCA is !
considered highly unlikely. Moreover, the data from the two ;
operating plants indicates that one tube tended to lead the rest by at least 10 percent of the tube wall thickness reduction, thus :
the participation by leakage of more than one tube is considered unlikely.
L s
i P
i t
Further, affiant sayeth not. .
P P
t 1
I i
10
t AFFIDAVIT l Before me, the undersigned authority, personally appeared W. D. Fletcher, who, being by me duly sworn according to law, deposes and says that he is ;
authorized to execute this Affidavit on behalf of Westinghouse Electric 1 t
Corporation (" Westinghouse") and that the averments of fact set forth in this affidavit are true and correct to the best of his knowledge, infor-mation, and belief:
/u' W. D. Fletcher, Manager Steam Generator Development :
and Performance Engineering Sworn to and subscribed before me this /f_ day of 8/tc2 t
1982.
' rult&& nMh /
Notary Public PauTit stomsna. NotAar Pusuc l
' settetVlut 8050 Auf4NENT COUNIT MV CON #IS$10N IIPitt3 MARCH 10,1986 Wember, Pennsylvania Association of iforeries i
l i
I
STATEMENT OF QUALIFICATIONS AND EXPERIENCE W. D. FLETCHER EXPERIENCE My name is W.D. Fletcher; I am presently Manager, Stean Generator Development and Performance Engineering in the Nuclear Technology Division of the Westinghouse Electric Corporation.
I graduated from Hardin-Simmons University in 1950 with a Bachelor's degree in Chemistry and from Fordham University in 1960 with a Masters degree in Chemistry.
I was employed with the Vitro Laboratories from 1951 to 1955, where I performed research on organo-phosphorus compound synthesis, reaction kinetics and mechanisms of organo-phosphorus compounds, i phase studies, bench scale and pilot plant production of organo-phosphites, high and low temperature kinetic studies of boron hydride synthesis, and. electro-kinetic studies of electrophoretic deposition of inorganic oxides in the manufacture of reactor fuel elements. !
r In 1957 I began my employment with Westinghouse and have been engaged in development work on the heterogeneous catalysis of- !
reactions between hydrogen and oxygen produced through radiolysis ,
l of reactor coolants, reaction kinetics and mechanisms, catalyst ,
development and evaluation in high temperature and pressure aqueous l solutions; evaluation and study of reactor coolant contaminants l
and means of coolant purification; study of behavior of fission and corrosion products in reactor coolants; in-pile studies of
- 1 ?
- reactor coolants as pertains to chemical shim technology; reactor plant chemistry control, analyses, and data collection and interpretation of all operating reactor systems designed by Westinghouse.
Since 1970, I have been directly involved in development and design activities related to Westinghouse steam generators.
Under my direction, steam generator programs related to operations have been executed. involving chemistry and materials as well as specific design configurations.
As Manager, Steam Generator Development and Performance Engineer-ing, I am responsible for three design-development groups that involve steam generator thermal / hydraulics, advanced concepts design and analysis and design of field modification to steam generators, t l
I am a member of the American Chemical Society, the National Association of Corrosion Engineers, the American Nuclear Society, and the American Society of Mechanical Engineers.
I l
I I
l
PUBLICATIONS
" Update of Operations with Westinghouse Steam Generators" American Nuclear Society 1977, D.C. Malinowski and W.D..,. Fletcher.
" Operating Experience with Westinghouse Steam Generators",
Nuclear Technology, 1975 W.D. Fletcher and D.D. Malinowski.
" Water Technology for Nuclear Power /PWR's", Industrial Water Engineering, 1971, W.D. Fletcher.
" Primary Coolant Chemistry of PWR's", W.D. Fletcher, the International Water Conference of the Engineers Society of Western Pennsylvania, Pittsburgh, October,.1970.
" Post Accident Iodine Cleanup by Containment Filters and Sprays."
Presentation at Tampa, Florida, May 21, 1968, J.D. McAdoo and
. W.D. Fletcher.
" Effects of Coolant Chemistry on Corrosion.and Corrosion Products",
W.D. Fletcher, Am. Nuc. Soc., Seattle, June 1969.
EURAEC-1972 (NCAP-3690'-41 " Description and Evaluation of the Boron Concentration Meter Utilized at the SENA (Franco-Belge)
Reactor Plant", January 1968, W.D. Fletcher.
WCAP-3269-57 "The Post-Irradiation Examination of Saxton Fuel Cladding Corrosion Products", March 1966, L.F. Picone and W.D. Fletcher.
WCAP-3269-63 " Fission Products from Fuel Defect Test at l
Saxton", April 1966, W.D. Fletcher and L.F. Picone.
- a
- WCAP-2964 " Stability of Alkali in Reactor Coolant", 1964, ;
W,D. Fletcher.
WCAP-2656 " Analysis of Fission Products in Saxton Primary Coolant", August 1964, W.D. Fletcher, t
" Water Technology of the Saxton Nuclear Experiment", Division of Water and Waste Chemistry, 4, 46 (19 6 4 ) , W. D. Fletcher and
- R.F. Swift. * !
" Flame Photometric Determination of Lithium Produced by B-10 (n , a) Li-7 to Measure Boron-10 Burnup in Reactors Utilizing Chemical Shim Control,: Presentation at Gatlinburg, Tenn.,
i Oct. 6-8 1964, B.D. LaMont and W.D. Fletcher.
- WCAP-3716 " Ion Exchange in Boric Acid Solutions with Radio-
)
i active Decay", November 1962, W.D. Fletcher.
WCAP-1689 Rev. "The Behavior of Stainless Steel Corrosion Products in High Temperature Boric Acid Solutions", May 1961, W.D. Fletcher, A. Krieg and P. Cohen.
WCAP-4097 '" Inorganic-I6n-Exchanger Materials for Water Puri-fication in CVTR", August 1961 GCVNA-135), N. Michael, W.D. Fletcher, et al..
l
'~ .
t WCAP-3730 " Interactions Between Stainless Steel Corrosion ;
l
- Products and Boric Acid Solutions", March 1960, W.D. Fletcher, j
- )
"Some Performance Characteristics of Zirconium Phosphate and Zirconium Oxide Ion Exchange Materials", Trans. Am Nuc Soc.,
3, 46 (1960) , N. Michael and W.D. Fletcher, i
WCAP-1206 " Internal Recombination Catalyst Studies",
May 4, 1959, W. D. Fletcher and D.E. Byrnes.
WCAP-ll10 "A Semi-Flow System for the Study of Catalytic !
Combination of Hydrogen and Oxygen in Aqueous or Slurry System", l February 1959, W.D. Fletcher.and W.E. Foster.
l
" Electrophoretic Deposition of Metallic and Composite Coatings",
Plating 42, 1255 (1955). -
" Post LOCA Hydrogen Generation in PWR Containments", W.D. Fletcher, !
M.J. Bell, R.T. Marchese, and J.L. Gallagher, American Nuclear Society. ,
i PATENTS ;
U.S. Patent, "Information Storage Systems and Methods for l Producing Same",
U.S. Patent, " Boron Concentration Meter". !
U.S. Patent, " Electrophoretic Coating Dispersion Formulations".
i l -
a i
= ;
l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l i
In the Matter of )
)
SOUTH CAROLINA ELECTRIC & )
GAS COMPANY, et al. ) Docket No. 50-395 OL
)
(Virgil C. Smnmer Nuclear ) l Station) ) l AFFIDAVIT OF MICHAEL D. QUINTON REGARDING THE V. C. SUMMER NUCLEAR STATION STEAM GENERATORS Michael D. Quinton, being first duly sworn according to i law comes forward and states:
My name is Michael D. Quinton. I am employed by South i
Carolina Electric & Gas Company as Director of Mechanical j Engineering within the Nuclear Engineering Department. In this position I have been assigned specific responsibility j to monitor technical developments relative to steam generator tube wear problems associated with Westinghouse Model D !
steam generators and particularly Model D-3, which is the model of steam generators installed in the Summer Station !
three loop unit. I possess a degree in mechanical engineering, received from the University of South Carolina in 1973. I am a registered professional engineer in the State of South l T
Carolina and I have been involved in the operation, design and construction of nuclear power plants for over 12 years. ,
I (Before entering college I was a qualified reactor operator / reactor technician in the U. S. Navy.)
&Dy b Y th_-.
-e -
I havo read the "Patition to Intervono and Request for Hearings" dated April 9, 1982 and filed by Fairfield United Action in this docket. The purpose of my Affidavit is to assess the significance of safety questions which FUA seeks to raise. I have concluded the issues raised by them 15' their petition do not present a significant risk to the health and safety of the public. I will describe the .
Applicants' in~olvement v with this problem and with utility and vendor efforts to identify the cause and develop an I
effective remedy. I shall also describe Applicants' commit-ments relative to interim operation of the Virgil C. Summer Nuclear Station and evaluation of proposed modifications i leading to installation of an effective and approved "fix." [
Through industry publications, information from Westing- f house, and information exchange systems such as NUCLEAR f NOTEPAD, we became aware of Westinghouse designed steam i generator operating problems at Ringhals (Sweden) and i'
Almaraz (Spain) (which both utilize Model D3 steam generators).
However, it was not until late December, 1981 that we were advised that these problems had direct application to the !
Summer Plant (also having Model D3 steam generators). ,
i Upon being so informed, we initiated an information gathering and evaluation effort involving frequent contact :
with Westinghouse as well as other utilities similarly !
affected to discuss the nature of the mechanisms observed, interim protective measures, and the possible remedies.
By letter dated January 20, 1982 the Director of the Division of Licensing of NRC's Office of Nuclear Reactor Regulation (NRR) requested that we provide.them with our
)
plans to address the problem as it relates to the Summer Plant. By a letter of the same date, the NRC Staff gave notification to the ASLB of the problem.
By letter dated February 19, 1982, Applicants responded to the request for our position on the matter. This letter was sent to all parties in this proceeding. In that letter, we committed to an interim operating program for the Summer Plant with the objective of precluding any significant steam generator tube wear pending permanent modification. Our response outlined our plans to proceed with initial core loading, low power testing, and escalation up to and con-tinued operation at the 50% level (or a higher level if justified by the information available at the time). Addi-tionally, we committed to shut down after approximately two months of operation at the escalated level and inspect (eddy current testing) tube rows 47, 48, and 49 for indications of tube wear. At the time of that response, there were no plans for providing internal or external instrumentation as installed at some of the other plants with similar problems.
However, since that time we have committed to install internal instrumentation in two tubes in one steam generator.
This commitment is contained in a Nichols to Denton letter dated April 14, 1982, which has been sent to all parties.
Westinghouse has indicated that a modification to correct the tube wear problem may be available by late summer, 1982.
It is estimated that implementation of the modification in all three (31 steam generators will take approximately three to four months.
It is my opinion that the' operation of the V.C. Summer Nuclear Station Unit No. 1 under the conditions set forth in the Nichols to Denton letter of February 19, 1982 as supple-mented by the instrumentation commitment in the Nichols to Denton letter of April 14, 1982 presents no significant risk to the health and safety of the public. This is based upon
'the results to date of the Westinghouse analysis and test programs at the two operating foreign reactors and Duke Power Company's McGuire unit as well as other Westinghouse studies. (See Fletcher Affidavit.) In addition to health and safety considerations, which are of primary concern, the economic incentive for Applicants to avoid operations at power levels posing any significant risk of steam generator damage with the cost penalties attendant to such damage provides every reason for us to adhere to the monitoring and testing programs outlined in our February 19, 1982 letter.
To address FUA's proposed Contentions (B1 and B2) specifically, I note first that an underlying premise in those contentions is that the Summer unit will be operated at a power level at which flow-induced vibrations in the preheater region will act to cause tube wear. This premise is incorrect.
The Applicant,s' commitment is to limit operation of the unit to 50% power (or to an appropriate level of power above 50%1 which precludes significant tube damage.
i
The mechanism for the inducement of, tube wear in the Model D steam generator cited by FUA is in agreement with our current information on the subject. FUA has referenced the Chesnut to Youngblood Memorandum statement that the increased turbulence is experienced at feed flow rates of approximately 50% in Model D3 steam generators. Our commit-ment to limit operation to 50% at this time is consistent with the NRC memorandum. Based on available information, operation at this level precludes the tube wear problem.
Since the accelerated tube wear problem will not arise during interim operation (or after a permanent modification is made to preclude the problem), there is no basis to postulate, as FUA does, tube rupture.or multiple tube rupture, possibly in combination with PORV failure or possibly leading to LOCA events, as a consequence of accelerated tube wear. Nevertheless, a few comments are in order on those matters to correct the impression that might otherwise be left by FUA's statements.
Westinghouse has performed analyses for postulated double :
end steam generator tube rupture events for all Westinghouse designed. nuclear steam supply system plants. (See Fletcher Affidavit, page 6.)' While FUA properly points out that the FSAR (5.2-161 gives the design basis tube failure as a double ended rupture of single tube, it is also true that this accident will ystult in a transient which is no more severe than that associated with a reactor trip from full power and thus requires no special treatment insofar as fatigue evaluation is concerned.
(Id.)
FUA's discussion of potential problems with the Power I
4
Operated R311ef Valva (PORV) on the pressurizor is misplaced.
The PORVs at the Sur.mer Plant are of the specific model tested by the Electric Power Research Institute (EPRI). The results of those tests conclusively demonstrate the PORV's operability ("EPRI PWR Safety and Relief Valve Test Program, Safety and Relief Valve Test Report, April 7, 1982). The
" anomalies" in safety valve (not PORV) performance referred to in the Nichols to Denton letter are in no way related to PORV operability.
Applicants are well aware of the potential for premature tube wear problems in Westinghouse Model D steam generators. We recognize.the need for corrective action and are working with the vendor and the NRC to develop, verify and install a suitable modi-fication to eliminate the problem. Based upon our understanding ,
of the mechanisms involved, we have developed and committed to an interim operating program to insure significant steam generator tube wear does not occur. By virtue of our actions in this matter, operation of the V.C. Summer Nuclear Station does not present undue risk to the public health and safety.
- M ;
Michael D. Quinton' ,
i SWORN to before this 23 rd da'y of April , 1982 H/d a . %' s ut< (L.S.)
Notary Public for South Carolina My Commission expires: /0 s ? W
' ; e '
- f
- ATTACHMENT A I
$g"% *% o gK l
February 18, 1982 3 .j SECY-82-72
! t 8
%, . . . . f .
POLICY ISSUE (Information)
For: The Commissioners From: William J. Dircits, Executive Director for Operations
Subject:
OVERALL STEAM GENERATOR PROGRAM Puroose: To transmit the enclosed report, " Steam Generator. Status Report."
Discussion: As a result of concerns expressed by the Chairman regarding the overall program addressing steam generator problems, I have requested the staff to review NRC and industry programs related to this subject and to develop a comprehensive, integrated program for addressing the issue. Furthermore, I have emphasized the need for and asked the staff to consider joint NRC/ industry programs that will promote resolution of the problem. I have assigned the lead responsibility to RES to :
work with NRR to pull together a plan of action. (See February 17, 1982 memorandum to Robert B. Minogue from William 'J. Dircks,
Subject:
Overall Steam Gene'rator Program.)
As a starting peint, RES, with input from NRR and I&E, has prepared the enclosed report, " Steam Generator Status Report."
The report sumarizes steam generator degradation experience and d.Is safety and economic significance; discusses the current regtlatory approach for ensuring safe steam generator. operation; describes current corrective actions and their limitations; sumarizes NRC, industry..and foreign research and development ,
activities; and discusses the long-term approach for addressing steam generator problems. The report is intended to provide a
" snapshot" of the steam generator issue as it exists today and .
to serve as a background and foundation for development of the proposed integrated program. The staff is currently working to develop such a program and will discuss its feasibility and fann in a later report.
.' O kWillia
/ lf %l. . Dircks .
Executive Director for Operations
Enclosures:
- 1. Dircks' Memo 12s >S:n.v.s vzd ' 1 9 A9 39 C9 E9NSU
- 2. status Report us m PA .
[ddM*M 2
Contact:
oc 20555 aAwtindruN 41 86g$ k443-5903 Strosnider, RES i
&p* C th g UNITED STATES e o NUCLEAR REGULATORY COMMISSION
{ E WAsHINGTO N, D. C. 20555 .
t, b .,o . ' February 17, 1982 i
MEMORANDUM FOR: Robert B. Minogue, Director Office of Nuclear Regulatory Research t
FROM: William J. Dircks Executive Director for Operations
SUBJECT:
OVERALL STEAM GENERATOR PROGRAM As you know, as a result of concerns raised by the Chairman associated with the overall program to address the steam generator problems, I discussed this issue with the various Office Directors involved. As a result, I am assigning you the lead responsibility working jointly with NRR to pull together a plan of action to address the steam generator concerns. You should consult with and get input from other offices, particularly IE, as necessary.
I envision the developnent of two papers for the Commission's information and consideration:
- 1. The first paper, to be developed in a short time frame, should address principally the question of "What is going on now?", and "What has happened to date?" Tha general framework of this paper would include: ,
- a. A discussion of the problems (both regulatory and economic) and their safety significance; -
- b. What programs are now going on within the NRC, DOE, USN, industry (including EPRI and AIF), and international programs. The NRC efforts include the NRR efforts on Unresolved Safety Issu'es
'A-3, 4, 5, as well as ongoing RES efforts;
~
- c. The short term fixes that have been proposed and/or implemented for individual problems over' the years and some insight into
. difficulties;
- d. Staff recommendations on the short term actions (for all affected reactors) believed to be prudent to ensure that public health and i
safety are protected; and *
- e. A general idea of possible direction to attack the problem in the long term.
l l 2. Recognizing that current efforts (NRC and regulated industry) are l
probably not nearly enough, a second paper, again working jointly with NRR, should be developed. This would include what additionally' k D { OMD-
~
r.,
STEAM GENERATOR STATUS REPORT '
FEBRUARY 1982 e
e U.S. NUCLEAR REGULATORY COMMISSION Q$\0WS
.l'., i i
l i
t must be done, and a plan of action for a joint research effort to provide a definitive resolution. To assure a creditable product, close coordination with industry will be r'equired. Perhaps a lead group representing the industry, such as EPRI, can, together :
with NRC staff, ACRS, and other consultants as necessary, form a steering group to identify what must be done, who should do what to assure canplete coverage without overlap and the priorities of work - to include the inspection program. l I would like you to report to me at the weekly meetings regarding progress, problems, or action I can take to move this program along in an expeditious [
manner.
+
. .i '
i '
/
db
William J. Dircks l Executive Director ;
for Operations CC: Chairman Palladino a- Commissioner Gilinsky Commissioner'Bradford Commissioner Ahearne Commissioner Roberts H.R. Denton/NRR R.C. DeYoung/IE l
l
I. Problem Defintion A. Summary of Tube Degradation ,
Degradation of steam generators (SG) manufactured by each of the three pressurized water reactor (PWR) vendors has resulted due to a combination of steam generator mechanical design, themal hydraulics, materials selection, fabrication techniques, and secondary system design :
and operation. In the early and mid-1970s, Westinghouse (W) S.G. experienced caustic stress corrosion cracking, and W and Combustion Engineering (CE) :
S.G.s experienced tube thinning (wastage). These modes of degradation were due to difficulties encountered with phosphate secondey water ;
chemistry. Because of these difficulties, most W and all CE plants r converted to an all volatile (AVT) secondary water treatment. Although ;
this conversion greatly reduced.the occurence of stress corrosion cracking l and wastage, other degradation modes including denting (defomation of !
the S.G. tubes due to corrosion of the carbon steel support plates) l
. began to occur. l r
Babcock and Wilcox S.G., which have a significantly different !
design from W or CE and have operated exclusively with AVT water chemistry, had relatively good operating experience in their early years of operation. i Nevertheless, th.ey have experienced numerous tube leaks. The principal !
modes of degradation in B&W units have been fatigue crack growth, confined !
primarily to limited sets of tubes located on the open inspection lane, l and more recently erosion-corrosion and primary side intergranular a ttack. '
To date, many different foms of steam generator degradation have been identified including: stress. corrosion cracking, wastage, intergranular attack, denting, erosion-corrosion, fatigue cracking, '
', pitting, fretting, and support plate degradation. One or more of these foms of degradation have affected at least 40 operating PWRs and have resulted in extensive S.G. inspections, tube plugging, repair, or replacement.
Recently, foreign Westinghouse S.G.s of the same design as McGui.re have experienced tube wear associated with flow induced vibration due.to a ;
new integral preheater design. References 1, 2, and 3 present detailed discussions of domestic S.G. operating experience.
The economic impact of steam generator degradation has been significant. Approximately 23% of non-refueling outage time has been attributed to steam generator degradation. The cost of such outages in tems of replacement power alone is very high. However, perhaps the greatest financial costs incurred to date are those associated with steam generator replacepent. Replacement of the Surry Unit 1 and Unit -
2 S.G.s cost approximately $200 million, including cost of makeup power.
Replacement of the Turkey Point S.G.s, currently in progress, will cost an estimated $460 million. NRC staff time involved with these activities is estimated at 6000 manhours for Turkey Point (which included time for a hearing) and 2000 manhours for Surry. Less radical operations also incur significant costs. Recent tube sleeving operations at San Onofre '
involved repair of approximately 7000 degraded tubes at a cost of $70 million. Proposed sleeving of 3000 tubes at R E. Ginna has an estimated cost of $20 million.
B. Safety Significance i The safety significance of S.G. tube integrity can be divided
1 into three categories: tube failures under nomal operating conditions; j tube failures concurrent with postulated accident conditions; and personnel exposure associated with S.G. inservice inspection (ISI), repair, and replacement. ,
The majority of the S.G. tube failures that have occurred under nonnal operating conditions were small stable leaks sometimes requiring plant shutdown, inspection, and corrective actions, but for the most part small enough (e.g., below technical specification leak rate limit) that operations continued until a scheduled shutdown.
However, four significant S.G. tube ruptures have occurred in domestic PWRs since 1975. These events occurred on February 26, 1975, at Point Beach Unit 1. Septenber 15, 1976, at Surry Unit 2, October 2,1979, at .
Prairie Island Unit 1 and on January 25', 1982, at R. E. Ginna. The first three of these events were evaluated in NUREG-0651, " Evaluation of Steam Generator Tube Rupture Events." The report includes an evaluation .
of ' system response, operator action, and radiclogical consequences during the three events. The leak rate associated with these events ranged from about 80 gpm to 390 gpm. The conclusion of the report is that no significant offsite doses or systems inadequacies occurred during the tube rupture events analyzed. However, the potential for more significant- consequences was recognized and a number of procedural recommendations were made to correct the deficiencies that were noted.
The present disposition of each of the recommendations is discussed in a recent memo to Commissioner Bradford from W. Dircks (Ref. 4). The present design basis for assuring that plants are acceptably protected
- against S.G. tube rupture events is a postulated double-ended rupture of a single S.G. tube. This assumption is intended' to provide a bounding.
leak rate for 'a spectrum of rupture geometries in a single tube and a
. spectrum of smaller leaks in multiple tubes within a single S.G. The consequences of multiple tube failures, in excess o.f the design base, have not yet been rigorously studied. Rapid degradation between inspections of a large number of tubes could create th potential for multiple tube failures in the event of a plant transient or failure of a single tube and the accompanying jet impingement and tube whip could cause failure of additional tubes. Furthennore, the potential for complicating circumstances involving multiple equipment failures such a's the stuck open PORV during the Ginna incident and possible steam bubble fonnation in the primary
. system'have not been evaluated. Another concern is ruptures in multiple S.G.s. In this event, unless the plant can be rapidly depressurized and brought onto Residual Heat Removal, there is the potential to continuously lose emergency core cooling water outside of containment. The above n concerns are being addressed as part of the TMI Action Plan. Item.I.C.1 .
l in the TNI Action Plan addresses S.G. tube failures coupled with other failures (such as a stuck open safety relief ' valve in the secondary system), ruptures of multiple tubes, and simultan'eaus ruptures in mul,tiple ~
S.G.s. The purpose of this effort is not to expand the plant design basis but to assure that operator emergency procedures provide proper guidance for safely controlling the plant during these types of events.
Although rigorous analyses of many of the scenarios postulated above have not been completed, ISI, leak rate limits, and tube plugging requiiement's are intended to guard against such occurrences -(3ee Section II). In summary, the consequences of S.G. tube ruptures under nonnal operating conditions have been small; however, such events .can present a significant challenge to plant operators and safety systems.
, 9 ,.
l' l
During postulated accident conditions, such as main steam line
! break (MSLB), feedwater line break, or LOCA, the S.G. tubes are subject to increased pressure differentials and possible pressure waves (e.g.,-
subcooled decompression phenonena) and vibrational loadings. These loads increase the potential for failure of degraded S.G. tubes which could exacerbate the accident sequence.. In the event of MSLB, failed
- S.G. tubes would provide a leakage path ficm the primary to secondary
,. system and several potential leak paths for radioactivity to the environment would then exist. In the event of a LOCA, the core reflood rate could be retarded by steam binding. This phenomenon is associated with a cold leg break, in which reflood of the core requires displacing steam generated in the core through the hot leg, the affected steam generator, and out of the cold leg break. S.G. tube failures would create a secondary to primary leak path which aggravates the steam binding effect and could lead to ineffective reflooding of the core. Analytical and experimental evaluations of this phenomenon are contained in References 4 and 5.
Large MSLBs and LOCAs are considered extremely low probability events, but are postulated as bounding conditions. More realistic events might include small and intemediate size MSLBs or LOCAs. Although these postulated accidents pose a less severe challenge to S.G. tube integrity, tube rupture (s) leading to or following such events could have serious consequences. This is particularly true if fuel damage has occurred as in the case of Three Mile Island.
The final area of concern is the radiation exposure of personnel involved in S.G. inspection, repair, and replacement. Reference 3 presents a summary of data on S.G. related personnel exposure for selected plants from 1974 to 1980. In recent years, as much as 25% of some plants annual occupational exposure has resulted from routine S.G.
inspection and maintenance and as high as 60% for S.G. replacement.
Recent tube sle.cving operations at San Onofre incurred 3500 man rem exposure and similar operations are planned for other plants.
4 II. Regulatory Approach -
The NRC approach to assuring S.G. tube integrity under all
~
operating conditions is based on inservice inspection (ISI), primary to secondary leakage rate limits, and preventive tube pluggirrg requirements.
Guidance for perfoming ISI is provided in R.G. l.83, " Inservice Inspection of S.G. Tubes," and plant technical specifications include requirements
> for ISI. Typical plant specifications require periodic inspections of i
, 3% of the S.G. tubes in,the plant and augmented ISI in the event tube degradation is detected. Required frequency of inspection is generally flexible enough to allow inspections to be perfomed concurrent with -
refueling outages. Certain incidents such as tube leakage require '
unscheduled ISIS. Furthemore, many plants with extensive degradation problems have licensing amendments imposing higher frequency and larger size inspections. The ISI requirements vere developed largely through a canbination of engineering judgement and operating experience. More rigorous statistically based ISI programs have.been developed ts part of Unresolved Safety Issues A-3, A-4, and A-5 (see Section V). The purpose of the required ISIS is to detemine if tube degradation is occurring in the S.G., assess the rate of tube degradation based on results of successive inspections, and identify those tubes requiring plugging or repair.
,. a . .
0 . g Primary to secondary leak rate limits are an extremely important requirement for ensuring safe s.G. operation. Scme foms of tube degradation
- have been observed to degrade tubes beyond the prescribed plugging limit
, during the interval between inspections. Technical Specification primary to secondary leak rate limits requiring shutdown, ISI, and corrective actions provide protection against unacceptable levels of degradation between inspections. Many serious conditions of tube degradation have been detected by monitoring of primary to secondary leakage and subsequent inspection. Primary to secondary leak rate limits exist in each plant's techni~ cal specifications. The bases for these limits are twofold.
First, the leak rate limit ensures that the calculated dosage contribution from tube leakage will be limited to a small fraction of the allowable limits in the event of a'S.G. tube rupture or MSLB. Second, the leak .
rate limit is intended to correspond to a defect size that would not be expected to result in tube rupture under normal or postulated accident conditions.
Finally, decradation limits for tube pluccino exist in the plant Technical Specifications. Criteria for establishing the tube plugging limits are presented in R.G.1.121, " Basis for Plugging Degraded Pressurized Water Reactor Steam Generator Tubes." These criteria require that the plugging limit include margins for eddy current testing error .
and continued degradation between inspections. Thus, it is important to have a good estimate of the rate of degradation based on successive ISI results and an understanding of the degradation phenomena.
The primary focus of the current NRC philosophy is directed at
, maintaining primary system integrity. This is accomplished primarily . ,
through the requirenents described' above for ISI, leak rate monitoring, ;
and tube plugging. In a sense, it is directed at treating the symptoms
~
and not the cause of S.G. degradation, which lies primarily in secondary systen design and operations. This philosophy has been debated extensively, but the current position rega'rds eliminating the problem at its source as an industry responsibility. '
III. Current Corrective Actions ;
l An effective solution to S.G. tube degradation problems would require major changes in S.G. mechanical design, themal-hydraulics, materials selection, fabrication -techniques, and changes in the secondary l
~ system design and operation. Elimination of S.G. degradation requires a l systems approach integrating all of these considerations. There are no '
simple corrective actioris. This is particularly true for those plants l which have significant operating time and have experienced S.G. degradation.
Design changes in operating S.G.s that would be necessary to eliminate i degradation problems are virtually impossible. For example, tube to !
tubesheet crevices already contaminated with corrosive environments are ,
virtually impossible to clean, carbon steel support plates cannot be !
replaced with more corrosion resistant materials, and residual fabrication
~
i stresses cannot be remcved. Thus, corrective actions may prolong S.G. I life, but tube degradation is expected to contihue in operating plants.
Once the secondary systen is contaminated by an aggressive environment it is difficult to reverse the adverse affects. For example, caustic !
stress ' corrosion cracking and wastage, due to residual phosphate' water chemistry conditions, still continue in some plants long after conversion
_ . xmamm m _-
are in use. Th2se _ fixes include such actions as tube sleeving, sludge
' lancing, soaking and flu'shing, reduced operating temperatures to slow corrosion, boric acid injection'to arrest denting, support plate modifications
- q to retard denting, S.G. replacenent, and improvements in secondary A systen design and operation. Secondary system improvements include - "
Tr prapt correction of condenser in-leakage, condenser retubing, removal of copper based alloys from the secondary, system, and addition of ~demineralizin!
systems. An industry cons.tituted secondary water chemistry guidelines cmmittee, under chaimanship of EPRI, is developing generic chemistry limits and operating ~ guidelines. NRR.has been in contact With this committee for the past yea'r and will review a copy of the draft ~ reports prior to issue. Chenical cleaning has also been proposed but has not" .
been implemented due to uncertainties' regarding its longer-tem affect on S.G. integrity. Industry efforts :are currently underway to' eliminate these uncertainties and chemical cleaning may become a viable option in the near future.- These fixes have. met with varying degrees-of successp
- ' but nonef of then is a' panacea'. ~ Furthennore', short -term solutions to one '
probleni may. create other problems. Conversion'from phcsphate to. AVT-water chemistry, which' minimized wastage and stress corrosion cracking d-but was followed by denting, is a case in point.
~
~
.s
~~
Finally it should 'be noted tha't the majority of the plants under review for' operating licenses have S.G.s of similar design to those curre'ntly in operation, so that the potential for S.G. tube degradation -
exists in these plants ~as well.- .
. IV. NRC, Industry, and Foreign' Research 'and Development Activities .
NRC's steam generator research crocram addresses improved eddy e current inspection techniques for steam generator tubing, stress corrosio'n jt cracking'of steam generator tubing.and evaluation of tube integrity. .
v: - -
. The objective of the eddy-current program is to upgrade and ~ "
in prove eddy-current inspection probes, techniques ano associated instrumentatio for inservice inspection of steam generator tubing to improve the ability to identify and characterize tube defects. Specific objectives include improving defect detection and characterization as affected by tube .
diame'ter and thickness variations, tube denting, probe wobble, tubesheet and tube support interference, and defect location and . type. .
- n. ' .
. t , . . , - .
,~ The stress corrosionJer~acking ' program is developing data and models which will be used to predict the stress corrosion cracking .
. initiation and seWice' life of Inconel 600 ' steam generator tubing. The testing program includes variables which influence ' stress corrosion -
cracking such as temperature, stress, strain and strain rate, metallurgical '
structure,s and processing, and ingredients in the primary and secondary c ool a nt.. .
i "A steam generator, with service induced degradation will be .
used fo'r the validation of the accuracy and co.ofidence limits of nondestructive inspection instrumentation and techniques; buist and collapse tests on .
field degraded tubes to validate tube integrity models; and for developing data .for validation of previously developed' stress corrosion cracking
, predictive models, chemical cleaning and decontamination, dos'-rate e
! reduction and secondary side characterization. In addition, statistically .
l based sampling models for insenice ins'pection programs will be confimed
! and/or improved utilizing the first ever confimed data base.
.( . ( ,7 I There are many ongoing programs addr'essing' S.G. issues a -
EPRI, most of which are sponsored by the S.G. Owner's Group, and t..
rest by. EPRI itself. The programs address the following areas:
chemistry and corrosion, (2) materials selection and testing, (3)(1) themai hydraulic and structural testing and analysis, and (4) nondestructive examination (HDE). Efforts in the chenis.try and corrosion area are directed at examining the causes of corrosion.related degradation such as denting, intergranular attack, and stress corrosion cracking, and identifying po'tential fixes such as alternative sec.ondary water chemistry treatments. Materials selection and testing efforts are directed at -
characterizing.and evaluating.the suitability of alternative tubing and-S.G. materials. .This includes consideration of new heat treatments for..-
tubing and ccrnpatability of S.G: tubing with structural materials. ,
Testing and~ analysis.in.thennal hydraulics and structures is directed at.
secondary side S.G. design and perfomance and their effect on.S.G. tube infegrity..The;EPRI nondestructive examination programs focus on development of improved inspection techn.iques. .These techniques include multiple f requency/multiparameter eddy current- testing, automatic eddy current -
signal analysis, profilometry for quantifying dent configuration and strai'n levels in dented tubes, .and methods for evaluating the condition of the. tube support plates.- In addition, EPRI has established the HDE Center in Charlotte, NC, dedicated to providing good NDE techniques, and
, effectively transferring research and development results to' the industry.
Research and development activities underway on steam generators e outside the USA' are being funded at high levels in several countries.
The Japanese are conducting a very large program'.with emphasis on themal/
hydraulics, and' also on water chemistry and tube testing. To date, we' have received 1ittle infomation on the progress or results of their programs. The French have work underway on eddy current NDE, crevice
- i. chenistry, and decontamination. There is work underway in Sweden on
? water chemistry. The Gemans have wrk underway in eddy current HDE, .
and at KWU on primi.ry side decontamination and secondary side cleaning; however, Geman steam generators are tubed with Incolloy 800 so much of
.their research is less relevant to ours. Finally, the Italiant have underway a large program which will allow then to make new designs to avoid current and possible. future problems.
V. Long Tenn Approachl
.m . i :-
A. ' Unreso[ved Safety I'ssu s A-3, A-4, and A-5 Reg r' ding s
Steam Generator Tube Integrity .-
s . .. .. . , .
JIn 19 8 the'NRC established Unresolved Safety Issues A-3, A- 1 j 4, and A-5 (USI) regarding degradation in W, CE, and B&W steam generators,
, respectively. A draft report, NUREG-0844, presenting the proposed NRC - ,
staff. resolution of these generic safety issues has been prepared and is currently being reviewed by NRR management prior to transmittal to the Committee for Review of Generic Requirenents and the Cmmission and publication for public ccrnment. The report integrates technical studies in the areas of systems analyses, inservice inspection (ISI), and tube integrity to establish improved criteria for e'nsuring adequate tube integrity and safe steam generator operation under all conditions.
We
l 7
._..s r
1 i
In the systens ' analyses portion o'f the report, the consequences of steam generator tube failures during nonnal operation and postulated loss-of-coolant and main steam line break accidents are evaluated. The evaluation considers predicted ' fuel behavior, emergency core cooling ;
systen perfonnance, radiological consequences, and containment response. l The results of the systems analyses lead to proposed criteria for establishin3 .
a tolerable level of steam generator leakage during postulated accidents. I ISI techniques are then evaluated, and statistically based ISI programs presented which, if implemented, would provide additional assurance that ,
no more than the tolerable level of tube leakage, defined by the systems !
analyses, would occur during nonnal or postulated accident conditions.
In the tube integrity portion of the report, the behavior of degraded tubes during nonnal and postulated accident conditions and tube plugging criteria are evaluated. Proposed changes in operating procedures ;
and design changes 'to minimize tube degradation are also identified. l
' ' Implementation of the proposed requirements and criteria developed in the program for resolution of the USI are not expected to i totally eliminate S.G. degradaiion. The intent of the proposed requirements is to establish a logical approach to evaluating steam generator tube ,
integrity and ensuring safe steam generator operation. The draft NUREG- .
0844 recommends criteria and requirements that can be used to evaluate i current and future degradation programs in steam generators. The establishment of maximum allowable steam generator tube leak rates during postulated accident conditions and associated tolerable number of defective tubes i is a major contribution to the evaluation of steam generator tube degradation problems. It provides objective criteria against which steam generator i tube integrity can be evaluated. Similarly, the development of statistical t
ISI programs provides a rational, scientific basis that can be used to establish and evaluate ~ISI reouirenents that will ensure the above criteria are satisfied. Results from NRC S.G. research programs are expected to lay the experimental basis for many of these criteria. ;
In keeping with the NRC's current and past philosophy on this issue, the proposed regulatory requirements developed in the draft report focus on ISI programs and techniques and tube plugging criteria. 7 The primary responsibility for attacking the problem at its source and ;
eliminating S.G. degradation is the industry's. However, several of the i requirements proposed in NUREG-0844 are intended to promote industry [
For example, one requirement is to ensure that efforts in this area.
all operating plants have implemented an approved secondary water chemistry monitoring and control program. This is a requirement in the most -
recent version of the NRR standard review plan for licensing of new :
plants. In addition, this type of program has been implemented at some '
but not all operating plants. Under this requirement, it is the industry's responsibility to establish specific water chemistry limits and effective .
monitoring techniques. This will ensure that each utility at least considers the importance of secondary system water chemistry and puts in the effort to develop a comprehensive water chemistry program. Simila rly, i ISI requirements for condensers are proposed. .These requirements will hopefully reduce the frequency of condenser in-leakage and encourage ,
l
1
, . 1 l
utilities to improve condenser perfomance. Us6 of noncopper based alloys when retubing condensers and feedwater heaters is also a requirens Additional require::ents are proposed for plants in the preoperating license stage and many reccmmendations for operating and future plants l are made. The intent of the proposed requirements as stated in the I
report is to leave primary responsibility for correcting tne S.G. proble'm in the hands of the industry, to allow the industry flexibility in addressing the issue, but at the same time, to strongly encourage proper industry actions.
B. - Cmprehensive NRC/ Industry Program The preceeding review has attempted to summarize the status of the S.G. issue at this tinie. As indicated, the NRC has many ongoing efforts to address this multifaceted problem. However, to date, joint
- NRC and industry cooperative efforts on this issue have not been extensive.
This is due largely to the different focuses on the issue. NRC is primarily concerned with requiring adequate ISI and corrective actions to ensure primary system integrity, while the industry has been concerned with developing fixes to prolong S.G. service life and reliability. NRC 1 and industry efforts have been primarily complementary in nature. ;
However, to the' extent that reliability implies safety and vice-versa 1 the NRC and industry efforts are synonomous. Therefore, the staff is pursuing the development of a joint NRC and industry program to address both near-tem and long-tem actions required for continued safe operation of steam generators and ultimate resolution of the S.G. degradation problem. The intent is to. evaluate the degree to which the NRC can .
expand its role ~ in prevention of tube degradation and work with the industry to solve' this problem. Efforts to detemine the feasibility of this type of cooperative program have been initiated and proposals for a joint NRC and industry program, will be presented in a later document.
I a
,s.,s .
REFERENCES
- 1. Eisenhut, Liaw, Strosnider, " Summary of Operating Experience wi th Recirculating Steam Generators,", NUREG-0523, January 1979.
- 2. Liaw, Strosnider, " Summary of Tube Integrity Operating Experience with Once-Through Ste.am Generator," NUREG-0571, March 1980.
- 3. SECY-81-664, "Infonnation Report - Steam Generator Tube Experience,"
from W. J. Dircks to the Commissioners, November 24, 1981. I
- 4. Memorandum for Commis'sioner Bradford from'W. J. Dircks, Status of Recanmendations Made in NUREG-0651, " Evaluation of Steam Generator !
Tube Rupture Events," to be transmitted.
- 5. EG&G Idaho, Inc. Report TREE-NUREG-1213 (NUREG/CR-0175), " Investigation of the Influence of Simulated Steam Generator Tube Ruptures During Loss-of-Coolant Experiments in Semiscale MOD-1 Systems," May 1978.
- 6. EG&G Idaho', Inc. Report CAAP-TR-032, " Steam Generator Tube Rupture -
Effects on a LOCA," f,'ovember 1978.
e D
t
\, , . . ,
. h00 D APR 0 21982
., .bLWLbu U L DEBEVOISE & LIBERMAN l
I
. i. . . .
^
/y %,.
g UNITED STATES NUCLEAR REGULATORY COMMISSION ATTAp NT. B
- c. j WASHINGTON, D. C. 20555 f
- JAN 2 01982
~ -
Docket No.: 50-395 ..- -
'Mr. T. C. Nichol's -
Vice President and Group -
Executive-Nuclear Operations - -
South Carolina Electric & Gas Company ,
P. O. Box 764 Columbia, South Carolina 29218
Dear Mr. Nichols:
Subj ect: . Plans for.Use of Westinghouse Model D Steam Gen,erators .- l Within the .last few months, there has been considerable interest paid to - '
operating nuclear power plants with pre-heater steam generators (Model. D) ~
manufactured by Westinghouse Electric Corporation. This interest was ,
initiated by tube failures and degradation in steam generators of .this model at non-domestic nuclear power plants. We understand.that you propose to, -
use Westinghouse Model D pre-heater type steam' generators at your facility. ,
Because of the safety. concerns relative to steam generator. tube ' damage, we consider the potential for such damage to be of safety significance.
Therefore, we' request that you provide us with your plans to address this problem at your facility within 30 days of receipt of this letter. We !
are especially interested to know whether or not you are relying on the .
results of the Westinghouse test program or testing at operating plants,
. what instrumentation you may propose for detection of flow-induced vibrations, and what testing and start-up pro.cedures yo'u propose for you own facility. , ,
Your full cooperation in this matter is appreciated.
The reporting and/or record keeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not'
~
. required under P.L.96-511. .
Sincerely, *
\
Darr . isenhut, Director -
' Division Licensing Office of Nuclear Reactor Regulation ec: Se~e next page .
9 b
- 9
, ~
n C
- 0. .
. .) '
Mr. T. C. Nichols, Jr. -
Vice President & Group Executive '. '
Nuclear Operations. .
South. Carolina Electric"& Gas Company .
P. O. Box -764 - .
. Columbia, South Carolina 29218 . ,
cc: ' .Mr. Henry Cyrus ,
Senior Vice President' '
- South 'CaroTina' Public Servic'e Authority -
223 Nor'th Live Oak Drive -
~
Monc,ks Corner, South Carolina 29461 -
J. 'B. V$otts, Jr. . .Esq. .
Debevoise & Liberman '
1200 17th Street, N. W. -
Washington, D. C.. 20036 ,
- Mr. Mark B. Whitaker, Jr. . . .
,, Group Manager - Nuclear Engineering & Licensing .
South Carolina Electric & Gas Comp ~any
.. ~
~ -
P. O. Box 764 - .
~
Columbia, S'outh Carolina 29218 Mr. Brett Allen Bursey .
Route 1, Box 93C ' . . .
Little Mountain, South Carolina 29076 Resident Inspector / Summer NPS .
! c c/o U. S. NRC-Route 1, Box 64 .
., Jenkinsville, South Carolina 29065 ,.
Dr. John Ruo'ff .
" Post Office Box 96 .
Jenkinsville, S'outh Carolina 29065., ,
Mr. James P. O'Reilly :
U. S. NRC, Reg an~II : .'
101 Marietta Street
. Suite 3100 . .
Atlanta, Georgi.a 30303 .
s
- \ ,
- . r .
' 8
- c. . ,
e
?'
0' ATTACHMENT C Sp "Ig
/
g' '
kg UNITED STATES NUCLEAR REGULATORY COMMISSION 5 ej W ASHINGTON, D. C. 20$55 s,,..
- N MEMORANDUM FOR: The Atomic Safety and Licensing Board for Virgil C. Summer Nuclear Station FROM: Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
SUBJECT:
BOARD NOTIFICATION - PREHEATER TYPE STEAM GENERATOR (BN-82-02)
Enclosed is a memorandum dated December 7, 1981, providing a summary of a meeting held on November 20, 1981, with the Duke Power Company and Westinghouse Corporation. Information was presented regarding the results of testing of the Model D steam generator (similar to those at the Virgil C.
Summer Nuclear Station) at two foreign reactors. Also enclosed is a letter from the Duke Power Company dated December 29, 1981, describing the results of its sequence of plant operation and steam generator inspection. To date, no indication of steam generator tube wear has been observed at the McGuire plant. The staff is closely monitoring the McGuire operation and steam generator test program and is evaluating information from the plant and the Westinghouse test program as it relates to the Virgil C. Summer Nuclear Station. We will keep the Board informed.
@W Robert L. Tedesco, Assistant Director for Licensing
- Division of Licensing
Enclosures:
- 1. Summary of Meeting held on November 20, 1981, dated December 7, 1981
- 2. Letter from Duke Power Company to NRC dated December 29, 1981 s
. . O. C -
Mr. T'. C.' Nichols , Jr. . !
Vice President & Group Executive Nuclear Operations South Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina .29218 cc: Fr. Henry Cyrus .
Senior Vice President South Carolina Public Service Authority 223 North Live 0ak Drive -
Moncks Corner', South Carolina 29461 J. B. Knotts, Jr. , Esq. -
Debevoise & Liberman 1200 17th Street', N. W.
Washington, D. C. 20036 Mr. Mark B. Whitaker, Jr.
Group Manager - Nuclear Engineering & Licensing South Carolina Electric & Gas Company' P. D. Box 764 .
Columbia, South Carolina 29218 Mr. Brett' Allen Bursey Route 1, Box 93C Little Mountain, South Carolina 29076 Resident Inspector / Summer NPS' c/o U. S. NRC Route 1, Box 64 Jenkinsville, South Carolina 29065 .
Dr. John Ruoff Post Office Box 96 -
Jenkinsville, South Carolina 29065 Mr. James P. O'Reilly U. S. NRC, Region II' 101 Marietta Street .
Suite 3100 - '
Atlanta, Georgia 30303 4
O e
C,. U l DISTRI3UTION OF BOARD NOTIFICATION V.C. Summer Docket No. 50-395 ACRS Members Dr. John H. Buck Dr. Robert C. Axtmann Mr. Brett Allen Bursey Mr. Myer Bender George Fischer, Esq. Dr. Max W. Carbon Herbert Grossman, Esq. Mr. Jesse C. Ebersole Dr. Frank F. Hooper Mr. Harold Etherington Joseph B. Knotts , Jr. , Esq. Dr. William Kerr Christine N. Kohl, Esq. Dr. Harold W. Lewis Mr. Gustave A. Linenberger Dr. J. Carson Mark Mr. Randolph R. Mahan Mr. William M. Mathis Alan S. Rosenthal, Esq. Dr. Dade W. Moeller Richard P. Wilson, Esq. Dr. David Okrent Atomic Safety and Licensing Board Dr. Milton S. Plesset Panel , Mr. Jeremiah J. Ray Atomic Safety and Licensing Appeal Dr. Paul G. Shewmon Panel Dr. Chester P. Siess Docketing and Service Section Mr. David A. Ward Document Management Branch
[
e i ,*,
i W
a
,. , ' %\ph b%ha l
. D o n e 1 % w r. u L.osrawy y 1Noe ;
m.... s.......
- r. . . . n . . . . . . . . . h. .
h;p Cg l
... x c c......,, s,...,...........
m g p_
.m._. .. . .. . ne, .e, 2,. i,ai C. 1
.. ... .... .Pv.. 9l n,.o, t
i l
Mr. Harold R. Denton, Direstor 1 Of fice of Nuclear Remetor Regulatinn '
U. S. Nuclear Regulatory Coom'incion {
Wochington, D. C. 20555 '
A*.tantion: Ms. E. C. Adennam. Chief Licensing Branch No. 4
/l Re McGuire Nuclear Station i Docket No. 50-369 Dear Mr. Denton !'
On October 31, 1983 Duke Power Company notified the NRC Staff of tube degradation . i on a non-domestic Westinghouse plant with Mode] D stense generatore similar to those i at McGuire Nuclear station. Subsequently a meeting was held on November 20, 1981 -
in Bethesda to brief the Staf f on the details of the problem as well as Duke Power Company's plans for operation of McGuire. In particular Mr. H. B. Tucker outlined ,
i o p1 maned sequence of plant operation. s t em.= generator tube inspection and instru- i smentation installation. The purpose of this letter is to update the NRC Staff on i the status of this effort to date and on plans for future operation. I Eddy current testing was performed on the 'A' St cam Generator to determine if a i
threshold power level exi st ed at which tube vibration in the preheater was initia-tcd. Rows 49, 48 and 4 7 were examined. This testing was conducted after two weeks operation at approximately 50% power and again af ter one week operation at 75% power. l Thio dicestertesting was conducted differential probe. by Babcock and Wilcox co pany personnel utilizing a .590" l A Zetec M1Z-12 multi-frequency apparatus was employed !
at frequencies of 130 khr, 200 khz, 400 khz and 550 khr. Since thie examination was !
looking for wear damage at tube support plate locations both 130-550 khs and 200- {
400 khs mixed outputs uere used to eliminate the dofoct signals for analysis. An ASEE Section XI type calibration standard was used. support piste signal leaving only I l
l Rocults of both of these inspections (i.e. efter operation at 50% and 75% power) i woro reviewed by Rabcock and Wilcow, Duke Power Crmpany, Westinghouse and EPRI NDT personnel. A comparison wi th the results of the preservice inspection van made, i No wear type indications were observed. j 1
During the November outage. three transducers were m.ounted around the feedwater monale on each'-stense generator. These t ransduc'e rs are i intended to provide an early indication of any gross snechanien) vibration inalde the prehester and will eventuall ('
bo used in conjunction with the internal Instrumentation when installed. To date no cignals have been noted which correlate to preheater/ tube vibration. Resonance peak which have vibration been observed were caused by flow turbulence rather than any mechanical phenomenon.
ik i -- _
o u,
g .
i t
i !
t Mt. Rarold R. Denton l December 29, 1981 I Fago 2 ,
i Currently, the unit is in the startup phase. Plans are to increase power to 90%, l bold at that level for up.to 4 daye then increase to 2001 for one day. Power opera- ,
tien would then continue for up to 6 weeks at power levels up to approximately 75%. l The unit will then be shutdown, eddy current examination performed on all 4 steam i conorators and internal instrumentation mounted in one steam generator as described ;
in our November 20, 1981 meeting in Bethesda. Thie operating plan represents our best efforts to balance testing and operationn) needs with a prudent course of actio: ;
to cssure the integrity of the steam generators. Minor changes to this planned sequ !
once may occur due to unforneen circuantances: however, we will keep you advised of :
cny significant departure f rom this plan. !
Flocae advise if y u have any questione regarding this matter.
i V truly your .
(n. _ o.
. William O. Parker, J
- l CAC/Jfw ces Mr. P. R. Bents Mr. James P. O'Re111y, Regional Administrator ,
Senior Resident Inspector U. S. Nuclear Regulatory Commission (
McGuire Nuclear Station Region II 101 Marietta Street, Suite 3100 l Atlanta, Georgia 30303 ;
i i
i I
i s
\
. .# C UE. 7193; j Docket Nos: 50-369 and 50-370 .
LICENSEE: DUKE POWER COMPANY .
FACILITY: McGuire Nuclear Station. Units 1 and 2
SUBJECT:
SUMMARY
OF HEETING HELD ON HOVEMBER 20. 1981 A meeting was held with the licensee on November 20. 1981, to discuss preheater-type stean generator tube problems. in foreign reactors as related to the operation of McGuire Unit No.1. A list of attendecs is shown on Enclosure 1.
The major briefing presentation was nade by Westinghous'e and included the results to date of testing of the Model D steam generator at two foreign reactors. These steam generators are of similar design (Model D) to those in the. McGuire Unit 1 pl ant. Eddy Current testing has revealed.that there are indications in the outer rows of tubes in essentially all steam generators s.o tested. . Westinghouse has '
initiated a program of testing and tube examination along with analytical evalu- -
, ations to determine the initiator of this tube degradation phenomena. Westinghouse believes at this time that tube degradation can be attributed to excitation of the tubes from hi0 h fluid velocities and/or non-uniforn velocity distribution. .
'The licensee stated that McGuire Unit 1 was shut down en Novenber 16. 1981 and that 3
an operating plan for Unit I had been developed (see Enclosure 2). Rows 47. 48 and 49 in "A" Steam Generator (S/G) were Eddy Current tested. The results were negative with possibly one slight indication. The licensee has installed external vibratiori monitors (3 transducers per S/G) on each S/G. Upon completion of this inspection effort, the unit will be restarted and a power level of 75% established, approx.
November 23. Upon completion of. the traditional 75% plateau power ascension testing
( 2 weeks), the unit will be shutdown and S/G Eddy Current testing repeated on S/G "A". During unit operation nonitor instrumentation will be evaluated. Interim operation on the above basis appears appropriate at this time.
The licensee indicated that further operation at 90% power is contemplated to com-plete some ascension testing ( 1 week) provided no indications above 20% are dis-covered in S/G "A". Following this week of testing, evaluation of the monitoring, and evaluation of 50-90% power data, they would decide on escalation to 100% power for one day. Data evaluation at this time would determine whether or not the licensee would plan to continue operation at an acceptable power level or shut down at that time for further EC testing. -
. c r.
C M ) 7(J(o omcc > ...................... . . . . ............... ........................ ........................ ........................ ........................ . . . . . . . . . . . . . . . . . . .
Mme > . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...................... ...................... ........................ . . . . . . . . . . . . . . . . .
CATE ) . . . . . . . . . . . . . . . ...................... ...................... . . . . . . . . . . . . . . . . . . . . ....................... . . . . . . . . . . . . . . . . . . . . . . ....................
oau m o m wacum4o. OFFIClAL RECORD COPY usy. ini-w
(- -
e
~f... .
- j Since the infornation presented by Westinghouse was proprietary, the licensee agreed to document the information pursuant to 10 CFR 2.79.0. ,
.. i 1
i I
Ralph A. Birkel. Project Manager Licensing Branch No. 4 Division of Licensing
Enclosures:
- 1. Attendance List .
- 2. McGuire Nuclear Station.
Unit 1. Operating Plan e
. e 8
.
- w
,v_-
So t
/
{-
O g
kJ.- ~ k_ ' '
. .m-McGuire .
i.
Mr. William B. Parker, Jr.
Vice President - Steam Production Duke Power Company P.O. Box 2178 -
422 South Church Street .
Charlotte, North Carolina 28242 cc: Mr. W. L. Porter Shelley Blum, Esq.
Mr. A. Carr 1716 Scales Street
~
Duke Power Company Raleigh, North Carolina 27608 P.O. Box 2178 422 South Church Street Mr. Paul Bemis Charlotte, North Carolina 28242 Resident Inspector .
c/o U.S. Nuclear Regulatory Commission Mr. R. S. Howard . P.O. Box 216
. Power Systems Division Cornelius, North Carolina 28013 Westinghouse Electric Corp. .
P.O. Box 355 -
Pittsburgh, Pennsylvania 15?30 Mr. E. J. Keith EDS Nuclear Incorporated ' ~
220 Montgomery Street '
San Francisco, California 94104 Mr. J. E. Houghtaling ,
NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Mr. Jesse L. Riley, President -
The Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28207 J. Michael McGarry, III, Esq.
DeBevoise & Liberman 1200 Seventeenth Street, N.W.
Washington, D. C. 20036 Ms. M. J. Graham Resident Inspector McGuire NPS ,
c/o U.S. Nuclear Regulatory Commission
.P.O. Box 216 -
l Cornelius, North Carolina 28031 O
n
- t. . . . . ,'.-
\. s ATTENDANCE LIST .
McGuire Nuclear Station. Units 1 and 2 i
, hovember 20. 1981 DUKE POWER COMPANY
~
Skip Copp '
. H. B. Tucker C. W. Hendrix, Jr.
A. L. Sudduth WESTINGHOUSE CORP.
H. J. Connors ,
I. C. Ratsep H. A. Weaver N. P. Mueller B. H. Bowman .
Antonio Aldeanueva
- 0. J. Woodruff K. L. Huffman '
T. F. Timraons -
Deryk R. Grain - -
~"
NRC STAFF Ralph A. Birkel William V. Johnston Philip Itatthevis Victor Denaroya .
. Conrad McCracken E. G. Adensam .
K. N. Jabbour C. Y. Cheng W. J. Collins - -
J. Rayan j s T. J. Kenyon i R. L. Tedesco
. j Nick Economos Alan Herdt Keith Wichman Denn'..; Crutchfield Emett Murphy E. Igne OTHERS
-:. Jose I. Villadoniga - I Consejo De Seguridad Nuclear - Spain I Julian - Gorosarri - !
, Alanaraz Power Plant 2 = c ". '. : :: :,
3co . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
...... i ..P,r,0j,e.ct ,,H,a,n,a.g,t , ,,,C.,,,!{,.,,,A,lga ra,z,,
eif Ericson - i ner) . . ....... ... ..... ... . . . . . . . . . . . . . . . .
........ g gg .a g Sci ence"0fftee"- " " " " " " " " " " " """""*"""" '
j oin >} ...... .......... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......................
ppu m p m meucae OFFICIAL. RECORD COPY usa m mi- m l
b .
f- ,
3 ftGUIRE TUCIIAR STATlal - UNIT 1 OPERATilG PIAN .
o CONDUCT EC EXM11 NATION ,
"A" S/G
~
RdWS 49, 48,' 47,I o INSTALL EXTERNAL INSTRLfE1TATIOJ ,
"A" S/G o EVALUATE EC EXM11NAT10N RESULTS IF lielCATIONS' <20% C0tiTINUE OPEF6110N . .
IF ItDICATIONS >20%, INSPECT 4 S/G, EVAllMTE
}
o ESCALATE TO 75% POWER, C&PLETE TESTS, (~2 hEEKS) ,
- FDNITOR INSTRLIiBRATION, EVALUATE o SHLTfDOWN - CONDUCT EC EXAMINATION .
"A" S/G R0WS 49, 48, 47,5 o EVALLMTE EC EXAMINATION RESULTS -
IF IfoICATIONS <20% CONTItUE OPERATION IF INDICATIONS >20%, IRSPECT 4 S/G, EVALUATE O
a ',",
e 9
r- a
. . . . t -
v ,
- MC GulRE NUCLEAR STATION - UNIT 1 - '
t l
' . l OPERATINGPLAM .
o ESCALATE TO 90%, CUPLEIE TESTS Gh80 o KlilTOR INSTRitBiTATION, EVALLMTE - .
o EVAUJATE 50-90E POER DATA
-IFSATISFACIORY,CONTINi)EOPERATION
- IF LNSATISFACTORY, OPERAlE AT REDUED POWER o ESCALATE PWER TO LOO % 0%XIriN 1 DAD o AT THIS PolNT EITER:
!L! aft b W
) '
Of,ANDINSTALLATIONOF
- 2) SHlJp,10A_
ADD.
f TR E IN INSgRLtgS Oh ]
.r o
EN ECT INSRCTION 14 S/G .
INSTALL liffERIML INSTRUENTATION 2 llBES,.1 S/G .
EVAUJATE 50,- 100% POER OPERATION PLUS OTER DATA ..
o DEIERMINE APPROPRIATE NEAR TERM OPERATI,NG C0talTIONS t!S$ OF COMBIED l%IN, AUXILIARY FEED Nn771FS POER LEVEL .
o RETURN To POWER
? b
'sq 0
. .o
-,# . h MEETING
SUMMARY
DISTRIBUTION b) 'ATTAcHMDIT D Docket File -
G. Lear NRC PDR MAR 1 a 1982 . . W. Johnston Local PDR -
S. Pawlicki TIC /NSIC/ Tera V. Benarcya
. . , Z. Rosztoczy LBil Reading W. Haass -
H. Denton/E. Case D. Muller
. R. Ballard .
. D. Eisenhut/M. Jambor W. Regan R.' Ma.ttson B. J. Youngblood * '
A. Schwencer F. Congel F. Miraglia 0. Parr !
J. Miller ~
F. Rosa i G. Lainas W. Butler i R. Vollmer ' '
J. P. Knight R. Houston R. Bosnak * ' ~
F. Schauer L. Rubenstein R. E. Jackson
- T. Speis Project Manager ses.em,, M. Srinivasan Attorney, OELD J. Stolz ,
M. Rushbrook . S. Hanauer OIE (3) W. Gannill ACRS (16) W. Minners ,
F. Schroeder E. Adensam D. Skovholt M' Ernst NRC
Participants:
L. Hulma'n C. Berlinger e
V. Benaroya K. Kniel P. Sears G. Knighton S. Chesnut A. Thadani K. Kiper J. Stang ,
J, Kramer D. Zienenn -
bec: Applicant & Service List
' k9 6
en cao
' ' uq#o 4 g UNITED STATES 8 g NUCLEAR REGULA' TORY COMMISSION
( j WASHINGTON, D. C. 20555
%,..../
Dock 2t Nos.: STN 50-454 STN 50-455, STN 50-456, STN 50-457 50-369, 50-370, 50-390, 50-391, STN 50-546, STN 50-547, 50-445, 50-446, 50-395, 50-413, 50-414, 50-498, 50-499, 50-400, 50-401, 50-402 '
and 50-403 MEMORANDUM FOR: B. J. Youngblood, Chief, licensing Branch No.1, DL P
FROM: S. Chesnut, Project Manager, Licensing Branch No.1, DL
SUBJECT:
SUMARY OF MEETING ON WESTINGHOUSE MODEL D STEAM GENERATORS I A meeting was held February 19, 1982 in Bethesda, Maryland with representatives from Westinghouse and various utilities which had purchased Westinghouse Model D steam generators. A list of attendees is given in Enclosure (1). '
Westinghouse representatives presented the results of their initial evaluations of accelerated tube wear on Model D steam generators noted at two non-domestic plants in Ringhals, Sweden (Model D3) and Almaraz, Spain (Model D3). Eddy current data and information derived from internal and/or external flow monitoring instrumentation of scale models and other Model D steam generators at McGuire (D2) '
and Krsko, Yugoslavia (D4) was also discussed.
Westinghouse representatives reported the the accelerated tube wear was flow related and had only been observed at two non-domestic plants. The accelerated wear was attributed to turbulence in the preheater region caused by the feed inlet impingement plate and flow limiter. Initial data from the instrumented steam generators showed that the onset of the increased turbulence occured at high feed flow rates (approximately 50% for Models D2 and D3, 70% for D4, DS).
Westinghouse indicated that increased instrumentation and additional analyses would continue until a resolution was reached. Westinghouse also discussed several preliminary design change possibilities which would decrease the flow-induced vibrations and turbulence. Westinghouse indicated that it would make recommendations to utilities but that each utility would respond to the staff as to appropriate measures to prevent the accelerated tube wear.
, h;($,' l n v 7 S. H. Chesnut, Project Manager Licensing Branch No. 1 Division of Licensing
Enclosure:
As stated cc: See next page
.i/ j
O 6 .
SHEARON HARRIS Mr. J. A. Jones j Yice Chairman
- Carolina Power & Light Company Post Office Box 1551 Raleigh, North Carolina 27602 i George F. Trowbridge, Esq.
Shaw, Pittman, Potts &
Trowbridge ,
1800 M Street, N. W. '
Washington, D. C. 20036 Richard E. Jones" Esq.
Associate General Counsel Carolina Power & Light Company .
411 Fayetteville Street Mall Raleigh, North Carolina 27602 .
Thomas S. Erwin, Esq.
115 W. Morgan Street Raleigh, North Carolina 27602 M.- David Gordon Attorney Associate General .
State of North Carolina P. O. Box 629 Raleigh, North Carolina 27602 .
George Maxwell Resident Inspector / Harris NPS c/o U. S. Nuclear Regulatory Commission Route 1, Box 315B New Hill, North Carolina 27562 Charles D. Barham, Jr.
~
Vice President and Senior Counsel Carolina Power & Light Company Post Office Box 1551 Raleigh, North Carolina 27602 James P. O'Reilly Nuclear Regulatory Commission, Region II r Office of the Executive Director for Operations 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303
. c. -
a
( (JUTHTEXAS Mr. G. W. Oprea, Jr. Mrs. Peggy Buchorn Executive Vice President Executive Director Houston Lighting and Power Company Citizens for Equitable Utilities, Inc.
P. O. Box 1700 Route 1, Box 1684 Houston, Texas 77001 Brazoria, Texas 77422 Mr. J. H. Goldberg William S. Jordan, III Esq.
Vice President - Nuclear Engineering Harmon & Weiss
& Construction . 1725 I Street, N.W.
Houston Lighting ar.d Power Company Suite 506 P. O. Box 1700 Washington, D.C. 20006 Houston, Texas 77001 Mr. D. G. Ba,rker Brian Berwick, Esq.
Manager, South Texas Project Assistant Attorney General Houston Lighting and Power Company Environmental Protection Division P. O. Box 1700 P. O. Box 12548 Houston , Texas 77001 Capitol Station Austin, Texas 78711 Mr. M. L. Borchelt Central Power and Light Company William M. Hill P. D. Box 2121 Resident Inspector / South Texas Project Corpus Christi, Texas 78403 . c/o U. S. NRC P. O. Box 910 Mr. R. L. Hancock Bay City, Texas 77414 City of Austin "
Electric Utility Department P. O. Box 1088 Austin, Texas 78767 Mr. J. B. Poston Mr. Lanny Sinkin Assistant General Manager for Operations Pat Coy City Public Service Board P. O. Box 1771 Citizens Concerned About Nuclear Power 5106 Case Oro San Antonio, Texas 78295 San Antonio, Texas 78233 Jack R. Newman, Esq. Mr. Cloin Robertson Lowenstein, Newman, Axelrad & Toll Manager, Nuclear Licensing 1025 Connecticut Avenue; N.W. Houston Lighting and Power Company Bashington, D. C. 20036 P. O. Box 1700 Houston, Texas 77001 Melbert Schwarz, Jr., Esq.
Baker & Botts Charles Halligan One Shell Plaza Bechtel Power Corporation -
Houston, Texas 77002 P. O. Box 2166 Mr. E. A. Saltarelli Brown & Root, Inc.
P. O. Box 3 r. John T. Collins Houston, Texas 77001 Nuclear Regulatory Commission, Region IV Office of theExecutive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011
o
('
- 0)
Mr. R. J. Gary -
Executive Vice President and General Manager Texas Utilities Generating Company 2001 Bryan Tower-Dallas; Texas 75201 cc: Nicholas S. Reynolds, Esq. J. Marshall Gilmore Debevoise & Liberman 1060 West Pipeline Road 1200 Seventeenth Street, N. W. Hurst, Texas 76053 Washington, D. C. 20036 Mr. Robert G. Taylor Spencer C. Relyea, Esq. Resident Inspector / Comanche Peak Worsham, Forsythe & Sampels Nuclear Power Station 2001 Bryan Tower c/o U. S. Nuclear Regulatory Dallas, Texas 75201 Commission
. P. O. Box 38 Mr. Homer C. Schmidt Glen Rose, Texas .76043 Manager - Nuclear Services Texas Utilities Services, Inc. Richard Fouke 2001 Bryan Tower Citizens for Fair Utility Regulation Dallas, Texas 75201 1668-B Carter Drive Arlington, Texas 76010 Mr. H. R. Rock -
Gibbs and Hill, Inc. Mr. John T. Collins 393 Seventh Avenue U. S. NRC, Region IV
~~ New York, New York 10001 611 Ryan Plaza Drive Suite 1000 Mr. A. T. Parker Arlington, Texas 76011 s Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230
/ David J. Preister Assistant Attorney General Environmental Protection Division P. O. Box 12548, Capitol Station Austin, Texas 78711 Mrs. Juanita Ellis, President Citizens Association for Sound Energy 1426 South Polk Dallas, Texas 75224
- r. i I
e 9
e
0
( .
Mr. T. C. Nichols, Jr.
Vice President & Group Executive ..
Nuclear Operations South Carolina Electric & Gas Company Post Office Box 764 -
Columbia, South Carolina 29218 cc: Mr. Henry Cyrus .
Senior Vice President South Carolina Public-Service Authority ,
223 North Live Oak Drive .
Moncks. Corner, South Carolina 29461 J. B. Knotts, J'r. , Esq. .
. i Debevoise & Liberman 1200 17th Street, N. W.
Washington, D. C. 20036 Mr. Mark B. Whitaker, Jr.
Group Manager - Nuclear Engineering & Licensing South Carolina Electric & Gas Company Post Office 764 Columbia, South Carolina 29218 Mr. Brett Allen Bursey - -
Route 1, Box 93C Little Mountain, South Carolina 29076 .
Resident' Inspector / Summer NPS c/o U.. S. NRC Route 1, Box 64 Jenkinsville, South Carolina 29065 .
Mr. James P. O'Reilly U. S. NRC, Region II 101 Marietta Street -
Suite 3100 -
Atlanta, Georgia 30303 '
, i l
h 0 . . '
e
~ ' '
t . .. , ,
U
. . . (; .
Mr.b.W.' Shields ,
Senior Vice President - Nuclear Division ~ .
Public Service Company of Indiana ,
P. O. Box 190 New Washington, Indiana. 47162 - -
,.cc: Charles W. Campbell, Esq. Mr. J. J. Harrison, Jr. ;
Vice President and General Counsel U. S; Nuclear -Regulatory Comission '
Public Service of Indiana Re'sident Inspectors Office :
1000 E.. Main' Street -
3738 Marble Hill Road Plainfield, Indiana 4616B Nabb, Indiana 47147 ,
Mr. William Kortier Mr. E. P. Martin Water Reactor Divisions General Manager Westingbouse Electric Corporation Wabash Valley. Power Association P. O. Box 355 P. O. Box 24700 ,
Pittsburgh, Pennsylvania 15230 Indianapolis, Indiana 46224 '
Mr. P. L. Wattelet Mr. James G. Keppler -
Sargent & Lundy Engineers U. S. NRC, Region III -
55 East Monroe Street '
799 Roosevelt Road '
Chicago, Illinois 60.603 Glen Ellyn, Illinois 60137 -
Harry H. Voigt, Esq. ,
LeBoeuf, Lamb, Leiby & MacRae - -
1333 New Hampshire Avenue, N. W.
Washington, D. C. 20036 Thomas M. Dattilo, Esq.
311 East Main Street Madison, Indiana 47250 .
' Joseph B. Helm, Esq. l Brown, Todd & Heyburn Sixteenth Floor .
Citizens Plaza ~
Louisville, Kentucky 40202 -
David K, Martin, Esq.
As'sistant Attorney General .
Room 34, State Capitol - ' -
Frankfort, Kentucky 40601 Mrs. David G. Frey Sassafras Audubon Society .
2625 S. Smith' Road -
Bloomington, Indiana 47401 e
. . 2.'. <.
Q: -
V Mr. Louis 0. DelGeorga ~
Director of Nuclear Licensing Commonwealth Edison Company . .
Post Office Box 767 !
Chicago, Illinois 60690 cc: Mr. William Kortier Atomic Power Distribution Westinghouse Electric Corporation .
Post Office Box 355 Pittsburgh, Pennsylvania 15230 .
Peul M. Murphy, Esq.
Isham, Lincoln & Beale ,
One First Natfonal Plaza .
~
42nd Floor '
Chicago, Illinois 60603 C. Allen Bock, Esq. -
Post Office Box 342 i
' Urbana, Illinois 61801 '
Thomas J. Gordon, Esq.
Waaler, Evans & Gordon 2503 S. Neil . :
Champaign, Illinois 61820 Ms. Bridget Little Rorem Appleseed Coordinator .
117 North Linden Street Essex, Illinois 60935 Mr. Edward R. Crass -
Nuclear Safeguards and Licensing Division Sargent & Lundy Engineers 55 East Monroe Street Chicago, Illinois 60603 ,
U. S. Nuclear Regulatory Commission Resident Inspectors Office RR#1, Box 79 -
Braceville, Illinois 60407 Mr. James G. Keppler U. S. NRC, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 i
L
?.
r l . .
? ~
. .O - -
G .
Mr. Louis 0. De1 George .
Oriector of Nuclear Licensing Commonwealth Edison Company e Post Office Box 767 ,
Chicago, Illinois 60690' cc: Mr. William Kortier Mr. James G. Keppler Atomic Power Distribution -
U. S. NRC, Region III Westinghouse Electric Corporation 799 Roosevelt Road Post Office Box 355 . Glen Ellyn, Illinois 60137 Pittsburgh, Pennsylavania 15230
< Paul M. Murphy, Esq. .
Isham, Lincoln & Beale '
One First National Plaza 42nd Floor Chicago, Illinois 60603 Mrs. Phillip B. Johnson -
1907 Stratford Lane Rockford. Illinois 61107 .
Ms. Bridget Little Rorem Appleseed Coordinator 117 North Linden Street Essex, Illinois 60935 Dr. Bruce von Zellen Department of Biological Sciences ' -
- Northern Illinois University DeKalb, Illinois 61107
~
Mr. Edward R. Crass l
Nuclear Safeguards and Licensing Division
. Sargent & Lundy Engineers 55 East Monroe Street .
Chicago, Illinois 60603
< Myron Cherry, Esq.
~
Cherry & Flynn
- . Suite 3700 Three First National Plaza Chicago, Illinois 60602 f
Mr. Willia' m Fourney U. S. Nuclear Regulatory Commission Byron / Resident Inspectors Office 4448 German Church Road ,
Byron,,I111nois 61010 ,
Ms. Diane Chavez-602 Oak Street, Apt. f4 Rockford, Illinois 61108 G
CATAWBA C -
C Mr. William O. Parker ~
Vice President - Steam Production Duke Power Conparty i P.O. Box 33189 -
Charlotte, North Carolina 28242 cc: William L. Porter, Esq. North Carolina Electric Membership Duke Power Conpar1y Corp.
P.O. Box 33189 3333 North Boulevard Charlotte, North Carolina 28242 P.O. Box 27306 Raleigh, North Carolina 27611 J. Michael McGarry, III Esq.
~
Debevoise 8.Liberman Saluda River Electric Cooperative.
1200 Seventeenth Street, N.W. Inc.
Washington, D. C. 20036 207 Shemood Drive Laurens, South Carolina 29360 North Carolina MPA-1 ,
P.O. Box 95162 James W. Burch, Director Raleigh, North Carolina 27625 Nuclear Advisory Counsel i 2600 Bull Street Mr. F. J. Twogood Columbia, South Carolina 29201 Power Systems Division .
Westinghouse Electric Corp. Mr. Peter K. VanDoorn P.O. Box 355 Route 2, Box 179N Pittsburgh, Pennsylvania 15230 York, South Carolina 29745 Mr. J. C. Plunkett, Jr. James P. O'Reilly, Regional Administra' tor NUS Corporation U.S. Nuclear Regulatory Comission, 2536 Countryside Boulevard Region II Cleamater, Florida 33515 101 Marietta Street. Suite 3100 Atlanta, Georgia 30303 Mr. Jesse L. Riley, President Carolina Environmental Study Group 854 Henley Place Charlotte. North Carolina 28208
. Richard P. Wilson, Esq.
Assistant Attorney General S.C. Attorney General's Office P.O. Box 11549 ,
Columbia, South Carolina 29211 Walton J. McLeod, Jr. , Esq.
General Counsel South Carolina State Board of Health J. Marion Sims Building 2600 Bull Street Columbia, . South Carolina 29201 l
l
Mr H. G. Parris ;
Manager of Power '
i Tennessee Valley Authority l 500A Chestnut Street, Tower II Chattanooga, Tennessee 37401 cc: Herbert S. Sanger, Jr. , Esq.
General Counsel ^
Tennessee Valley Authority
- 400 Comerce Avenue l E11B33 Knoxville, Tennessee 37902 l
Mr. W. Lucp -
Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230 i Mr. David Lambert '
Tennessee Valley Authority 400 Chestnut Street. Tower II Chattanooga, Tennessee 37401 Mr. J. F. Cox '
Tennessee Valley Authority 400 Comerce Avenue, W10B85C Y,noxville, Tennessee 37902 Resident Inspector / Watts Bar NPS c/o U.S. Nuclear Regulatory '
Comission Rt. 2 - Box 300 l Spring City, Tennessee 37831 Mr. David Ormsby Tennessee Valley Authority 400 Chestnut Street, Tower II Chattanooga, Tennessee 37401 ,
i James P. O'Reilly, Regional Administrator U.S. Nuclear Regulatory Comission, Region II 1
101 Marietta Street. Suite 3100 Atlanta, Georgia 30303
? .
' ko
'McGuire ,
Mr. William O. Parker, Jr.
Vice President - Steam Production Duke Power Company P.O. Box 2178 422 South Church Street Charlotte, North Carolina 28242 cc: Mr. A. Carr James P. O'Reilly, Regional Administrator.
Duke Power Company U.S. Nuclear Regulatory Commission, P.O. Box 2178 Region II 422 South Church Street 101 Marietta Street, Suite 3100 -
Charlotte, North Carolina 28242 Atlanta, Georgia 30303 Mr. F. J..Twogood Power Systems Division Westinghouse Electric Corp.
P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. E. J. Keith EDS Nuclear Incorporated 220 Montgomery Street-San Francisco, California 94104 -
Mr. J. E. Houghtaling NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Mr. Jesse L. Riley, President The Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28207.
J. Michael McGarry, III Esq. . -
DeBevoise & Liberman 1200 Sdventeenth Street, N.W.
Washington, D. C. 20036 Shelley Blum, Esq.
1716 Scales Street Raleigh, North Carolina 27608 Mr. Paul Bemis Senior Resident Inspector '
c/o U.S. Nuclear Regulatory Commission P.O. Box 216 Cornelius, North Carolina 28013 O
. -\
, u .
('.
Enclosure 1 \
Attendees List NRC Westinghouse E. Doolittle '
J. Schuities K. Kiper , D. MNalinowski C. D. Sellers N. Mueller .
B. Turov11n. - -.
T. Timmons J.'Gleim , R. Twogood E.~G. Adensam J. McAdoo S. Chesnut D. Sheats G. Lanias G. L. Calhoun W. Johnston D. R. Grain C. Cheng . W. E. Pennell T. A. Ippolito T. J. Kerrigan TVA K. N. Jabbour S. B. Burwell P. G. Ioannides P. Bemis .
A. R. Herdt Swedish Nuclear Power Plant A. Taboada ,
K. Wichman L. G. Larsson R. L. Tedesco .
B. Erikigaard B. Senseney
- R. A. Birkel McGuire
-~
J. Rajan E. J. Brown P. Bemis J. Youngblood G. A. Copp J. I. Villadoniga K.j5. Canady Oak Ridge National Lab. Krsko (Yugoslavia)
C. V. Dadd D. Ieretic Argonne National Lab. Westinghouse Nuclear (Spain)
J. Jendrezejczyk R. DeHaro ,
M. Wa.nbsganss J. Gorosami F. Alomar Texas Utilities B. S. Kacko E. Alarcon S,outh Carolina Electric & Gas Company R. Clary C. A. Drice M. D. Qu1,nton l
- -* ATTACHMENT E SOUTH CAROLINA ELECTRIC & gas COMPANY po.t orries som re.
COLumstA, SouTM CAnoWNA 29218 T. C. Niewo ws. J a.
v.u mu. ., ~. r ou,-.
umw ww .
h ,y 19, 1982 Mr.. Harold R. Denton, Director Office of Nuclear Reactor Regulation O. S. Nuclear Regulatory Ccmnission Washington, D. C. 20555
Subject:
Virgil C. Sununer Nuclear Station D:x:ket No. 50/395 Westinghouse Mel D-3 Steam Generators
Dear Mr. Denton:
4 On January 22, 1982, South Carolina Electric and Gas %mny (SCE&G) received Mr. D. G. Eisenhut's January 20, 1982 letter requesting a respnse within thirty days regarding plants to address problems associated with the preheater section of Westinghouse Model D steam generators. SCE&G's Iresent plan is provided in this letter.
Tne Vd.rgil C. Sunrner Nuclear Station utilizes three (3)
Westinghouse Mel D-3 steam generators. me Mel D-3 incorprates a split flow design where feedwater flow enters at a mid section of 4
the Ireheater section and splits into an upaard and downward flow around the tubes and baffles.
%e tube wear mechanism was first identified in the Mel D-3 steam' generator preheater area at a non-dcznestic plant. After 113 effective full pwer days of operation, this plant experienced a primary to secondary leak in one tube in the prehea*ar section.
Inspections revealed wear on othe.r tubes in the outer three rows in this area.
Another non-dcmestic plant which had a similar operating history found similar indications but of less magnitude.
SCE&G is aware of other operating plants having Model D-2 and D-4 steam generators, and are following their progress as they evce:ed through their startup testing and operation.
As folice-up to previous meetings and frequent telephone discussions, SCE&G net with Westinghouse on Ebbruary 10, 1982 to discuss the nature of the mechanisms observed. We discussed with '
Westinghouse the Irogram that is in place to understand *Ja nature of the phencrnena to permit operation of all preheat s*aam generators.
The Westinghouse Irogram includes the use of scale model testing,
%' 3 - -mw,
),o
Mr. Harold R. Denton February 19, 1982 Page 2 analytical studies, wear testing, pericriic eddy current inspection of operating units and the use of diagnostic internal and external instrumentation in selected cperating plants.
As stated, installation of internal accelecrreters has been used ;
by Westinghouse as a diagnostic tool for verification of the analytical studies and concurrent lab testing. At the present time, diagnostic internal instrumentation is installed or being installed in a Model D-2 unit dcmestic plant and in Mcdel D-3 and D-4 in non-dcmestic plants. Field data collection frcm the diagnostic instrumentation has adequately met the verification needs of the Westinghouse evaluation prcgram and the diverse typ Model D units instrumnted makes it unnecessary to further instrument Mcdel D units at this time. Berefore, Westinghouse has concluded that in-plant, on-line, data availability is not needed for safe operation of the Virgil C. Suamer Nuclear. % e conclusion for not installing internal instrumentation at Virgil C. Sumner Nuclear Station is based on the following:
- 1) fne similarities with the Maiel D-3 unit currently instruTented and the adequacy of diagnostic instrument data currently collected.
- 2) The results to date of the Westinghouse evaluation Irogram including analyses, lab testing and opration plant data, support the start-up program and safe level of power operation defined below.
In our review of the Westinghouse program, we have determined that the results of the Westinghouse anaylsis and test programs and the experience and data gained frca these two operating plants appear to be adegaate. Accordingly, our current prcgram is outlined below:
- cbnduct normal low pwer testing.
- Conduct pwer escalation testing up to 50% power to ccmplete ,
the startup testing up to and including that level.
- Continue operation at 50% power for approximately 2 nonths or at a pwer level above 50% that has teen evaluated, based upon information available at that tine, to preclude significant tube damage.
- Shut down and Eddy current test roas 49, 48, and 47 of one steam generator.
[
Mr. Harold R. Denton Ebbruary 19, 1982 -
Page.3 ' -
'i s
,~ .~> r.
? t 1, f- .
Reevaluate available data to confim continued limited pwer operation capability until a nodification can be a m lished to resolve the problem.
- The proposed schedule may be altered at any time to prform nodifications.
Eblicwirug empletion of this phase of the p. yam SG&G and Westinghouse will jointly establish an operating pwer level for operation of the Virgil C. Sumer Nuclear Station prior to modification. In detemining this pw_r level consideration will be given to:
- Results of the eddy current inspection.
- The experience ard data frca other operating plants.
- Tne status of the Westinghouse analysis and test program.
- The status of any proposed Westinghouse nodification to alleviate or resolve the probim.
In summry SG&G has reviewed the Westinghouse program to approach this problem and will proceed with the outline of activities as described above. Additional details and scheduling information will be provided to the IEC as they becme available.
If you have any cuestions, please let us know.
Very truly yours, j .
< C. .
T. C. Nichols, Jr.
RBC:'ICN:1kb cc: Page 4
Mr. Harold R. Denton Ebbruary 19', 1982 Page 4 cc: V. C. Smmer T. C. Nichols, Jr. .
G. H. Fischer :
H. N. Cyrus H. T. Babb -
D. A. Nauman M. B. Whitaker, Jr. ;
W. A. Williams, Jr. .
O. S. Bradham R. B. Clary M. N. Browne A. R. Koon ,
G. J. Braddick I J. L. Skolds l J. B. Knotts, Jr. l B. A. Bursey :
C. L. Ligon (NSRC)
J. C. Ruoff J. P. O'Reilly H. Grossman !
F. F. Hooper G. A. Linenburger ;
- J. B. Cookinham R. R. Pahan NPCF File ,
i I
T i
b 9
l ATTACHMENT F SOUTH CAROLINA ELECTRIC a gas COMPANY cost orrice .on to.
)
CoLUMe:A, south CARQUNA 292t8
~
. . T.C.NicMots,Ja.
. _ _ v.a nu.. ... c . c=== April 14, 1982 u m o. n. .
~'~
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Cmmksion' Washington, D.C. 20555
- ~
Subject:
Virgil C. Sumer Nuclear Station Docket No. 50/395 Steam Generator Ins' e tation
Dear Mr. Denton:
This is to inform you that South Carolina Electric and Gas Capany (SCE&G) will install internal instrumentation in the A steam generator at the Virgil C.
Sunner Nuclear Station The instrunentation will consist primarily of biaxial accelerometers and will be installed in two tubes as selected by Westinghouse and SCS&G. We expect that work will begin on installation of this equipment in earl ~y May and' kill be couplete within a few weeks.
If you require further information, please contact us.
very truly yours, -
/
T. C. Nichols, Jr.
NBC:TCN:lkb cc: V. C. Sunner G. H. Fischer H. N. Cyrus T. C. Nichols, Jr.
M. B. Whitaker, Jr.
J. P. O'Reilly
~~
H. T. Babb D. A. Nauman C. L. Ligon (NSRC)
W. A. Williams, Jr.
R. B. Clary O. S. Bradham A. R. Koon M. N. Browne G. J. Braddick J. C. Ruoff J. L. Skolds J. B. Knotts, Jr.'
B. A. Bursey
. F. K. Mangan M. D. Quinton g1n ww