ML20078F868

Response Opposing TMI Alert 830921 Suppl to Petition to Intervene.Contentions Satisfy None of Specificity Requirements.Certificate of Svc Encl

ML20078F868
Person / Time
Site:	Three Mile Island
Issue date:	10/06/1983
From:	Blake E METROPOLITAN EDISON CO., SHAW, PITTMAN, POTTS & TROWBRIDGE
To:	Atomic Safety and Licensing Board Panel
References
83-491-04-OLA, 83-491-4-OLA, ISSUANCES-OLA, NUDOCS 8310110205
Download: ML20078F868 (42)

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. s 00CKETED USNRC 13 00T -7 Pl2:1g,I 1 /s/83 UNITED STATES OF Af@lgAF SElet.!W '-

NUCLEAR REGULATORY COMICiIMSTDMSERVICf BRANCH BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

) Docket No. 50-289-OLA METROPOLITAN EDISON COMPANY ) ASLBP 83-491-04-OLA

) (Steam Generator Repair)

(Three Mile Island Nuclear )

Station, Unit No. 1) )

LICENSEE'S RESPONSE TO TMIA SUPPLEMENT TO PETITION FOR LEAVE TO INTERVENE I. INTRODUCTION Pursuant to 10 C.F.R. 52.714(b), the Board's August 8, 1983 " Notice of Hearing On Issuance Of Amendment To Facility Operating License" (" Notice") ordered the petitioners in this j proceeding to file, no later than September 21, 1983, "a list of the contentions which they seek to have litigated in the i matter, *** [ setting) forth the bases for each contention

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with reasonable specificity." The Board further ruled that

"[c]ontentions shall be limited to matters within the scope of the amendment under consideration." Notice, at 4. Licensee 8310110205 031006 PDR ADOCK 05000289 G PDR

[$33()(($

responds herein to the "TMIA Supplement To Petition For Leave To Intervene," dated September 21, 1983.

Licensee begins the response to TMIA's proposed conten-ti-ons with a general discussion of the legal requirements for

contentions (insofar as here pertinent) and their application to this petition. Licensee then addresses each of TMIA's eight proposed contentions individually.

II. RESPONSE TO CONTENTIONS f

A. Requirements for Contentions

1. Scope of Hearing Notice A threshold requirement for an admissible contention is that it address a matter which is within the scope of the issues set forth in the Commission's Notice of Opportunity for Hearing in this proceeding. See Northern Indiana Public Service Co. (Bailly Generating Station, Nuclear 1), ALAB-619, 1

12 N.R.C. 558, 565 (1981); Portland General Electric Co.

(Trojan Nuclear Plant), ALAB-534, 9 N.R.C. 287, 289-90, n.6 3

(1979); Public Service Co. of Indiana (Marble Hill Nuclear Generating Station, Units 1 and 2), ALAB-316, 3 N.R.C. 167, 170-71 (1976).

2. Bases with Reasonable Specificity As the Board emphasized to petitioners (Notice, at 4), the Commission's Rules of Practice, at 10 C.F.R. 52.714(b), further

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require that a petition include with proposed contentions "the bases for each contention set forth with reasonable specif-icity."21/

There are several purposes which underlie the Commission's section 2.714(b) standards:

A purpose of the basis-for-contention re-quirement in Section 2.714 is to help assure at the pleading stage that the hearing process is not improperly invoked. For example, a licens-ing proceeding before this agency is plainly not the proper forum for an attack on applicable re-quirements or for challenges to the basic struc-ture of the Commission's regulatory process.

21/ Generally, pro se petitioners -- unschooled in the Commission's pleading requirements -- are held to less rigid standards of clarity and precision with regard to an interven-tion petition (though a totally deficient petition must be rejected). Public Service Electric & Gas Co. (Salem Nuclear Generating Station, Units 1 and 2), ALAB-136, 6 A.E.C. 487, 489 (1973). And even an inexperienced pro se party is expected to familiarize himself with the Commission's Rules of Practice.

Pennsylvania Power and Light Co. (Susquehanna Steam Electric Station, Units 1 and 2), ALAB-563, 10 N.R.C. 449, 450 n.1 (1979). Further, petitions drawn by counsel experienced in NRC practice must exhibit a high degree of specificity. Kansas Gas

& Electric Co. (Wolf Creek Generating Station), ALAB-279, 1 N.R.C. 559, 576-77 (1975).

Petitioners here can similarly be reasonably expected to exhibit a high degree of specificity in their pleadings. TMIA is represented by counsel experienced in NRC proceedings, and both TMIA and individual members of Joint Petitioners are sea-soned veterans of the TMI-l Restart proceedings, with at least four years of experience in active NRC intervention. As such, they are no strangers to he Commission's requirement that con-tentions have a " basis" stated with " specificity." They are also well aware of the existence of and availability of the Public Document Room, and the procedures for its use. Further, they know the significance of documents such as the Staff's Safety Evaluation Report in a given proceeding. Thus, the pe-titioners here are not entitled to the leniency which may be accorded novice pro se petitioners.

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Another purpose is to help assure that other l parties are sufficiently put on notice so that they will know at least generally what they will have to defend against or oppose. Still another purpose is to assure that the proposed issues are proper for adjudication in the particular proceeding. In the final analysis, there must ultimately be strict observance of the require-ments governing intervention, in order that the _

adjudicatory process is invoked only by those persons who have real interests at stake and who seek resolution of concrete issues.

Philadelphia Electric Co. (Peach Bottom Atomic Power Station, Units 2 and 3), ALAB-216, 8 A.E.C. 13, 20-21 (1974) (emphasis supplied; footnote omitted). '

The notice aspect of the requirement is a natural out-growth of fundamental notions of fairness applied to the party with the burden of proof. The Atomic Safety and Licensing Appeal Board has observed:

The applicant is entitled to a fair chance to defend. It is therefore entitled to be told at the outset, with clarity and precision, what arguments are being advanced and what relief is being asked. * * *

• So is the Board below. It should not be necessary to speculate about what a pleading is supposed to mean.

Kansas Gas and Electric Co. (Wolf Creek Generating Station, Unit No. 1), ALAB-279, 1 N.R.C. 559, 576 (1975) (emphasis supplied; footnote omitted). Moreover, the Licensing Board is entitled to adequate notice of a petitioner's specific conten-tions to enable it to guard against the obstructionism of its

, processes. As the Supreme Court noted, in upholding the Commission's requirements for a threshold showing of materiality for environmental contentions:

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• ** [I]t is incumbent upon intervenors who wish to participate to structure their partici-

, pation so that it is meaningful, so that it alerts the agency to the intervenors' position and contention. * *

• Indeed, administrative proceedings should not be a game or forum to engage in unjustified obstructionism by making cryptic and obscure reference to matters that "ought to be" considered * * * .

Vermont Yankee Nuclear Power Corp. v. Natural Resources Defense Council, 435 U.S. 519, 553-54 (1978) (emphasis supplied).

Yet important as the notice aspect of the standard is, the requirement for bases with reasonable specificity goes beyond t'he " notice pleading" allowed in the federal courts, which has been found to be insufficient for NRC licensing proceedings.

I' See Wolf Creek, supra, ALAB-279, 1 N.R.C. at 575, n.32 (1975).

! On the other hand, the regulation does not require the peti-tioner to detail the evidence which will be offered in support of each proposed contention. Peach Bottom, supra, ALAB-216, 8 A.E.C. 13, 20 (1974).22/ In short, the standard falls some-where in between, and "[t]he degree of specificity with which i

the basis for a contention must be alleged initially involves the exercise of judgment on a case-by-case basis." Id.

There are also certain practical considerations which should play a particularly important role here in the Board's 22/ See also Mississippi Power and Light Co. (Grand Gulf Nu-clear Station, Units 1 and 2), ALAB-130, 6 A.E.C. 423, 426 (1973); Houston Lighting and Power Co. (Allens Creek Nuclear Generating Station, Unit 1), ALAB-590, 11 N.R.C. 542, 548-49 (1980).

application of the " bases with reasonable specificity" standard to a particular proposed contention -- beyond the question of whether the proposed contention provides clear and precise notice of the issues on which Licensee may bear the burden of proof. First, the contention should refer to and address per-tinent documentation, available in the public domain, which is relevant to this facility and this project.3/ The Commission itself has recently emphasized petitioners' duties in this regard. See generally, Duke Power Co. (Catawba Nuclear Station, Units 1 and 2), CLI-83-19, N.R.C. (June 30, 1983). In Catawba, the Commission expressly recognized that "a person who invokes the right to participate in an NRC proceed-ing also voluntarily accepts the obligations attendant upon such participation." Id., slip op. at 10 (citations omitted).

The Commission further acknowledged the " substantial public in-terest in efficient and expeditious administrative proceed-ings." Id., slip op. at 11 (citations omitted). Based on these principles, the Commission held that a petitioner for in-tervention has an ironclad obligation "to diligently uncover and apply all publicly available information to the prompt for-mulation of contentions." Ibid. In the instant case, the re-quirement for specific reference to relevant documentation 3/ See Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Units 1 and 2), LBP-81-24, 14 N.R.C. 175, 181-84 (1981).

applies with special force to the Staff's TMI-1 Steam Generator Repair Safety Evaluation Report (NUREG-1019)("SER"), including its numerous attachments and appendices,4/ and to Licensee's

" Assessment of TMI-1 Plant Safety For Return To Service After Steam Generator Repair," Topical Report 008 ("TR-008"), and the "TMI-1 OTSG Failure Analysis Report," GPUN Technical Data Report No. 341 (July 1982) ("TDR-341"),5/ but also extends to other d'ocketed correspondence and published reports, as well as industry codes.

In addition, there should be either a reasonably logical and technically credible explanation, or a referenced and plau-

.sible authority, for the factual assertions in the contentions.

The petitioner's personal opinion alone is not adequate for .

this purpose.

See generally, Detroit Edison Co. (Enrico Fermi t i

Atomic Power Plant, Unit 2), LBP-78-11, 7 N.R.C. 381, 386-87,  !

aff'd, ALAB-470, 7 N.R.C. 473 (1978); Houston Lighting & Power Co.

( Allens Creek Nuclear Generating Station, Unit 1),

ALAB-590, 11 N.R.C.

542, 547-48 (1980); Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Units 1 and 2), .

LBP-81-24, 14 N.R.C. 175, 181-84 (1981).

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25, 1983. The Staff's SER was served on all petitioners on August Ei 5/ =

TR-008Petitioners were directly notified of the placement of [

(Rev.2) in the Public Document Room by a letter from Staff counsel to Licensee dated August 2, 1983. In addition, [g

'on September 29, 1983, Licensee provided each petitioner with si; an individual copy of TR-OO8 (Rev. 3). TDR-341 has been avail- n able in the PDR since September 3, 1982. g E

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3. _ Challenges to Regulations All rules and regulations of the Commission, and the underlying bases for those rules and regulations, are immune to attack in an individual licensing proceeding unless a petition is first addressed to the licensing board for an exception or waiver.

The sole ground for a petition for waiver or exception shall be that special circumstances with respect to the subject matter of the particular proceeding are such that application ,

of the specific challenged rule or regulation (or provision thereof) would not serve the purposes for which the rule or regulation was adopted.

i The petition must be accompanied by an affidavit in support of that basis for the petition. Other parties must be afforded an opportunity to respond to the I petition, including leave to submit reply affidavits. If the i

licensing board determines that a prima facie showing has been made in support of waiver or exception, it shall, before so ruling, certify directly to the Commission for a determination on the matter. If the licensing board does not determine that a

such a prima facie showing has been made, it must deny the e petition. 10 C.F.R. $2.758; Potomac Electric Power Co. a u

(Douglas Point Nuclear Generating Station, Units 1 and 2), m E

ALAB-218, 8 A.E.C. 79, 89 (1974). C

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B. Response to Contentions TMIA Pronosed Contention 1(a) asserts generally that there is insufficient assurance that tube ruptures during operation or specified accident transients "will be detected in time and prevented" to avoid impermissible radiation releases, because of the inadequacy of "[p]ost repair and plant performance testing and analysis including the techniques used, empirical information collected, and data evaluation," and because of the inadequacy of " proposed license conditions." However, TMIA has failed to cite any authority whatsoever in support of its broadbrush condemnation of both Licensee's testing program and the proposed license conditions. Thus, Proposed Contention 1(a) is lacking in basis, and should be rejected for that reason alone.

Moreover, despite the extensive discussions in the perti-nent documentation of the testing program as well as the pro-posed license conditions, TMIA has failed to identify a single (alleged) inadequacy in either the testing program or in the proposed license conditions.

Indeed, the proposed contention gives no clue that TMIA even read the relevant documents. TMIA has thus ignored TR-008, Appendix A, which is devoted exclu-sively to a discussion of Licensee's testing program, which combines cold and hot precritical testing with the power 3

, escalation program to create "a progressive testing program." "

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Appendix A describes the goals of Licensee's testing program, discusses the scope and chronology of the testing program, and summarizes the results of testing to date. TMIA has also over-looked the Staff's review of Licensee's testing program, which is fully documented in the SER. (See, e.g., SER 593.4, 3.7, at 16-27, 30-33 (Staff concludes that Licensee has " established an acceptable pre-critical and post-critical test and power escalation program to confirm that the leak tight integrity of the repaired OTSG is maintained").)

Similarly, TMIA has failed to address -- or even to refer-ence -- the individual proposed license conditions, which are listed in the SER, $5.2, at page 46, and discussed throughout the SER.6/ Nor has TMIA pointed to any inadequacies in the procedural changes proposed by Licensee to prevent recurrence of steam generator corrosion.7/ In fact, TMIA has not even ac-knowledged Licensee's stricter administrative controls on the 6/ 1 Licensee has recently proposed several modifications to 3

the (SeeStaff's GPUNreuommended license letter, H.D. Hukill conditions numbered 4 and 5. g 3, 1983.)

to J.F. Stolz, NRC, dated October  :

7/ Except in those instances where the context otherwise requires, ~

Licensee's reference in this proceeding to the terms

" corrosion" and " corrosive" when used in relation to the damage ..

to the TMI-l steam generator tubes means the intergranular E

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stress assisted cracking phenomenon that took place when three conditions occurred simultaneously, namely (i) a sufficiently

• high tensile stress, (ii) a susceptible material micro- _

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structure, and (iii) an aggressive environment, and the combi-

. nation of these conditions damaged the tubes.

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introduction of potential chemical contaminants or the stricter controls placed on reactor coolant system chemistry to maintain a noncorrosive environment. These controls, designed to prevent future chemical contamination and corrosion, are discussed in TR-008, at pages 29 through 32 and 33. The Staff's review of Licensee's stricter controls is documented in the SER, $$3.6 and 3.7, at pages 30 through 33, where the Staff concluded that Licensee's stricter controls (and other measures) " provide reasonable assurance that cracking of the OTSG tubes will not recur."28/

TMIA's wholesale failure to address, or even reference, the extensive documentation available to it on the subjects of Proposed Contention 1(a) -- Licensee's post-repair testing pro-gram and the Staff's proposed license conditions -- render the proposed contention fatally defective under the Commission's recent decision in Catawba. See Catawba, supra, CLI-83-19, N.R.C. (June 30, 1983). Accordingly, TMIA's Proposed Con-tention 1(a) must be rejected as lacking in specificity and bases.

TMIA Proposed Contention 1(b) alleges that simultaneous tube ruptures in both steam generators (resulting in excessive -

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g controls, are included in the October 3, E in note 6, above. 1983 letter identified E

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radiation releases and/or uncoolable conditions) is not an incredible event because of "the enormous number of tubes in both steam generators which have undergone this repair process."

As its sole basis for this contention, TMIA cites a September 19, 1982 memorandum from Dr. Paul Shewmon (then Chairman of the ACRS) to Ray Fraley. But, as TMIA well knows, its reliance on the Shewmon memorandum is misleading and mis-placed.

The precise language of the Shewmon memorandum cited by TMIA was also the focus of a May 5, 1983 letter from Commission Chairman Palladino to Representative Edward J. Markey, which

.TMIA obtained through a Freedom of Information Act ("FOIA") re-quest lodged with the NRC.9/ At page 3 of the May 5 letter, Chairman Palladino sets forth the complete quote from the Shewmon memorandum:

I'm sure that the Staff will carefully go over GPU's NDE for the SG tubes, but I would be in-terested in learning more about what happens if a B&W plant with two SG has a tube break in both SG at the same time. It isn't an incredible event for this plant [a B&W plant with two SG],

and it seems to me that it might present the op-erators with at least a challenge the Subcommit-tee could look into.

Thus, read in context, the language which TMIA cites from the Shewmon memorandum provides no basis for TMIA's contention; 9/

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A copy of the May 5, 1983 letter is attached (" Attachment for the convenience of the Board and the other parties. i r

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that is, contrary to the implication of the wording of TMIA Proposed Contention 1(b), Dr. Shewmon never related the concerns expressed in his memorandum to "the enormous number of tubes in both steam generators which have undergone this repair process."

Indeed, a careful reading of the complete Shewmon quote reveals that the point Dr. Shewmon was arguing was that "a tube break in both SG'at the same time" was not "an incredible event" for any "B&W plant with two SG"; that is, Dr. Shewmon's concerns are not only unrelated to the number of tubes repaired a

at TMI-1, but the concerns are also not specific to TMI-1.

Chairman Palladino concurs:

[B]ased on subsequent discussions with Dr.

Shewmon, it is our understanding that he was not specifically identifying a potential safety hazard at TMI-1, but rather, he was indicating an interest in a generic B&W olant response and operator actions which, in his opinion, should be reviewed by the ACRS.

(May 5, 1983 letter, at 4 (emphasis supplied).)

This interpretation is consistent with the position of the ACRS Subcommittee on Metal Components, which has reviewed the f TMI-1 steam generator repair program and has indicated that i i

neither the Subcommittee nor Dr. Shewmon himself have remaining  :

concerns about the TMI-1 steam generator repairs, although they

. will continue to review steam generator issues "as a generic problem, distinct from TMI-1, related but separate." (See i

generally, Transcript of January 28, 1983 meeting, at 290-94, .

especially at 293-94 (Moeller, Major, Etherington).) I I

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Thus, TMIA has failed to provide any basis whatsoever for its Proposed Contention 1(b). While the Shewmon memorandum was distorted in an attempt to provide the appearance of some au-thoritative support for the contention, the memorandum -- read in context -- provides no logical basis for TMIA's allegations.

Moreover, the consideration of beyond-design-basis accidents -- such as multiple steam generator tube failures --

is simply beyond the limited scope of this proceeding. The NRC Staff has concluded that the TMI-l steam generator repair pro-gram has restored the entire reactor coolant system ("RCS"),

including the repaired steam generators, to "the original li-censing basis" (SER 55.1, at 45; cf., TR-008 at 4, 5, 74, 107).

And TMIA has failed to provide any basis to dispute the Staff's conclusions. For all these reasons, then, Proposed Contention 1(b) must be rejected.

TMIA Proposed Contention 1(c) asserts that the plant can-not be operated safely with the repaired steam generators " con- ,

sidering among other things interference which plugged tubes will have in the plant's ability to respond to transients and accidents"10/ because of "[t]he type of plug used, the number of tubes requiring plugging, and [the] choice of tubes to be

plugged, including failure to plug 66 degraded tubes."

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10/ TMIA's use of the expansive phrase "among other things" hardly enhances the specificity of its proposed contention.

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TMIA has completely failed to explain what is wrong with the types of plugs used in the TMI-l repair project; indeed, TMIA has not even indicated which of the various types of plugs used it seeks to challenge. The various types of plugs used were identified and described in detail in TR-008, at pages 56-57, and reviewed in the SER, $53.4.2 and 3.4.3, at pages 22 and 24 to 25. As the Staff observed, the tube plugging tech-niques employed in the repair program "do not differ signifi-cantly from those previously used at TMI-1. " (SER 63.4.3, at 24'.) Moreover, all types of plugs used in the repair program have been previous;y qualified for steam ' generator tube plug-ging, and have been used in other operating plants. (See SER 53.4.3, at 24-25; TR-008 at 56-57.) Particularly given the abundance of information available on the subject, TMIA's fail-ure to provide a reasonably specific basis for its assertions 5 impugning the types of plugs used is fatal to its allegations. s Similarly, TMIA has failed to adequately specify its z

concern about "the number of tubes requiring plugging." TR-OO8 reflects the results of analyses performed by Licensee to de- *

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termine if the steam generators could be safely operated with E up to 1500 tubes plugged, which included consideration of re- si F

duction in total flow and margin to departure from nucleate g

- boiling effects of asymetric flow distribution, effects on flow a

coastdown rate, ;a effects on steam generator mass inventory and 2

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TR-008 also describes the

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effects of tube plugging on small and large break loss-of-coolant accidents, as well as all other accidents and transients analyzed in the FSAR. In addition, TR-OO8 reflects Licensee's consideration of the effects of plugging on moisture carry-over.

(See TR-OO8 at 62-73.) The Staff reviewed Licen-see's analyses, and concluded that:

The thermal-hydraulic consequences of operation with plugged tubes are accept-able. Accident consequences of such operation meet criteria or remain bounded by the FSAR analyses.

(SER $5.1, at 45.)

Despite the extensive information available to TMIA, it has failed to provide any basis whatsoever to undermine the confidence of the Staff in Licensee's analyses of the thermal-hydraulic and accident consequences of operation with plugged .

tubes.

Again, there is no indication that TMIA even read the relevant documentation. TMIA's sweeping, generalized allega-tions of concern about "the number of tubes requiring plugging" thus also lack the requisite basis and specificity.

Nor has TMIA provided the requisite basis with specificity for its criticism of the " choice of tubes to be plugged, 1

l including failure to plug 66 degraded tubes."ll/ This allega-j

! tion amounts to little more than an attack on the pre-existing 1

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9 11/ The number of degraded tubes has been revised to 83.

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" plugging criteria," which are. reflected in Licensee's Technical Specifications (and are fairly standard within the industry), and which were not affected by the repair program and are thus patently beyond the limited scope of this proceed-ing. To the extent TMIA seeks to challenge not the plugging criteria themselves, but rather Licensee's application of those criteria, TMIA has failed to identify any bases whatsoever to support its concern.

The only specific concern TMIA identifies with respect to the " choice of tubes to be plugged" is the failure to plug a number of " degraded" tubes. Again, however, the allegation is nothing more than a challenge to Licensee's pre-existing Technical Specifications, which require the plugging or repair of " defective" tubes only, not " degraded" tubes.12/ Although the status of " degraded" tubes is monitored carefully, such tubes need not be repaired or removed from service, since --

I even if a tube had a 360 flaw -- the tube would not fail with i

less than 40 percent through-wall penetration. (See TR-008 at 60.) Accordingly, TMIA has failed to supply an adequate basis with specificity in support of its concern about "the choice of f tubes to be plugged," so that the allegations must be rejected.

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l 12/ A " degraded tube" is one containing " imperfections equal to or greater than 20% of the nominal [ tube] wall thickness",

whereas a " defective tube" is one containing "an imperfection of such severity that it exceeds the repair limit [40% of the nominal tube wall thickness)." (See Appendix A Technical Spec-ification 4.19.4.)

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In sum, TMIA Proposed Contention 1(c) must be rejected in its entirety as lacking the requisite specificity and bases, particularly considering the volume of relevant information available to TMIA on the subjects of the proposed contention.

In addition, to the extent Proposed Contention 1(c) seeks to challenge Licensee's " plugging criteria," the contention must be rejected as beyond the limited scope of the instant proceed-ing.

TMIA Proposed Contention 1(d) claims that the Staff and Licensee have failed to demonstrate the adequacy of the steam generator tube repair program because neither the Third Party

. Review Report nor the Staff's SER13/ are " credible documents."

In support of this position, TMIA sets forth, as we understand TMIA's arguments, eight somewhat interrelated criticisms of the reports. However, the

" credibility" of the documents is not at issue here and is not a proper subject for a contention. Rath-er, we here respond directly to each of TMIA's underlying enu-merated criticisms, which are -- in essence -- attacks upon the conclusions of the subject reports.

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TMIA first asserts that there are " inherent inconsis- '

w tencies" between the SER and the Third Party Review Report.  ;

13/ While TMIA generally attacks the Third Party Review and the SER, as usual TMIA makes no reference to the specific +.

sections with which they disagree. Licensee's response, there-fore, 4

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' disputed by TMIArests on assumptions as to those portions of the reports 3

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TMIA makes no effort to identify any of these so-called inconsistencies and Licensee therefore is unable to respond to this general assertion.

Accordingly, it should be rejected as  ;

lacking specificity. I Second, TMIA claims that the reports f ail to provide data or calculations to support their evaluations, assumptions and conclusions. While the SER itself contains only limited raw data, the attached consultants' report (SER Attachments 2 through 5) and analyses submitted by Licensee to the Staff (which are publicly available) contain a wealth of data sup-portive of the conclusions reached by Licensee and the Staff.

Similarly, the Third Party Review focused mainly on the two safety evaluations prepared by Licensee (SER Att. 6, February 18, 1983 Report at 4); all other documents received and reviewed by the Third Party Review Group are listed in Appendix D to the SER. Neither the Staff nor the Third Party Review Group is under any obligation to include all the raw data which supports the conclusions in their reports; however, this data j

is documented and available and petitioners are obliged to review the available underlying reports and provide specific details of their disagreements with that data. In these  :

circumstancec, then, TMIA's complaint that the SER and the Third Party Review Report include insufficient raw data must be rejected.

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Third, TMIA states that "much" of the analysis is based on laboratory conditions which do not take into account the age of the plant.

Here again, TMIA fails to identify which specific analyses it is referring to and, as such, its criticisms must be rejected as lacking any specific basis. But, beyond this, a significant portion of the tests and analyses were conducted on 1

actual samples of the TMI-l tubing or upon archive tubing sam-i ples. (See, e.g.,

SER $3.4.2, at 17; TR-008 at 21-22, 24, 26 4

(use of actual TMI tube samples); see also Staff's October 13, 1982 Safety Evaluation (sent to all parties).) In the absence of any explanation by TMIA of any disagreement with the use of i

. actual TMI-1 and archive specimens in testing or any identifi-cation of any additional testing which it claims should account for aging, i

TMIA's generalized complaints on the subject must be rejected as lacking the required basis with reasonable specif-icity.

i TMIA next attacks the qualifications of those individuals who participated in the reviews conducted by Licensee and the Staff, based upon four asserted deficiencies in the r.nalyses performed or the methods of performance. However, as discussed earlier, topics such as the expert qualifications of reviewers are not proper subjects for contentions. Instead, would-be in-I @

n tervenors should directly address the methods, and the substan-tive analyses and conclusions of reviewers. Accordingly, in f

'this context, the four asserted analytical deficiencies are addressed below.

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TMIA claims that reliance was improperly placed on linear fracture mechanics theory as opposed to non-linear theory in determining critical crack length. Licensee does not i

, understand the underlying reasoning behind this criticism, as linear fracture theory is generally more conservative than non-linear fracture mechanics and TMIA advances no reasons why this is not the case here. Beyond that, however, this allega-i tien, as discussed below, constitutes an impermissible attack 4

on the Commission's regulations. (See 10 C.F.R. 52.758 and p.

8, supra.)

Section 50.55a(g) of the Commission's regulations requires that components which are part of the reactor coolant pressure boundary meet the criteria of ASME Code Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components." Es-sentially, Article IWB-3000 of the Code sets forth the accep-

! tance standards for flaw indication; Article IWB-3122.4 states that a flaw which exceeds the referenced acceptance standards "shall be acceptable for service without the flaw removal, repair or replacement if an evaluation analysis, as described in IWB-3600, meets the acceptance criteria of IWB-3600." Arti-l cle IWB-3610, " Acceptance Criteria for Ferritic Steel Compo-i nents 4 in. and Greater in Thickness," states that such a flaw "may be evaluated by analytical procedures such as described in l Appendix A to calculate its growth until the next inspection or the end of service lifetime of the component." Article A-1100 l

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of Appendix A "provides a procedure for determining the acceptability of flaws * * *. The procedure is based upon the principles of linear elastic fracture mechanics" (emphasis supplied).

It is clear, then, that Licensee's linear fracture mechanics analysis is not only in conformity with ASME Code criteria, but is mandated by Commission regulation. TMIA has not petitioned for a waiver of 10 C.F.R. $50.55a(g); its criti-cism of the use of linear fracture mechanics analysis must therefore be rejected as a challenge to that regulation.

TMIA next asserts that axial symmetric stress analysis was improperly relied on in that it would not be applicable to all cracks. TMIA does not, however, supply any basis for this statement nor does TMIA suggest why axial symmetric stress analysis would not be applicable or what type of cracks have not been properly analyzed. As such, this statement is totally devoid of any specific basis to which Licensee could respond.

In fact, Licensee did consider both symmetric and asymmetric stresses, as appropriate for the specific geometry in qucstion.

(See, e.g., TR-008 at 84 (inclusion of lateral tube movement).)

In the absence of some specific basis for challenging Licens-ee's consideration of symmetric and asymmetric stresses, TMIA's sweeping condemnation of Licensee's stress analyses must be rejected.

TMIA also claims that the reviews and reviewers " fail [ed]

to analyze crack resistance on the basis of toughness as opposed to hardness which has no relation to crack resistance."

Licensee does not understand the basis for this contention --

none is stated by TMIA -- in that Licensee did not use material hardness as a calculational parameter. However, as discussed in TR-OO8, Licensee emoirically derived the threshold value for crack propagation in Inconel 600, rather than calculating a value to separately account for material properties such as t'oughness. (See TR-008 at 84). Material properties, including toughness, are thus inherent in the values used. Accordingly, TMIA's criticism of critical crack analyses has no basis what-soever, and -- in its present form -- is nonsensical. Its

.I criticism must therefore be rejected.

TMIA's last assertion in its Proposed Contention 1(d) is t

that the analyses failed to differentiate between the effects of thermal stress on small versus large cracks. Here, TMIA is As wrong and has obviously overlooked applicable information.

explained by the Staff, a linear fracture mechanics analysis, utilizing EPRI's "BIGIF" Code, was performed to determine when a crack of a given initial size can be expected to propagate through wall. The analysis included load cycles imposed by thermal factors. (SER 53.4.2, at 21.) Additional calculations i

were performed by Licensee of main steamline break ("MSLB")

conditions for various arc lengths. (TR-008 at 82-84 and

Figures IX-2 and IX-5; see also, SER 93.4.2, at 21. ) Thus, TMIA's allegation is baseless.

In sum,.TMIA has failed to provide any basis with the requisita specificity in support of its Proposed Contention 1(d); therefore, this contention should be rejected in total by the Bohrd.

TMIA Proposed Contention 1(e) castigates the Staff and_Li-censee.for thei'r alleged failure to " consider any alternative repair process," including " removal of the steam generators."

'However, TMIA cites no authority whatsoever to support its baro assumption that evaluation of such alternatives is required under the Atomic Energy Act. Indeed, the controlling law is to the contrary.

A would-be intervenor is, of course, free to argue that an

, applicant's propo' sal fails to meet specified applicable

' Commission requirements, and that the application should there-fore be d'enied. Eut the fact that the applicable requirements g

may slso be sa'tisfied --( perhaps even in a more satisf actory -

g manner -- by some other alternative is legally irrelevant.

See, e.g., s Consolidated Edison Co. of New York (Indian Point Station, Unit'No. 2), ALAB-188, 7 A.E.C. 323, 338-39 (1974);

. }

g.

Wisconsin Electric Power Co. (Point Beach Nuclear Plant, Unit

'2), ALAB-31, 4 A.E.C. 689, 693 (1971); Consumers P6wer Co.

.(Midland Plant, Units 1 and 2), ALAB-35, 4 A.E.C. u 711, 711-12

'(1971); Wisconsin Electric Power Co. (Point Beach Nuclear

=3

• l$Nd-a

, - w

Plant, Unit 2), ALAB-78, 5 A.E.C. 319, 330 (1972).

Accordingly, Licensee here was not obligated to consider alter-natives to kinetic expansion for the repair of the steam gener-ators.

For this reason alone, TMIA Proposed Contention 1(e) must be rejected.

Moreover, TMIA is mistaken in its allegation. Contrary to TMIA's assertions, Licensee and the Staff did consider alterna-tive repair methods (see, e.g., slides presented at April 7, 1982 meeting of Licensee and Staff, which are available in the Public Document Room and which discuss the " sleeving" and

" rolling" options), although the consideration of the steam generator replacement option is not reflected in the public record.

TMIA Proposed Contention 2(a) asserts that Licensee and the Staff have failed to demonstrate that the corrosion experi-enced will not reinitiate during plant operation, rapidly progress and attack the steam generators or other primary system components.

In support of this assertion, TMIA claims that neither the causative agent, the source of initiation nor the conditions under which the initiation of corrosion origi- .

nally occurred have been properly identified. TMIA then goes .

F

?

on to allege that, due to this asserted deficiency, no reliance i can be placed on the clean-up process, corrosion stress i

f analyses, procedures to eliminate corrosion or, indeed, upon

' conclusions that the causative agent has been removed from the d

RCS.

e 1

The causes of the stress corrosion of the TMI-1 steam gen-erator tubes, and the conditions under which it propagated, have been the subject of extensive analyses by Licensee and various Licensee and Staff consultants; these analyses have subsequently been reviewed and evaluated by the Staff. (SER

$3.1, at 4-8; see generally SER Atts. 2-4 (Staff consultants' reports), Att. 6, Third Party Review Report of February 18, 1983, at 7; TDR-341.) Despite this wealth of information,14/

TMIA -- as is its penchant -- fails to discuss this data or to provide any explanation of why it believes these analyses are incorrect or inadequate.

TMIA states no basis for its belief that there may have been causative agents other than those identified by Licensee and the Staff and, again, conveniently ignores the fact that Licensee has previously investigated and rejected other possible contaminants. (See TDR-341, at IV-1, V-1, V-2 and VI-8 through VI-10.)

In the absence of such ex-planations, TMIA Proposed Contention 2(a) must be rejected for 3

failure to set forth any basis with reasonable specificity.15/

14/ The analyses agree that the degradation of the TMI-1 7 steam generator tubes was due to a combination of soluble sulfur i 5

compounds at corrosive-inducing levels (most likely from thiosulfate the tank additions), the highly sensitized nature of :j Inconel 600 tubes and the high axial tensile stresses expe- J i

rienced during cooldown and cold shutdown following hot func-tional testing.

Report"), at 2 (SER 53.1, at 8; see also SER Att. 3 ("Dillon (rejecting carbon as source of contaminant).) .

[

15/ In that TMIA has provided no support for its allegation

'that the tified, Licensee source (s) of the corrosion has not been properly iden- i 9

sees no need to respond in detail here to m

(Footnote continued)

E::

b' L

Proposed Contention 2(b) again claims that there is no as-surance that corrosion will not reinitiate during operation, because: (1) Staff consultant Paul Wulp/ believes that the risks associated with cleaning are greater than simply living with a large sulfur inventory in the RCS; and (2) there is no assurance that cleaning will remove more than 50% to 80% of the contamination and that the residue will not cause reinitiatiaa.

With respect to TMIA's first claim, it should be noted that the Dillon report concerns cited by TMIA dealt with possi-ble risks during the clean-up process itself, not with risks associated with operation after completion of clean-up. (cee SER Att. 3, at 12.) In any event, leakage testing during and following the desulfurization process showed no evidence of corrosive attack. (TR-008 at 33; see also GPUN letter, H.D.

Hukill to J.F. Stolz, NRC, re post-repair testing for leakage, dated July 20, 1983, at 2 (" Attachment B").) Thus, there is no (Continued) l l

TMIA's assertions regarding the removal of contaminants from the RCS, or the reliability of the RCS clean-up, corrosion control procedures or stress analysis. Suffice it to s?y that these subjects are discussed in detail in both the SER a '

l TR-OO8, which TMIA again has not even addressed.

l lb/ Mr. Wu is a Staff employee, not a consultant to the Staff; Mr. Wu was the Staff recipient of a report from consultant R.L.

Dillon. (See SER Att. 3.) Thus, it is not Mr. Wu's views at I all that TMIA references. It is the views of Mr. Dillon in a l report to Mr. Wu. Even as to Mr. Dillon, as discussed infra, TMIA's reliance is misplaced.

l

basis for TMIA's claim that the RCS clean-up may adversely impact operation. Clean-up has already been conducted with no adverse effects, and subsequent favorable testing has been performed and reported. With the exception of the results of the successful completion of approximately thirty days of hot steam generator testing, this information is available to TMIA, but has been simply ignored. There is no basis for TMIA's proffered contention.

TMIA's Proposed Contention 2(b) further claims that the potential removal of 50% to 80% of the sulfur contamination may not be sufficient to prevent future reinitiation, but again cites no support for this assertion. The Staff (and its con-sultants) have evaluated the amount of sulfur remaining in the system and have determined that the concentrations are " removed to an acceptable extent." (SER $3.5, at 29; see also TR-OO8 at 27, 32-33.) While the cleaning process has reduced the level of sulfur in the RCS, it is likely that the concentrations ex-isting prior to the clean-up would themselves not have precipitated additional stress corrosion cracking. (See SER Att. 3, at 13-14.) The reduction in the level of contaminants merely provides added assurance that, in combination with other corrosion controls adopted by Licensee, reinitiation of corro-sion will not occur. (SER $3.7, at 33.) TMIA addresses none of these points.

In sum, then, TMIA has failed to meet the requirements of 10 C.F.R. $2.714(b) in that it has not provided a reasonably specific basis for its Proposed Contention 2(b).17/ According- '

i~

ly, Proposed Contention 2(b) must be rejected.

i TMIA Proposed Contention 2(c) alleges that neither the i

Third Party Review nor the Staff's SER are credible documents in their evaluation of "the causative agent, clean-up, or pro-

l. cedures to prevent contaminant reintroduction" and that those i

documents therefore cannot be used as a basis for concluding i

that the repairs are safe, because of the assertions set forth in Proposed Contention 1(d). TMIA provides no more support for the allegations here than it provided in Proposed Contention 1(d). Proposed Contention 2(c) must therefore also be rejected, for the same reasons.

l t

i l 17/ As a separate assertion under Proposed Contention 2(b),

TMIA claims that Licensee and the Staff have failed to consider i alternatives to the clean-up process. As discussed in response to Proposed Contention 1(d), Licensee is under no obligation to provide a consideration of alternatives under these circumstances.

I 1

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III. CONCLUSION For all the foregoing reasons, none of TMIA's proposed contentions should be admitted for litigation. Accordingly, TMIA's petition for intervention in this proceeding should be denied.

Respectfully submitted,

&f f. /BL,gff, George F. Trowbridge, P.C.

Ernest L. Blake, Jr., P.C.

Delissa A. Ridgway Diane E. Burkley SHAW, PITTMAN, POTTS & TROWBRIDGE 1800 M Street N.W.

Washington, D.C. 20036 (202) 822-1000 Counsel for Licensee Dated: October 6, 1983 i

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%, UNITED STATES y g g NUCLEAR REGULATORY COMMISSION 3O E WASHINGTON, D. C. 20555 e 7;-Am Fj DOCKET,ED L'%

ig* *cf4,e May 5,1983 CH AIRMAN og3 {]} .] f}2 l}h

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The Honorable Edward J. Markey, Chairman Subcomittee on Oversight and Investigations .

Co=nittee on Interior and Insular Affairs United States House of Representatives Washington, D.C. 20515 -

Dear Mr. Chairman:

This is in response to your letter of March 23, 1983 which raised questions resulting from my letter to you of March 21, 1983 and the Subcomittee on Oversight and Investigations' December 13, 1982 hearing. Responses to'the questions in your letter are enclosed.

Additionally, you reminded us of our promise made during the February 22, 1983 Energy and Environment Subcomittee hearing, to provide information relative to the exact status of completion of items in the TMI Action Plan. .

This information was provided to your staff by our Office of Congressional Affairs on March 25, 1983.

We hope this information resolves your outstanding questions with regard to these subjects.

Sincerely, >

~f

.gd:.%W Nunzio J. Palladino

Enclosures:

Response to Questions cc! Rep., Ron Marlenee k

', RESPONSES TO CONGRESSMAN MARKEY'S QUESTIONS

. c '

OUESTION 1: Could a tube rupture in both SGs of a two SG plant such as TMI-1 result in core damage?

RESPONSE

A single . tube rupture in both SGs of a two SG plant such as TMI-1 is not expe'cted to result in core damage. In order for core damage to occur, -

there must be a sufficient loss of primary coolant inventory such that core uncovery and excessive fuel heatup occur. The inventory loss through a ruptured tube in each steam generator could be adequately -

compensated for by the high pressure safety injection system flow.

. In the event of a SGTR, the primary objective of the operator is to iminimize any radiological releases. The operator accomplishes this by equalizing the pressure between the primary system and faulted SG to stop the leak of -adioactive primary coolant into the SG, and then isolating the faulted SG. The plant is then cooled down and depressurized using .the intact SG to remove decay heat. These actions are prescribed by emergency operating proc ~edure guidelines. In the event both SGs of a two SG plant had ruptured tubes, the operator woul'd be forced to accomplish the cooldown and depressurization using at least one faulted SG. The use of the faulted SG would result in continuous leakage of primary coolant to

, - the secondary system during the entire cooldown process. This leakage of primary coolant to the secondary system eventually can result in releases of radioactive material to the environment. However, as stated earlier this event would not be expected to lead to core damage.

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'0UESTION 2: Does the NRC agree with Dr. She'wmon that a tube rupture in both SGs at TMI-1 at the same time is a credible event?

RESPONSE

The staff's preliminary assessment indicates that the simultaneous failure of tubes in both steam generators of a t,<o steam generator plant is highly unlikely. The probability of such an event and its consequesces are currently being evaluated in the generic program relating to steam .

. generator tube degradation and tube rupture events. Tne probability of~ .

such an event will. therefore be better qualifief upon completion of that review.

As stated in the response to Question 1, staff analysis indicates that it is very unlikely that a tube rupture in multiple steam generators would lead to core damage. Since our review will ensure that when repairs are complete, the steam generators and nactor coolant system structural design basis and leakage limits are returned to within original licensing ba'ses for the plant, we conclude that our generic, analyses, those to date and ongoing, apply to TMI-1. Tne factors that iead the staff to this conclusion are as follows: ~

1. The licensee has th6 roughly quantified the corrosion conditio[of the steam generators by conducting 100% eddy testing (ECT)* of both . steam gene rato rs . The extent of ECT for the TMI-1 SGs is greater than that performed at any other operating plant. The techniques used and extent of ECT provide reasonable assurance that defects which may be present have been detected. -.

2. When considering SG tube ruptures, tubes in the fme span (the S2 feet open area between upper and lower tubesheets) are the primary concern, because this is the only location where the classic guillotine break is po s si b.i.e. Tubes within the tubesheets are restrained from separatino within the tubesheet c.revice. Therefore, although ieakage in the tube-sheet is possible, " tube rupture" in the classic sense is not. Greater .

than 95% of all' corrosion at TMI-1 too'K piace within the upper ' tube-sheet crevice, whert separation is restrained. All tubes in the free span of both TMI-1 SGs :that are icentified as significantly degraded' will be removed f rom'seryi'ce. Tne refore, both TMi-3 SGs. ~will be' returned to servi.ce under the same criteria as otber units which haye experienced corrosion; i

~

ECT is a means whereby the electrical conductivity of a tube is checked by passing a ccil with an induced voltage along the tube. if some form of tube degradation has occurred (such as corrosion) which has separated the tube metal, an electrical dis-continui ty exists. The electrical discontinuity will be proportional to the amount of metal which is missing. If 40% or more of the tube wall i$ missing, the tube is duaififCu_Leie_clive and must be repaired or removed frem service.

.  ? .

3. The most limiting initiating event for a SG tube rupture is the main steam line break (MSLB) accident. Under MSLB, maximum differantial pressure will exist on the tubes. For a tube to ' rupture during a MSLB, it would have to be uniformly degraded through by greater than 70% of its wall thickness. The tube plugging criteria of less than 40% includes a corrosion allowance for the next: operating period and an uncertainty allowance in ECT. Because most corrosidn mechanisms do not result in uniform degradation that would cause strdctural

. failure before an unacceptable leakage occurs, 40% plugging criterio'n

. .is very conservativi. This is evidenced by the fact that no SG tube.

ruptures (leakage in. excess of 100 gpm) due to corrosion have occurred

'since'1976, and only two occurred prior to that time.

/, . In addition to the conservatism of the tube plugging criteria, a number of other factors contribute to making tube ruptures in multiple SGs at

. TMI-1 highly unl.ikely. The responses listed below are specific to TMI-1.

a. ' A significant difference exists in the extent of corrosion between the two SGs. Although both SG have been repaired to the same criteria, a statistical difference exists as to the potential for continued corrosion. This factor reduces the probability that -

ruptures would occur in multiple SGs at the same time, even in the event of a MSLB.

b. The corrosion which has been found is circumferential, and in mo'st cases involves less than half the tube circumference. This results in sufficient tube wall remaining to maintain str'uctural strength, even for a MSLB.

. c '. Extensive pre-critical hot functional testing (approximately 6 weeks)'

wi.11 be performed to verify serviceability of the SGs. If undetected corrosion is pre ~sent or progresses'during this period, the resultant

~1eakagelwi'll provide an indication that additional repairs are, necessary. ,

d. Subsequent to criticality, power escalation will be slow (approxi-mately 12 weeks to reach 100% power). Once full power is' reached the plant will be shutdown and the SGs examined by ECT to monitor for continued corrosion.

l l e. Extensive eff6rts have been conducted to remove the contaminants I

and mitigate the possibility of recontamination of the reactor coolant system. .

To provide additional. perspective on this issue, the complete cuote from Dr. Shewmon's memorandum is:

"I'm sure that the Staff will carefully go over GPU's NDE for the SG tubes, but I would be interested in learnine core about what happens if a E&W plant with two SG has a tube break in both SG at the same time. It isn't an incredible event for this plant, and it'seems to l me that it might present the "operators with at least a challenge the

. . o 4

Thus, based on subsecuent discussions with Dr. Shewmon, it is our understanding that he was not specifically identifying a potential safety hazard at TMI-1, but rather,: he was indicating an interest in a generic B&W plant response and operator actions which, in his opinien should be reviewed by the ACRS.

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OUESTION 3: What analysis has the NRC or the ACRS done to evaluate the probability or consequences of this risk at TMI-1 .

and what were the reasons? -

RESPONSE

No probabilistic risk assessment of the subject event has been: performed -

by either the NRC staff or the ACRS for THI-1. Such an assessment is considered unnecessary by the NRC staff for reasons set forth in the responses to Questions 1 and 2, above for the restart of TMI-1. '

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QUESTION 4: Why has the NRC decided not to require consideration or resolution of this issue before restart?

RESPONSE

With reference to the issue of' tube ' rupture in both s' team generaiors, our '

Ma'rch'21, 1983 letter to you sta'ted that the Commission does n6t' consider -- ~

9 the irrplementation of TMI Action Plan Item I.C.1, which calls for emergency operating procedures to addmss events beyond the design basis, to.be necessary befom the restart of TNI-1. This consideration is based upon ,

the. fact that we intend to ensure.that, when the steam generator repairs are complete, the steam generator' and reactor coolant system structural design bases and leakage limits are mturned to within the original- .

licensing bases for the plant s6 that the possibility of the postulated event is not more likely at TMI-1 tha't at any other operating PWR as .

xplained in the response to Question 2.

It should be noted that the licensee has submitted with its steam generator repair safety evaluation a summary of operational guidelines for single and ' multiple tube ruptures in one or both steam generators. The staff is reviewing this sumary as one aspect of its TMI-1 steam generator' '

repair review effort, and will address the guidelines in the forthebming staff safety evaluation.

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~7-00ESTION 5: What accidents within the design basis for TMI-1, but outside the scope of the restart hearings, will not be resolved before restart?

RESPONSE

It is the Comission's intent that, at the time of restart, if permitted, -

all significant issues involving accidents within the design basis, but outside the scope of the restart. hearings for TMI-1, will be resolved.

With respect to Unresolved Safety Issues (in the context of Section 210 '

of.the Energy Reorganization Act of 1974), NUREG-0737 action items, and other generic issues, TMI-1 will be treated as an operating reactor.

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GPU Nucle:r Carp rati::n A U, )

- -- U Gar Po$t o"'c' So* '8o Route 441 South D$h{EJ,ED Miccietown, Pennsylvania 17057 0191

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717 944 7621 TELEX 84-2386 Writer's Direct Dial Number:

,83 00T -7 P12 :16 LFr'C 0: SEC;t.4 July 20, 1983 EF vi 5211-83-208 00CKEig Office of Nuclear Reactor Regulation At n: J. F. Stoln, Chief Operating. Reactor Branch No. 4 Division of I.icensing U. S . Nuclear Regulatory Co-4 ssion

'a'ashin gt on, D. C. 20555

Dear Sir:

Three Mile Island Nuclear Station, Uni 1 (TMI-1) l Operating License No. DPR-50 .

Docket No. 50-289 TMI-l Stea= Generator Repair Status Update our let:er of June 13, 1983 (5211-83-179), informed you of results of post repair tes:ing of the stea: generators perfor=ed to that date and plans f or '

future :esting. This letter is to inform you of additional test results through early July including the early results of reactor pri:tary syste=

cleaning.

Following repair of the residual tubes identified in our June 13.le :er, we conducted additional drip and bubble tests of both OTSG's on June 17 and June

26. These tests on OTSG "3" showed no observable leaks. The drip test on OTSG "A" identified five plugs (f our rolled plugs and one explosive plus botto= tube sheet) and one tube with drops of water clinging to the end. -No drops were seen to fall f ree over the 30 second obse:vation period f ro= any of

'hese six tubes. This particular tube (unplugged) had been inspected by ECT t

in the previous month and dispositioned as no detectable defects. This tube was not plugged. The leakage from the plugs was ex:remely snall and since fu:ure operation may seal the plug leakage with time, no further repair action regarding the dripping plugs is planned at this eine.

The bubble tes: on OTSG "A" identified seven plugs (one welded and six rolled) a.nd three unplugged tubes from which some gas bubbles were de:ected by visual observation. The welded plug was repaired. This vas the first bubble tes:

on this welded plug since 1: had been repaired. All of the tubes and rolled plugs passing nitrogen in CTSG "A" had very fine strears of :iny bubbles which did not cause surface disturbances of the wa:e layer. This amount of bubbling is less than that observed in previous :esting. Due to the small

a

cunt of leakage, no fur:her ac: ion was :sken regarding :he :hree tubes and six relled plu;s. The results of pre-critical he: fune:ional OTSC testing will be used to disposition these tubes prior :c cri:ical plan: opera:icn.

GPU Nuclear Corporation is a subsidiary of the Genera: ;ublic Uti!ities Corporation L

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- 2::t-:: 05 As in dica:e d in th e J6ne 13 le tt er. we a e ee...d ..u.< - .. e . .-' ..- .. _1., . e _e

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-he s:ea: generators and vill assess tha: infer =a:ien a: the =anage=en-reviev :ollowing s:ea: generator he: esting. The S:aff vill be infor=ed of

--....s v w.s cene_,usiens prior to proceedinz :o cri:icali:y, "a have ce=ple:ed approxima ely 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> cf RCS cleanup. The sulfur level in :ne RCS varer shers a s_1::: increase as a fune len

.. 3 a:e r . .

t=casurec as su of ti=e. The sulfste has increased frc= an ini:ial value of 220 ppb te approxi=a:ely 360 ppb as of July 18. This sulfur increase is substantially less than predicted based on assu=ing the =axi=u: levels cf con:a=ina:icn drawn f rc= svipe data frc= the areas of highes concentration (i.e., OTSG

~ upper head area).

Throughou: the RCS cleanup, ve have been =eni cring for pri=ary to secondary leakage by exa=ining the secondary side f or ac:ivity. No tri:iu= has been de:ected to da:e. Tritiu= is currently 4.4 x 10-3 uci/=1 in the RCS and has a =ini=u: detectable activity of about 5 x 10-6 u ci /=1. If ene assu=ed equal =ixing in the OTSG secondary side, the leak, if any, at the curren: 300 psi differenzial pressure is less than 2 x 10 3 gp= based on no detectable tri:iu=. There have been several Cesium-137 =easure=ents above the mini =u=

de:ectable activity of 1 x 10-7 uci/=1. These =easure=en:s have been interspersed with =easure=ents belov the =ini=u= de:ectable activity. The highest secondary side Cesiu=-137 ac:1vity which was =easured a#:er 10 days is 1.86 x 10-7 uci/=1. Since Cesium-137 is curren:1y 1.6 x 10-# uci/=1 in

-he RCS, this corres ends to a primary c secondary leak rate of equal to e s s _....- ..,_ x. 10- g .s _ .

Our schedule shows ce=pletien of.the 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> scheduled RCS cleanup en July o its ner=al cpera-

^4 vich 6 e 10 days required c res:cre RCS che=istry

-icnal range. "e expect to be prepared :c ce==ence pre-critical he: esting

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'83 OCT -7 P12 :16 UNITED STATES OF AMERICA NUCLEAR REGULAggygCp gISSION t 00CKETING & SE6 4.

BRANCH BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

METROPOLITAN EDISON CO., ET AL ) Docket No. 50-289-OLA

) ASLBP 83-491-04-OLA (Three Mile Island Nuclear ) (Steam Generator Repair)

Station, Unit No. 1) )

CERTIFICATE OF SERVICE I hereby certify that copies of " Licensee's Response to TMIA Supplement to Petition for Leave to Intervene" dated October 6, 1983, were served this 6th day of October, 1983, by deposit in the United States mail, first class, postage prepaid, to those persons on the attached Service List 0 lot . 04, Delifsai A. Ridcfw@ ) () U Dated: October 6, 1983

UNITED STATES OF .2_M.IRICA NUCLEAR REGULATORY COMMISSION 3EFORE THE ATOMIC SAFETY AND LICENSING 30ARD In the Matter of )

)

METROPOLITAN EDISON CO., ET AL ) Docket No. 5 0- 2 8 9-OLA

) ASLDP 83-491-04-OLA (Three Mile Island Nuclear ) (Steam Generator Repair)

Station, Unit No. 1) )

SERVICE LIST

• Shelden J. Wolfe, Chainnan Joanne Doroshow/I.cuise Bradford At=nic Safety and Licensing Board Three Mile Islard Alert, Inc.

U.S. Nuclear Regulatory Comissicn 315 Peffer Street Washington, D.C. 20555 Harrisburg, Perr.sylvarda 17102 Dr. David L. Hetrick Jane Lee

-Professor of Nuclear Engineering 183 Valley Road "niversity of Arizona - Etters, Pennsylvania 17319 Tucscn, Arizona 85721 Bruce Molholt, PPD.

Dr. James C. Lamb, III Haverford College De:artrent of Env:n:nmental Sciences Haverford, Pennsylvania 19041 ard E:sineering University of North Carolina Norran Aamodt Chapel Fi11, North Carolina 27514 R. D. 5, Bcx 428 Coatesville, Pennsylvania 19320

• Mary E. Wagner, Esg

• e Office of Executive Legal Director Atmic Safety and Licensing Appeal U.S. Nuclear Regulatory Comission Scard Panel Washington, D.C. 20555 U.S. Nuclear Fegulatory Camissicn Washingt~n, D.C. 20555 Docketing ard Service Section (3)

Office of the Secretary U.S. Nuclear Regulatory Ccmission Washington, D.C. 20555 t

l Atcznic Safety ard Licensing Board Panel U.S. Nuclear Pegulatory Ccmnission Washington, D.C. 205ao 9

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