ML20079E348

Answer to First Set of Interrogatories & Request for Production of Documents Re Steam Generator Repair. Certificate of Svc Encl

ML20079E348
Person / Time
Site:	Three Mile Island
Issue date:	01/13/1984
From:	Churchill B METROPOLITAN EDISON CO., SHAW, PITTMAN, POTTS & TROWBRIDGE
To:	THREE MILE ISLAND ALERT
Shared Package
ML20079E345	List: ML20079E338 ML20079E348
References
83-491-04-OLA, 83-491-4-OLA, ISSUANCES-OLA, NUDOCS 8401170255
Download: ML20079E348 (105)

Text

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% January 13, 1984 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic-Safety and Licensing Board In the Matter of )

)

)

)

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

) ASLBP 83-491-04-OLA

) (Steam Generator Repair)

)

(Three Mile Island Nuclear )

Station, Unit No. 1) ).

LICENSEE'S ANSWER TO TMIA'S FIRST SET OF INTERROGATORIES AND REQUEST FOR PRODUCTION OF DOCUMENTS Licensee hereby submits its response to Intervenor TMIA's First Set of Interrogatories and Request for Production of Documents, which1were hand-served on Licensee on December.30, 1983.

In Licensee's view, many of TMIA's interrogatories are be-yond the scope of the contentions as admitted by the Board, and hence-are beyond the proper scope of discovery. See 10 C.F.R.

S 2.740(b)(1), which requires that discovery shall relate only to these matters in controversy which have been identified by

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. . .'the-presiding officer . . ." In the spirit of cooperation and in-the interest of expeditious completion of these-proceedings, Licensee has nonetheless responded to those interrogatories, with the exception of Interrogatories 23-24, 27-33 and 58 for reasons as more specifically set forth below in its responses. In answering such interrogatories, Licensee is not waiving its right to object to the consideration of such subject matter during the course of these proceedings on the grounds that it is immaterial and irrelevant.

A number of the detailed Definitions and Instructions at pages 1-8 of TMIA's discovery request are objectionable on the grounds that.they are unnecessary, unduly detailed, burdensome, and beyond the scope of reasonable discovery. Licensee has expended considerable time and resources in attempting to respond fully and completely in a manner which will rcasonably provide TMIA with the information requested. Therefore, Li- +

'censee hereBy objects to the information requested in the Defi-nitions and Instructions to the extent they may be construed to request-information beyond that provided by Licensee.

Where the response to a given interrogatory is set out in existing documentation, Licensee frequently has referenced such material rather than repeating the information here. Where possible, references to documents contain section or page numbers. If none appear, the reference is to all of the infor-

.mation in the document. The referenced documents are being provided herewith, provided, however, that where they contain p . .

. o proprietary information to The Babcock & Wilcox Company and/or Licensee, the documents will be provided to TMTA subject to an appropriate proprietary information agreement signed by Licens-ee and TMIA and a Protective Order entered by the Board. Pro-prietary documents are indicated by an asterisk in the document list herein.

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/- I. INTERROGATORIES ADDRESSED TO LICENSEE Interrogatory 1 Describe all tests (including a description of which tubes were tested in the TMI-1 steam generator, and their location Lwithin the~ tube bundle), test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by_Li-censee, including a description of the loading conditions and an explanation as to_how the tests took into account the load time history effects on the tubes, to determine the fatigue life of TMI-1 steam generator tubes:

(a) before kinetic expansion repairs were performed; and ,

(b) after kinetic expansion rgpairs were performed.

Response

(a) A finite element thermal / mechanical stress analysis was performed for the OTSG with characterized tube material thinning (Reference Document 24). All design basis normal operating transients were evaluated. The analysis con-sidered the load (temperature and pressure fluctuatiens) time-history-effects on the tube's fatigue. Tube fatigue usage factors resulting from this analysis were calculated utilizing the procedure specified in ASME Section III for Class I pressure boundary components.

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s a (b) A series of mechanical tests were performed on prototypical mockups (Reference Documents 1 through 11). The test program is described in Section V of Licensee's Topical Report 008, Rev. 3. The test results are discussed in Refer-ence Document 18, Sections 2.5 and 2.6. These tests demon-strated that the " fatigue life" (i.e., the period of time over which the repair would continue to meet the required design goals) was adequate. The original design basis tube loads (Reference Document 24) were.used in these mechanical tests.

These tests covered the leak tightness and confirmed the fatigue axial load carrying capability of the chocen expan-sion technique for simulated and accident conditions, and showed the effect kinetic expansion will have on adjacent repaired tubes, and determined the effect of reexpanding previ-ously expanded tubes (Reference Documents 19, 21, and 22).

Key elements of the qualification program included determining' joint pullout strength (Reference Documents 1 through 10) before and after thermal cycling of the joir.ts and axial load cycling of the joints (Reference Document 15).

l Tube period changes induced by the tube expansion were also tested (Reference Document 12).

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I Interrogatory 2 Describe all tests (including a description of which tubes were tested in the TMI-1 steam generator, and their location within the tube bundle), test data (as defined in 14B, supra),

and analyses' developed and performed by or relied upon by Li-

-censee,. including a description of the loading conditions and an_ explanation as to how the tests took into account the load time history effects on the tubes, to determine the stress levels on the TMI-1 steam generator tubes:

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L 4, . 48 (a) before kinctic expansion repairs were performed; and (b) after kinetic expansion repa rs were performed.

Response

Analyses were performed to determine the tube loads /

stresses for normal-operating and postulated accident conditions (Reference Documents 24; 28; 29, Section 2.1.3; and 30). The loads / stresses'are applicable for both before and

- after kinetic expansion configurations.

Analyses of stresses in the repaired area are de-scribed in Reference Document 29, Sections 2.3 and 2.4. Tests were performed on a full-size steam generator mockup at the B&W Mt. Vernon, Indiana facility to measure the stresses induced in steam generator tubes due to the kinetic expansion repairs.

The. tests.were performed by mounting etrain gages on the steam generator tubes. (Reference Document 19, Section 4.2).

Laboratory tests were also performed to determine the changes in tube preload stresses resulting from kinetic expan-sion. (Reference Document 12).

Interrogatory 3 Describe all tests (including a description of which tubes were tasted in the TMI-1 steam generator, and their location

. within tie tube bundle), test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by Li-consee, including a description of the loading conditions and an explanation as to how the tests took into account the load time history effects of the tubes, to determine the effects of the corrosive contaminant on:

-(a) stress levels on the TMI-1 steam generator tubes; and g-- _

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(b) the fatigue life of TMI-1 steam generator tubes.

Response

(a) The stress levels seen by the OTSG tubes are un- ,

affected by the presence of a corrosive contaminant.

(b) " Corrosive contaminants" in the chemical state or quantity sufficient to initiate or propogate tube cracking are not present in the RCS. Therefore, the fatigue life of TMI-1 steam generator tubes is unaffected.

Interrogatory 4 Describe all tests-(including a description of which tubes were tested in the'TMI-1 steam generator, and their location within the tube bundle), test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by Li-censee, including a description of the loading conditions and an explanation as to how the tests took into account the load time history effects of the tubes, to determine the effects of the changed strength and dimensions of tubes which have been kinetically expanded on:

(a) stress levels on the TMI-1 steam generators tubes; and (b) the fatigue life of the TMI-1 steam generator.

Response

The-testing and analyses described in the replies to Interrogatories 1 and 2 demonstrated the load carrying capabil-

- ity of the repaired tubes. All such tests and analyses were performed with the correct geometries of the kinetically ex-panded tubes and-therefore simulated any changed strength char-acteristics caused by altering tube geometry. (Reference Documents 1 through 10, 12, 15, 19 through 22, 24, 28, 29, Sections 2.1.3, 2.3, and 2.4; and 30 ) .

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e a Interrogatory 5 Describe how the tests (including a description of which tubes.were tested in the TMI-1 steam generator, and their loca-tion within the tube bundle), test data (as defined in 14B, supra), and analyses developed and performed by or relied upon by Licensee described in Interrogatories 1 through 4 took account of crack size in determining:

(a) stress levels on the TMI-1 steam generator tubes, in particular, the effects of thermal stress, and (b) the fatigue life of TMI-1 steam generator tubes.

Response

In addition to the tests and analyses described in the responses to Interrogatories 1 through 4, the effect of crack size on stress lev ~els and the fatigue life of TMI-1 steam generator tubes was taken into account as follows:

(1) A fatigue evaluation in accordance with Section III of the ASME Code was perforraed with a fatigue strength re-duction factor of 5 to account for the presence of a crack.

The crack cross-section was utilized in the analysis to account for the size of the crack. (Reference Document 31, Sections VIII-D)

(2) The load carrying capability of a cracked tube for a postulated main steam line break design basis accident was calculated in Section VIII-B of Reference 31. The crtek cross-section was utilized to account for the size of the crack. See also Licensee's Topical Reort 008, Rev. 3, pages 82-84.

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(3) The fatigue crack growth was calculated in ac-cordance with the methodology of Appendix A of Section XI of the ASME Code. This methodology utilized n fracture mechanics analysis-technique which accounted for the effects of crack size. (Reference Document 29, Section 2.1.4; Reference Document 31, Sections VIII-C and D; and Topical Report 008, Rev. 3, pages 82-84).

Interrogatory 6A Describe how the tests (including a description of which tubes were tested in the TMI-1 steam generator, and their loca-tion within the tube bundle), test data (as defined in 14B, supra), and analyses developed and performed by or relied upon by Licensee described in Interrogatories 1 through'4 took account of crack location in determining:

(a) stress levels on the TMI-1 steam generator o tubes; and (b) the fatigue life of TMI-1 steam generator tubes.

Response

! See response to Interrogatory 5. The crack was as-1 I

sumed to be at the location on the tube which will result in the highest tube stresses and least fatigue life (i.e., in the top span where the cross flow is highest). (Reference Document l 29, section 2.1.3).

Interrocatory 6B I

Describe and state the basis for determining the-failure mode of:

(a) the kinetically expanded tubes; and (b) the TMI-1 archival tubes.

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Response

Failure mode of the kinetically expanded joint in this context is taken to mean one or more ways in which the joint would fail to (a) carry axial load or (b) sufficiently

' limit leakage.

The failure mode was demonstrated by test to be fail-ure by slipping due to application of sufficient axial force on the tube. The testing, which included TMI-1 archival tubes (representative of those in the OTSG), demonstrated that the load value at which failure occurred was greater than the maxi-mum design basis tube axial load value of 3140 pounds for a postulated main steam line break accident. Influences which could reduce the joint's ability to carry axial load or limit leakage, such as exposure to cyclic heating and cooling, expo-sure to cyclic axial loading, and testing at elevated tempera-

.ture, were included in the testing. (Topical Report 008, Rev.

3,Section V.C).

Leakage testing, likewise, indicated that the repair was soundrin as much as the tested leak rates, taken to the

-statistical upper bound and extrapolated to the number of tubes in'the steam generators, were much less than the Technical Specification criteria. Actual leak tests on the repaired steam generators have demonstrated that acceptable leakage rates have been achieved. (Topical Report 008, Rev. 3, App. A,Section II.A). See also responses to Interrogatories 1, 2 and 4.

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, ' Interrogatory 7-State the basis for using an axial lead of 1110 lb./100*F during the corrosion test referred-to in TDR-008. ,

Response

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The. axial load of 1100 pounds represents the maximum inormal operating transient tube load from generic analyses.

13 tis load results from a design basis reactor coolant system

,7 cooldown. -

. Interrogatory Q

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Has Licensee determined or quantified the specific load transfer-for testing purposes? If the answer is "yes", de-scribe Lnd. state tne ba' sis for determining the load transfer under:

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ideal conditions; and '-

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(b) service conditions. - '

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Response

Ye,s . It is' assumed the term " ideal conditions" -

refers to laboratory conditions and the term "sertice conditions"-refers to operational conditions in the generator.

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The transfer of load from the tube to the tubesheet was found during testing to be a function of kinetic expansion-length and tube and tubesheet surface conditions.  ; .

Prequalification testing (Reference Document 27)-pro ' ,

u 'vided:information for the optimization of expansion length. AJ six-inch effective length for the kinetic expansion was . i nelected for qualification testing. (Reference Document 18,

'Section 2.2.1).

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f Qualification testing utilized oxidized tubes and v .

tubecheats to simulate that found on the tubes and tubesheets at TMI-1. (Reference Document 16). Testing was performed to

, CD determine the relative effect of oxide thickness on kinetic ex-pansion joint integrity. (Reference Document 10). In this testing relative slip loads and leak rates for tubes with various oxide thicknesses were compared with those for clean tubes.

The hot functional testing which tested the joints in "servica conditions" is described in Reference Document 30, and included heatup and cooldown transients as well as operation at saturated temperature steady-state conditions.

Interrogatory 9 Describe all tests (including a description of which tubes were tested in the TMI-1 steam generator, and their location within the tube bundle), test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by Li-censee, to test plasticity failures, including a description of which tubes'have been tested for plasticity failures, and a de-scription of the plasticity failure analysis for all sections of the tubes. If only certain tubes were tested, state the basis for selecting the tubes which were tested, and for failing to select others.

Response

The tests and analyses which cover potential plastic-4 ity failures for all sections of the tubes are presented in Reference Document 31, Section VIII-B. The following is a brief summary of this referenced information:

i

a. Attachment 1 of Section VIII-B illustrates a method (based on tests on 304 stainless steel reported in the literature) for calculating ultimate stress at the plastically failed section of a cracked tube or plate.

b. Attachment 2 of Section VIII-B compares the calculated ultimate stress for plastic failure of a cracked Inconel 600 tube (per Attachment 1) with B&W test results for an actual section of TMI-1 tube (which was removed from the "A" generator location No. 13-63) and shows good agreement,

c. Attachment 3 of Section VIII-B calculates the minimum intact cross-section area (to prevent tube failure for a MSLB design basis accident) of a TMI-1 tube to be 64 percent if the defect is within the upper (or lower) tubesheet.

d. Attachment 4 of Section VIII-B calculates the minimum intact cross-section area to prevent tube failure for an MSLB design basis accident of a TMI-1 tube to be 72 percent if the defect is between tubesheets at TMI-1.

It was not considered necessary to test other tubes in TMI-1 because tube 13-63 from the "A" generator was consid-ered representative of other degraded tubes. Also, testing of tube 13-63 was only considered confirmatory in nature-since the calculational method used is considered adequate even without the testing of tube 13-63.

The results are summarized in Topical Report 008, Rev. 3, page 84.

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Interrogatory-10  !

Describe all tests (including a description of which tubes were-tested in the TMI-1 steam _ generator, and their location within the tube bundle), test J.ata (as defined in 14B, supra),

and analyses developed and performed by or relied upon by L1-consee, to determine the effects of a-simultaneous rupture in both steam generators, including a simultaneous rupture in both steam generators occurring in conjunction with a LOCA.

Response

i Because postulation of a simultaneous tube rupture in

• both steam generators is not a design basis accident under NRC requirements, Licensee's tube repair program relies on no tests

. or analyses for determining the effects of such an event.

Interrogetory 11

_Po you claim that in the course of corrosion, some tubes failed. earlier than others? State the basis for your answer.

Response

No evidence has been uncovered which indicates that some tubes failed earlier than others. All tube failures occurred between the hot functional testing-in September 1981 and system pressurization in November 1981. (See Reference

- Document 33 Section III). This analysis is consistent with our understanding of the cracking mechanism and the facts sur-

- rounding the hot-functional testing in 1981. During hot func-tional testing, had tube failures. existed, leakage from primary

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to secondary sides would havs been detected when the RCS was at full pressure. No such leakage was reported.

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In addition, the circumferential orientation of the

-defects could only have been caused by an axial stress. At operating pressure, the maximum stress is in the hoop direction and this would have produced a longitudinal crack. Therefore, the orientation of the cracks dictate that the cracks occurred when thel system was depressurized or depressurized to the point where the axial tube loads exceeded the pressure loads. This condition existrid only during cooldown and cold shutdown fol-lowing the hot functional testing. (Reference Document 44).

Extensive tests conducted by B&W and others. confirmed that cracking of the type observed in the TMI tubes can be produced at ambient temperatures in aeratsd thiosulfate contam-inated borated water. (Reference Documents 45 and 46). Such cracking, depending on the environmental and stress parameters, would occur somewhere between ten and several hundred hours.

Interrogatory 12 If your answer to the above interrogatory is "yes":

(a) Describe all tests, (including a description of which tubes were tested in the TMI-1 steam generator, and their location within the tube bundle), test data (as defined in 14B, supra), and analyses developed and performed by or relied upon by Licensee to determine why some tubes failed earlier than others.

(b) Describe precisely how pre- and post-repair testing and analyses of the TMI-1 steam generator took account of the fact that certain tubes had failed earlier than others.

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Response

Not applicable. See response to Interrogatory 11.

Interrogatory 13 Do you claim that the repaired tubes have been returned to the original design basis?

(a) If the answer is "no", state the basis for determining _this, and explain precisely why this is or is not safety significant.

(b) If the answer is "yes", state the basis for this determination.

Response

Yes. The basis for this determination is set out in Section V of Licensee's Topical Report 008, Rev. 3.

Interrogatory 14 Did Licensee perform the "Rockwell hardness test" on any corroded tubes before the expansion repair?

(a) If the answer is "no", state the basis for not to perform the "Rockwell hardness test" on cor-determining'before roded tubes the expansion repair.

(b) If the answer is "yes", describe the test performed, which tubes were tested, and test data (ac defined

! in 14B, supra), and analyses developed, including a description of the loading conditions for the test.

Response

i j No. The hardness of a corroded TMI-1 steam generator l

tube sample was measured by alternate means at the Babcock and Wilcox Company, Lynchburg Research Center. (Reference Document 23, Section 2.14). Microhardness was measured as a function of

! distance from the top of the sample. These measurements were l

L l

made ever approximately two inches of tube wall at the top of Sample BlO-29.

Measurements were made with a Wilson Tukon hardness tester equipped with a Diamond Pyramid indenter and operated with a 100 gram load. Hardness measurements were made 1-1/2 mils in from both the inside and outside surfaces and at mid-wall along the length of the sample.

Interrogatory 15 Did Licensee perform the "Rockwell hardness test" on cor-roded tubes after the expansion repair?

(a) If the answer is "no", etate the basis for determining not to perform the "Rockwell hardness test" on cor-roded tubes after the expansion repair.

(b) If the answer is "yes", describe the test performed, which tubes were tested, and test data (as defined in 14B, supra), and analyses developed, including a description of the loading condit.4ons for the test.

Response

No' . Additional hardness data on the corroded TMI-1 steam generator tubes was not required to support any conclu-sions concerning the effectiveness or adequacy of the repair l process.

Interrogatory 16 l

Did Licensee perform tests to evaluate the " toughness" of l corroded tubes before the expansion repair?

H (a) If the answer is "no", state the basis for determining not to perform tests to evaluate the " toughness" of l corroded tubes before the expansion repair.

(b) If the answer is "yes", describe the test perforced, which tubes were tested, and test data (as defined l

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in 14B, supra), and analyses developed, including a description  !

of the loading conditions for the test. '

Response

Licensee performed tensile load and elongation tests on tube specimens removed from locations numbered 13-63 (crncked) from Steam Generator A and 33-30 (uncracked) from Steam Generator B. The yield strength, ultimate tensile strength and elongation results confirmed that the material toughness had not been degraded (Reference Document 33, Section 2.16; Reference Document 31, Section VII-B, and Reference Document 33).

Interrogatory 17 Did Licensee perform the tests to evaluate the " toughness" of corroded tubes after the expansion repair?

(a) If the answer is "no", state the basis for determining not to perform tests to evaluate the " toughness" of corroded tubes after the expansion repair.

(b) If the answer is "yes", describe the test performed, which tubes were tested, and test data (as defined in 14B, supra), and ann.i.yses developed, including a description of the loading conditions for the test.

Response

Confirmatory bending tests were performed on samplos from a kinetically expanded corroded tube. The tube samples were removed from location 78-32 of Steam Generator A. This test indicated thct failure was ductile. (Reference Document 18 Section 2.6.4; Reference Document 32).

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Interrogatory 18 l

Describe the relationship between the pre-repair "Rockwell hardness test" and tests to evaluate " toughness." j Responsg There is no relationship between the "Rockwell hardness test" and tests to evaluate "tcughness."

Interrogatory 19 Describe the relationship between the post-repair "Rockwell hardness test" and tests to evaluate " toughness."

Response-There is no relationship between the "Rockwell hardness test" and tests to evaluate " toughness."

Interrogatory 20 Describe the " lead test" program, including an explanation asLto whether the tube samples being used for the lead test program have been expanded prior to the testing sequence.

Response

The lead test program, described in Reference I

Document 47 and in Licensee's Topical Report 008, Rev. 3, pages 25-27, has been designed to test OTSG tubing under conditions representative of those that will exist in the OTSG's during l

future operations.

Two types of full-section tube specimens are being used-- . The first, described a " lead" specimen, includes tubes which have not been expanded. Some lead cpecimens have been l

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i exposed to expansion products as designated below. These tube

. sections are representative of the unrepaired portions of the tubes below the upper tubesheet. Lee.d tube specimens with and without addy current indications are being used.

'The second type of full tube specimens are designated repair. specimens. These tubes have been explosively expanded into simulated OTSG tubesheets (Id. at 47, Figure 2) prior to testing. The expanded region and transition are representative of those areas in the upper tubesheet of the OTSG's.

Specimens being used in the long term tests have been exposed to:the chemical protective agent and products of the explosive expansion. (Reference Documents 47, Table 3, 48, and 49).

Ir.terrogatory 21-State the basis for determining that the " lead test" pro-gram will detect tube degradation.

Response

The lead test program has been designed to duplicate the conditions of-material, stress and environment that will actually be present in the as-repaired OTSG's. The program is

-described in Reference Documents 47 and in Topical Report 008, Rev. 3, pages 25-27.

Materials are actual OTSG tube sections. These tube sections'are being exposed in both the unexpanded condition, t

simulating the unrepaired portions of the tubes, and in an expanded condition identical to that used in the actual repair.

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Some tube sections have been further prepared by ex-posure to the hydrogen peroxide sulfur removal process used in the plant. Also, the form and concentration of surfact deposits in the lead tests is representative of that in the actual OTSG tubes.

Stresses in full tube specimens are representative of those generated in OTSG tubes during normal operations including heatup,. operation, cooldown and shutdown (Reference Documents 50, 51). In addition, specimens of actual OTSG tubes are statically loaded to near yield strength - significantly higher than normal operating stresses.

Environment is contolled to be representative of the environment that will be experienced during operation. Syn-thetic reactor coolant containing the maximum amount of contaminants allowed by specifications (Reference Document 52) ,

is being used. Oxygen levels appropriate to the operating modes are used. Temperatures and heatup/cooldown rates are representative of those actually experienced in the plant.

Interrogatory 22 Describe the-smallest crack opening displacement de-tectable by your methods.

Response

'During the performance of the long-term test program, eddy current testing (ECT) will be the most sensitive non-destructive test employed. The ECT sensitivity is discussed in p- -

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4 Licensee's Topical Report 008, Rev. 3, pages 78-81.

Metallography will be performed at the completion of each test cycle. The resolution of the equipment to be used will be better than .0001." -

Interrogatory 23 Describe and state the basis for deciding not to use non-linear fracture mechanics theory and analysis for testing and analyzing the residual tube properties in the TMI-1 steam generators, including the properties of those tubes which were circumferentially cracked?

Interrogatory 24 If you claim that non-linear fracture mechanics analysis is inappropriate for testing and analyzing the residual tube properties in the TMI-1 steam generators, including the properties of circumferentially cracked tubes, describe and state the basis for this determination.

Objection to Interrogatories 23 and 24 Licensee objects to Interrogatories 23 and 24 on the grounds'that they are beyond the scope of the admitted conten-tions in this proceeding. In its Memorandum and Order of November 29, 1983, at.9-10, the Board expressly struck from t

TMIA's proposed Contention 1.d the issue of whether non-linear fracture mechanica theory should have been used.

- Interrogatory 25 Describe the empiricr.1 data which exists to support the fracture mechanics calculations used by Licensee to test the residual tube properties in the TMI-1 steam generators,

! including the properties of circumferentially cracke d tubes.

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Response

Empirical data which supports the fracture mechanics calculations used by the Licensee is presented in Topical

-Report 008, Rev. 3, Fig. IX-4, and Ref. 29. This data was uti-lized-to develop the constants in the Paris equation (presented in Ref. 29, pg. 8).

i Additional eepirical data is present in the attached figure which is contained in Janes, L.A. " Fatigue Crack Propa-gation Behavior of Inconel-600," International Journal of Pressure Vessels and Piping, Volume 5, 1977.

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TMIA INTERROGATORY NO. 25 Response: Fig. No. 1

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9 Specimen 183, R = 0.050 -

-d A Specimen 185, R = 0.333 * -

E Specimen 184, R = 0.600

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- Annealed Inconel 600 g -

tested at 400 cpm in 3 -5 air at 8000F (4270C) g g 10 ,,

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Figure 6-4 The effect of stress ratio upon the fatigue crack growth behavior of annealed Inconel 600.

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i Prepared by EPRI based on data contained in James, L. A., Fatigue Crack Propagation Behavior of Inconel 600, International Journal of Pressure Vessels and Piping, Vol. 5, 1977.

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c ., . . . .- _.

I[terrogatory 26 Describe and state the factual basis for determining which type of plug has been used in the various locations of the TMI-1 steam generators.

Response

If the expanded portion of a tube was to be plugged, and the tube was to be stabilized, a B&W welded plug with a stabilizer fitting was used. Westinghouse roll plugs were used for plugging the expanded portion of all other tubes.

Interrogatory 27 Describe and state the factual basis.for determining the effect which the plugging of 1,500 tubes would or will have on load distribution.

Interrogatory 28 Describe all tests (including a description of which tubes were tested in the TMI-l stsam generator, and their location within the tube bundle), test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by Li-censee to ddtermine the effects which plugging 1,500 tubes will have on load distribution.

Interrogatory 29 l

l Describe and state the factual basis for determining the revised load for an individual tube after it has been plugged.

Interrogatory 30 Describe all tests (including a description of which tubes were tested in the TMI-l steam generator, and their location within the tube bundle), test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by Li-censee to determine the revised load for an individual tube l after it has been plugged.

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Interrogatory 31 Describe and state the factual basis for determining how the new load distribution referred to in the preceding in'ter-rogatory affects the fatigue life of an individual tube.

Interrogatory 32 Dascribe all tests.(including a description of which tubes were tested in the TMI-1 steam generator, and their location within the tube bundle), test data, (as-defined in 14B, supra),

and analyses developed and performed by or relied upon by Li-censee to determine how this new load distribution referred to in the preceding interrogatory affects the fatigue life of an individual tube.

Interrogatory 33 Describe precisely how pre- and post-repair testing and analyses of the TMI-1 steam generator took account of the new load distribution identified in the preceding interrogatory.

Objection to Interrogatories 27-33 Licensee objects to Interrogatories 27 through 33 on the grounds that they are beyond the scope of the admitted con-tentions in,this proceeding. Contention 1.c, the only conten-tion related to the plugging of steam generator tubes, express-ly does not encompass an allegation with respect to the effect of-plugging on tube load or load distribution. See Memorandum and Order, January 9, 1984, at 4-5.

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Interrogatory 34 Describe in precise detail the history of plug retention at TMT-1. Include the complete failure analysis history.

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Response

There have been no instances of failure of a plug to i be retained within a tube at TMI-1.

Interrogatory 35  !

Describe all tests (including a description of which tubes were tested in-the TMI-11 steam generator, and th9ir location within the tube. bundle), test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by Li-consee-to determine the plug retention capability of tuben which suffered corrosion damage and had been expanded berfore plugging.

Response

A. Welded Plugs The welded plugs used on the expanded portion of TMI-l steam generator tubes were designed to be welded to the existing tubesheet weld after the protruding tube end is machined off to 0.030" from the surgace of the tube sheet, similar to Design I. Analyses were undertaken to confirm the integrity of the existing tubesheet welds as a' connection point 3,

for the' welded plugs. Metallographic examinations of tube sam-ples indicated that axial and circumferential cracking of the tube ends extends typically into the portion of tube behind the seal weld. For the tube end behind the seal weld, fine IGSAC cracks have been found near the heat affected zone (HAZ) and L -down in the inside surface of tube approximately 3/8" from the ,

top tube-end. Although IGA has been noted in the HAZ area, the metallurgical: analysis also shows that at least 50% of the tube ,

- Larea-in each polished specimen is not affected by the IGA (Ref.

i 16). The maximum weld penetratioh into the tube wall is about 50%. Characteristically, the bottoms of the cracks are se-verely blunted, indicating cracks arrest within the weld pene-tration zone (Ref. 17)< It is also concluded by the above analyses and tests that the original tube-to-tubesheet weld and the cladding on top of the UTS were not cracked due to the IGSAC. From the metallurical point of view, this can be explained that the cast structure (welds and weld deposits cladding) of the alloys is expected to be more resistant to corrosion attack than wrought structures in the tubes.

B. Roll Plugs The roll plug is designed to achieve a leaktight joint by mechanically colling the plug over the original roll region for the tube end, which extends below the tubesheets outer surface.

Metallographic examination of removed tube samples as well as fibdrscopic examination of tubee in the OTSG indicate axial and circumferential cracking in the tube and down to the elevation of the tubesheet weld and circumferential cracking in the roll transition region (Reference Document 53).

There were no cracks identified in the roll region below the elevation of the new weld. The potential impact on roll plug performance due to a circumferential crack in the tube at the elevation of the seal weld was investigated in the Westinghouse qualification program by preparing a test specimen with a simulated circumferential crack (36 machined) at the

~

i .

elevation of the seal weld. The test specimen underwent thermal cycling and subsequent leak testing. Test results in-dicated performance within the plug design criteria.

Acceptable perfcrmance of roll plugs insertad in ki-netically expanded tubes was confirmed in several ways. Li-censee concluded that the material properties of the area of the expanded tube below the weld had not been adversely affect-ed by the kinetic expansion based on crack growth tests, which showed that the expansion had no significant effect on existing cracks or the microstructure in the area of the cracks. Refer-ence Document 18 at Section 2.6.4. Hardness tests further demonstrating that the kinetic expansion produced less hard-ening working of the repaired area than the rolling process itself. Id. at 5 2.8.11 and Topical Report 008 page 40. In addition, a second laboratory test was conducted by Westinghouse using tube and tube sheet tect blocks. Kinetic expansions Vere performed by Foster Wheeler on the test blocks and the pre- and post-expansion tube I.D. measured. The aver-age change in tube I.D. (.00034 inch) was extremely small.

Third, no significant leakage from tubes plugged in the kinetically expanded region was found during final bubble testing or hot functional testing.

Interrogatory 36 Describe and state the basis for determining acceptable leakage for plugged tubes.

Response

The.two plug types referred to in the response to In-terrogatory 26 have different design leak characteristics. The welded plugs were designed for zero leak rate, while the me-chanical roll plug was designed with an acceptance criteria of 1.6 drops per minute per plug maximum leak rate. This criteri-on was derived based on'an assumed total of 4400 roll plugs installed in the steam generators and a limit of 0.1 gpm total leakage for both. steam generators during plant operation.

Interrogatory 37 Describe the number and. location of plugged tubes which were identified as leaking unacceptably, requiring further repairs, during post-repair testing.

Response

Of the tubes which were plugged in the expansion region, only six were identified as leaking during post-repair testing and* repaired. They are in the following locations:

A-135-72 A-113-03 A-138-69 A-69-126 A-124-95 A-148-38 Interrogatory 38 i Describe and state the factual basis for determining which l tubes had unacceptable leakage identified during post-repair

! testing.

l f

R~p' nm Licensee performed a preliminary bubble test in May 1983, and a final bubble test on June 17 in the "B" generator and on June 26, 1983 in the "A" generator to determine which of the tubes that had been plugged in the Kinetic expansion region were leaking. The test procedure was the same for each test and is decribed in Topical Report 008, App. A, pages 108 and 115.

Interrogatory 39 Describe all tests (including a description of which tubes were tested in the TMI-l steam generator, and their location within the tube bundle), test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by Li-censee to determine the cause of the lack of integrity of these leaking plugs.

Response

Of the six leaking plugs identified in the response to Interrogatory 37, five were B&W welded plugs, the other a Westinghouse rolled plug.

Visual inspection of the welded plugs determined that the cause of the leakage was a defective welds. The plugs were replaced. Visual inspection indicated that the new welds were satisfactory, and subsequent leak testing demonstrated that the new plugs did not leak.

The rolled plug was removed and replaced by a new rolled plug. Bubble testing thereafter indicated no leakage, leading to the conclusion that the condition of the host tube was not the cause of the leak.

- . . _ . . . . , . . _ . _ - - - . ~ . . . . , - - - - - - - - - - , - - - . . . ~ . - . - - , - - - - , . - - - - -

I

In addition, the QA records packager for plug instal-lations were reviewed.

Interrogatory 40 Describe all tests (including a description of which tubes were tested in the TMI-1 steam generator, and their location within the tube bundle), test data (as defined in 14B, supra),

and analysis developed and performed by or relied upon by Li-censee to determine whether plugged tubes which have undergone the kinetic expansion repair process will interfere with the plant's ability to respond to transients and accidents?

Response

Licensee has confirmed that the plug qualification programs for plugs in unexpanded tubes remain valid for ex-panded tubes. See response to Interrogatory 15.

Interrogatory 41 Describe mechanistically your definitions of safety and safety significant as they relate to steam tube repairs and failures.

Response

Safety and safety significance are used in the context of meeting the licensing basis for steam generator tubes.

Interrogatory 42 Do you claim that "a turbine trip at maximum power" could result in stresses sufficient to cause a rupture of the repaired portion of a steam generator tube? Describe all tests, (including a description of which tubes were tested in the TMI-1 ster.m generator, and their location within the tube bundle), test data (as defined in 14B, supra), and analyses developed and performed by or relied upon by Licensee to deter-mine this.

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(a) If the ansvar is "yes", describe the signifi-cance of these stresses, state the basis for your determina-tion, and explain precisely why this is or is not safety sig-nificant.

(b) If the answer is "no", state the basis for this determination.

Response .

A turbine trip at maximum power will not result in stresres sufficient to cause rupture of the repaired portion of a steam generator tube.

A turbine trip will result in an automatic reactor trip, and the plant will be stabilized at reactor coolant conditions which are comparable to " hot standby" conditions (RCS temperature at or above 532*F). This results in less tube load than for a design basis cooldown transient. Thus, signif-icant changes in the OTSG shell to tube temperature difference and primary and secondary pressures from the power operating conditions are not produced as a result of a turbine trip.

The OTSG tube axial loads and stresses which are induced by a turbine trip at maximum power would be lower than and would be bounded by those loads / stresses which have been utilized to evaluate the integrity of the repaired OTSG tubes.

Interrogatory 43 Do you allege that " thermal shock from an inadvertent actuation of emergency feedwater at high power" could result in stresses sufficient to cause a rupture of the repaired portion of a steam generator tube? Describe all tests, (including a description of which tubes ware tested in the TMI-1 steam gen-erator, and their location within the tube. bundle), test data

-(as defined in 14B, supra), and analyses developed and performed by or relied upon by Licensee to determine this.

(a) If the answer is "yes", describe the signifi-cance of these stresses, state the basis for your determina-tion, and explain precisely why this is or is not safety sig-nificant.

(b) If the answer is "no", state the basis for this determination.

Response

Thermal shock from an inadvertant actuation of emer-gency feedwater at high power will not result in stresses suf-ficient to cause a rupture of the repaired portion of a steam generator tube.

The term "inadvertant actuation of EFW" is understood to mean a single non-mechanistic failure that results in starting of the emergency feedwater pumps while the plant, in all other respects, is operating normally at high power. This will not result in the injection of emergency feedwater into the steam generator.

The design of the TMI-1 EFW system is such that once the EFW pumps are initiated, the actual flow to the OTSG's is controlled by valves which respond to a flow demand signal gen-erated by the OTSG level control system. The water level in the OTSG at high power levels is much higher than the OTSG EFW level setpoint at which the EFW flow centrol valves are initi-ated to open. The EFW pumps are initiated by signals other than and independent of the OTSG level. Therefore, inadvertant actuation of the EFW pumps will not result in EFW injection into the OTSG and will not result in any change to the OTSG tube stresses.

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Even if EFW injection into the OTSG were to occur, the resulting thermal stresses would not result in stresses sufficient:to cause rupture of the repaired portion of a steam generator tube. The location of any thermal shock stress

. condition, due to impingement of cold water that could occur on a tube the was repaired, would be remote from the repaired

. portion of the tube. Thus, the direct thermal shock stress effects would only affect a portion of the original tube. The only affects of the actuation of emergency feedwater at high power would be a slight decrease in the average tube tempere-ture. Consequently, only a slight change in tube load would occur. This is based on the tube temperatures and pressures analyzed in Reference Document 24.

t Interrogatory 44 Do you allage that " rapid cooldown" following a LOCA could result in strecses sufficient to cause a rupture of the repaired portion of a steam generator tube? Describe all tests, (incruding a description of which tubes were tested in the TMI-l steam generator, and their location within the tube bundle), test data (as defined in 14B, supra), and analyses developed and performed by cur relied upon by Licensee to deter-mine this.

(a) If the answer is "yes", describe the signifi-cance of these stresses, state the basis for your determina-tion, and explain precisely why this is or is not safety sig-nificant.

(b) If the answer is "no", state the basis for this determination.

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Response

A " rapid cooldown" following a LOCA will not result in stresses sufficient to cause a rupture of the repaired por-tion of a steam generator tube. This conclusion is based upon the LOCA event describesd in Reference Document 24 which includes a rapid cooldown.

The corresponding tube load value determined in Ref-erence Document 24 is 2641 pounds. The kinetic expansion repaired portion of a steam generator tube has been qualified by testing (see responses to Interrogatories 1 and 2), to loads in excess of 3140 pounds (the main steamline break condition).

Therefore, since the repaired portion of a steam gen-erator tube has been qualified for a much higher loading condition, the repaired portion will maintain its structural integrity for a " rapid cooldown" following a LOCO condition.

Interrogatory 45 Do you allege that " rapid cooldown" following a LOCA could result in stresses sufficient to cause a rupture of the repaired portion of a steam generator tube? Describe all tests, (including a description of which tubes were tested in the TMI-1 steam generator, and their location within the tube bundle),-test data (as defined in 14B, supra), and analyses developed and performed by or relied upon by Licensee to deter-mine this.

(a) If the answer is "yes", describe the signifi-cance of these stresses, state the basis for your determina-tion, and explain precisely why this is or is not safety sig-nificant.

(b) If the answer is "no", state the basis for this determination.

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l

Response

A " rapid cooldown" following a LOCA will not result in stresses sufficient to cause a rupture of the repaired por-tion of a steam generator tube. This conclusion is based upon l l

the LOCA event described in Reference Document 24 which ,

l includesarapjdcooldown. l The corresponding tube load value determined in Ref-~

erence Document 24 is 2641 pounds. The kinetic expanslon repaired portion of a steam generatIor tube has been qualified by testing (see responses to Interrogatories 1 and 2), to loads in excess of 3140 pounds (the main steam line break condition).

Therefore, since the repaired portion of a steam gen-erator tube has been qualified for a much higher loading condition, the repaired portion will maintain its structural integrity for a " rapid cooldown" following a LOCA condition.

Interrogatofy 46 Do you claim that " restart" could result in stresses suf-ficient to cause a rupture of the repaired portion of a steam generator tube? Describe all tests, (including a description of which tubes were tested in the TMI-I steam generator, and their location within the tube bundle), test data (as defined in 14B, supra), and analyses developed and performed by or relied upon by Licensee to determine this.

(a) If the answer is "yes", describe the signifi-cance of these stresses, state the basis for your determina-tion, and explain precisely why this is or is not safety sig-nificant.

(b) If the answer is "no", state the basis for this determination.

4

B O

Response

" Restart" of TMI-1 will not result in stresses suffi-cient to cause rupture of the repaired portion of a steam gen-erator tube. The expansion joint has been qualified to le, ads /

stresses far in excess of those involved in the " restart" process. See responses to Interrogatories 1 and 2.

Interrogatory 47 Describe and state the basis for determining that the cor-rosive environment has been eliminated.

Response

The basis for determining that the corrosive environ-ment has-been eliminated is described in Licensee's Topical Report 008, Rev. 3, Sections III.D, IV, IX.A. The source of the stress corrodant of the tubing was sodium thiosulfate and this has been ecmpletely removed from the system. The

, byproduct, nickel sulfide, was partially removed from the system and no demonstrated risk has been identified with small quantities of. nickel sulfide remaining on the tube surfaces.

Numerous steps have been taken to insure corrosion reinitiation will not occur. These steps include:

l 1. Reduce the amount of sulfur species in the RCS l

via the peroxide cleaning. (Reference Document 59).

2. Administratively controlling sulfate, chloride and fluoride to below 100 ppb each and establishing analysis programs to assure this. (Reference Document 52).

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3. Assuring lithium levels are at the higliest per-mitted concentrations within Technical Specifications. limits to-

-take advantage of lithium's corrosion inhibiting effect. (Ref- )

erence Document 52). ,

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4. Running extensive long term corrosion'tssts y N 3

simulating actual. reactor operation to insure that sulfur ,

induced problems that might occur are understood and evaluated. 4 To date these tests have not revealed any corrosion problems. ,

. 4 (Reference Documents 47 and 58). <

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5. Running extensive short term corrosion tests to ,, '

characterize the cracking parameters and to help formulate.lthe ,.

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failure scenario. (Reference Documents 45, I6, 56'S'ection I-

. +l 2.16, .57, 62).- "

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6. Evaluate the sulfer-water system thermodynamics.

, t (Reference Document 55). It is concluded.from thin svaluation ,

. that under reducing conditions nickel sulfids on they tibe; sur[

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faces is stable and will not cause cracking.c o ,o -

e s-Interrogatory 48 , 1

y Do you claim that the concerns raised by Mr. Gillon with j ,

' respect to fears that the clean up process may cause the corro . '

sion which damaged the steam generatofs v.o reinitiate, have anya continuing relevancy now that the c. lean up has been completed? ~ '

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. . (a) If the answer is "yes",.Elescribe the relevancy, '

state the basis for your determination,2and explain precisely-  %

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why this is or is not safety significant. 3 .. -

.(b) If the answer is "no", state the basis for this determination.

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s o Response

'4 No. The concerns raised by Mr. Dillon have no rele-C7 '

vance after the cleanup has been completed. Neither corrosion

' tests (Reference Document 60) performed during the process de-velopment nor current long term loop tests (Reference Document s

61) have produced any indication that peroxide cleaning process can reinitiate the corrosion process which produced the tube

< failures. Measured leak rates from testing subsequent to the

.3: cleaning provides additional confidence that reinitiation did not occur.

e- ,

/ Interrogatory 49 Do you claim that the sulfur contamination remaining after ythe cleaning process poses no-risk of reinitiation of IGSCC?

(a) If the answer is "yes", describe the effect, y' < ' .

state the basis for your determination, and explain precisely

'why'this is or is not safety significant.

(b) If the answer is "no", state the basis for this determination. ,

Response

E Based on the test data available to GPUN and the sat-

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7 e ^ 'isfactory completion of the hot functional testing, GPUN does O f' not believe that crack reinitiation will occur when the RCS j f f ,. chemistry is.within anticipated operating limits.

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c The results of the short term. corrosion tests deter-B mined that the steam generator corrosion was associated with

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} y, q oxygenated metastable sulfur species (Reference Documents 45 and 46). However to assess that the presence of sulfide

containing oxide films on tube surfaces and contamination of

. the bulk coolant at the maximum specification limits would not produce corrosive attack, Licensee began a series of tests to study long term effects (Reference Document 47). These tests are being conducted by Westinghouse Electric Corp. Research and Development Center. The long term corrosion test program started in October 1982 and is scheduled to run approximately 17 months. (Reference Document 58).

These tests are being conducted to simulate environ-mental and operational conditions which are representative of the worst case chemistry conditions which could exist in the primary system within administrative specification limits. The objective of the test is to verify that the metallurgical, en-vironmental, geometric and surface conditions which existed after kinetic tube expansion in the steam generators and after the peroxide cleaning are not detrimental to tube integrity.

Td date two loops of the long term corrosion tests have been operating for approximately 300 days. The second two i loops which were subjected to peroxide cleaning have been oper-ated for approximately 160 days. Interim examinations of tube specimens and C-rings have revealed no evidence of corrosion initiation, corrosion reinitiation, or propagation of existing defects. It is significant to note that from the beginning C-ring specimens have shown IGA (surface etching effect) ap-proximately one grain deep; in no case has there been any evi-dence of growth of this 1GA. It is also significant to note 4

,-- . . , . . . , -.~, -

J that one autoclave has been operating with 100 ppb of thiosulfate (measure as SO+D4+U) (Reference Document 58).

The long term corrosion tests provide positive evi-dence that return to operation of the steem generators in their current condition poses no threat of corrosion under the established chemistry cotrols.

At this time the hot functional testing has been com-pleted and no known degradation of steam generators has occurred as assessed by leak rate during HFT and subsequent system pressurization. Chemistry specifications were maintained during the hot functional test and posed no particular control problems. Temperatures of 550F were reached, the system was subjected to several cooldown cycles and finally returned to decay heat cooling. (Reference Document 30).

The successful HET as well as the subsequent layup

_ period has ddded significant credence to the work which has been performed to assure safe operation of the reactor as well as confirming the results of the. ongoing long term corrosion test.

Interrogatory 50 l

Do you claim that an inventory of 0.1 ppm sulfate in solution would have no corrosive effect on the steam generator tubes or the RCS?

'(a) If the answer is "yes", describe the effect, state the basis for your determination, and explain precisely why this is or is not safety significant.

i .

(b) If the answer is "no", state the basis for this determination.

Response

Yes. Test data indicates that a concentration of 0.1 ppm sulfate in solution in PWR reactor coolant will have no corrosive effect on steam generator tubes or reactor coolant system materials under shutdown, hot functional testing, or operating conditions. Long term corrosion testing (lead test)

(Reference Documents 47 and 58) being performed at the Westinghouse R&D Center for Licensee is experimental confirma-tion that low level sulfate concentrations will not be corro-sive to RCS materials (including steam generator tubing).

Testing at B&W Alliance Research Center (Reference Document 45) and Oak Ridge National Laboratory (Reference Document 46) provides-additional supporting evidence for the above conclu-sion.

Interrogatory 51 Do you claim that the release of the 20-50% of the sulfur l remaining in the oxide corrosion film will have any corrosive l effect on the steam generator tubes or the RCS?

l (a) If the answer is "yes", describe the effect, state the basis for your determination, and explain precisely why this is or is not safety significant.

(b) If the answer is "no", state the basis for this determination.

Response

No. Most of the sulfur remaining in the oxide corro-sion film is present as nickel sulfide, with the remainder as smaller quantities of iron and other metallic sulfides.

Release.of sulfur from these relatively insoluble compounds, even in the presence of-an oxidizing agent, is slow. Further oxidation to sulfates would occur rapidly once soluble sulfur compounds were formed. The sulfates are removed by normal plant ion exchange processes and a monitoring program provides additional assurance that unanticipated buildups will not occur. Under reducing conditions, such as encountered during power operation, sulfide is the favored form and significant or detectable release is not expected. (See response to Interroga-tory 49).

Interrogatory 52 State the basis for determining that the chemical composi-tion of contaminants which remains'in the steam generators after cleaning is acceptable.

Response

The possible introduction of chemical impurities and the effects of existing levels of impurities are discussed in l TR-008, Rev.'3,Section VI A. Ensuring that contaminants re-maining in the OTSG's after chemical cleaning are at acceptable levels was primarily accomplished by controlling contaminants in materials used during OTSG repairs.

Consumablo materials such es explosivss, polysthylena candles, surface wetting agents, and flush water were required to conform to specified levels for halogens, sulfur, and heavy metals (Reference Documents 34 and 17). Each lot of such materials was individually analyzed to conform with this speci-fication prior to use in the repair operations.

Non-consumables (tools) were required to meet surface cleanliness requirements by specification (Reference Document 34).

The OTSG's were flushed with elevated pH water fol-lowing removal of expansion debris. Control of the chemical species present in the flush water (Reference Document 61) en-sured that soluble contaminants were removed to an hcceptably low level.

The acceptability of existing levels of chemical contaminants has been confirmed by testing. The surface wetting agent which was put on tube surfaces prior to expansion was tested (Reference Document 39) to ensure that it would not cause or accelerate IGSAC.

Short term corrosion tests (Reference Document 62) were performed on actual OTSG tubes that had been kinetically expanded. After expansion these tubes were cinsed to remove loose debris; no effort was made to flush to a condition equiv-alent to that in the OTSG's after flushing. The short term samples were then subjected to the hydrogen peroxjde cleaning process used in the plant.

, - . . . . c.----,,,--- . ~ ,:- -. ~ . - - - - --.- . - , - - - . ~~.,- -- - -

Examination of the short term corrosion test samples after testing in simulated reactor coolant containing high levels of impurities (Reference Document 62, Table 1, page 3) did not reveal any evidence of IGSAC.

Specimens for the long term corrosion tests (Refer-ence Document 47) are being tested as-removed from the OTSG's (before repair), exposed to the chemical wetting agent, and exposed to the explosive expansion process and peroxide cleaning. No evidence of intiation or progression of corrosive attack has been sean to date. (Reference Document 58 and TR-008, Rev. 3, Sections IV and IV-A).

Interroaatory 53 State the basis for determining that the explosive residue remaining in the steam generators after cleaning is acceptable.

Response

The kinetic expansion process was rigidly controlled to assure that any residue remaining after cleaning would be acceptable. Four methods of control were used through,out the development and implementation of the kinetic expansion process. There were: 1) control of all materials introduced into the OTSG's, 2) methods to limit adherance of debris and residue to OTSG surfaces, 3) methods to clean-up debris folloi!-

ing explosive expansion, and 4) specifications to assure that the clean-up process was effective in returning the OTSG's to an acceptable cleanness level. A description of each of these four methods of control is given below:

-- . . . _ - . , _ . - , . - _-., _ __. _.r._ _, ,-e,-- .- ,- -.-.,- -

All materials which were introduced into the TMI-1 OTSG's as a result of the kinetic expancion process were controlled by GPUN specifications (Reference Documents 17 and

34) to limit the introdt!ction of harmful contaminants into the generators. Quality Assurance data packages (Reference Decuments 35 and 36) were prepared for the explosives to document that all explosive materials used complied with the GPUN specifications. Section 2.7 of Reference Document 18 addresses cor. trol of contaminants throughout the kinetic expan-sion project in more detail.

A surfactant was applied to all primary side surfaces of the OTSG's prior to each round of explosive expansions to limit the adherance of debris and residue to primary side OTSG surfaces. Sections 1.4.3 and 2.9.3 of Reference Document 18 describe the use of the surfactant as a precoat in the explo-sive expansion process. A Manufacturing Specification (Refer-i ence Document 37) was prepared and submitted to the surfactant manufacturer. This specification controlled contaminant l- levels. The qualification program demonstrated that the surfactant does not promote damage to steam generator material

, (see Reference Document 38). Additionally, tests were l

conducted as a part of the kinetic expansion program to demon-l strate that the surfactant wculd not promote further damage to sulfur precracked OTSG tube material (see Reference Document 39).

l

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An extensive program was conducted to assess cleaning methods and develop the optimum method to remove debris and residue from the OTSG's following the explosive expansion work.

Reference 9 contains test descriptions and data from this work.

Section 2.9 of Reference Document 18 discusses the development and implamentation of the post-expansion clean-up process.

The clean-up process used consisted of three steps:

1) removal of loose particulate debris by vacuuming, 2) flushing the generators to remove particulate debris and 'resi-due, and 3) use of felt plugs blown through the OTSG tubes to verify that debris was removed. This work is described in Section 2.9 of Reference Document 18.

The flush procedure used at TMI-1 was governed by

! Reference Document 41 operating specification. The flush water quality acceptance critaria were used to determine when the steam generators had been returned to an acceptable cleanliness level. A sdmmary of the results of the OTSG post-expansion flush is provided in Reference Document 42.

Interrogatory 54 Do you claim that sodium thiosulfate, or other sulfur-bearing species resulting in residual sodium thiosulfate, was introduced into the reactor coolant system in:

(a) July, 1980; (b) May, 1981; (c) September, 1981; and (d) any other date in TMI-1 history.

l _. - - . _ - _ _ . . _ . _ _ _ . _ _ _ , - - . . , - . . , . . _ . - - - ~ - . _ _ - -- __

n- ..

Response

-(a)- Yes

(b) Yes (c) Yes L

(d) No Interrogatory 55 l Describe the-precise circumstances of each introduction of l

sodium thiosulfate, or other sulfur-bearing species resulting in: residual sodium thiosulfate, referred to in the preceding interrogatory,.and state _the factual basis for determining theseLeircumstances.

Response See Reference Document 33,Section III.

Interrogatory 56-l Describe ~ steps taken by Licensee to determine the extent

'of damage , and to rid the system of' contamination, in response to the instances of introduction of sodium thiosulfate, or i other sulfur-bearing species resulting in residual sodium b ' thiosulfate,. referred to in the preceding interrogatory, L including whether Licensee reported such instances to the NRC.

t i Response

~

At the times of induction no potential for damage was recognized. Consequently, removal of the sodium thiosulfate l . occurred'through routine use of various exchange purification systems.

Since the 121 and conductivity increases violated no L existing specifications or regulations, no report to the NRC

[ ..

was required.or made.

1 l

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Interrogatory 57 Describe any steps taken by Licensee to determine if damage had occurred within the reactor coolant system as a result of introduction of sodium thiosulfate, or other sulfur-bearing species resulting in residual sodium thiosulfate, referred to in the preceding interrogatory.

Response

See response to Interrogatory 56.

Interrogatory 58 Has Licensee received NUREG-0691? If so, describe the precise circumstances as to when this was received, and de-scribe any steps taken to determine if damage had occurred within the reactor coolant system as a result of the findings contained in NUREG-0691?

Objection to Interrogatory 58 Licensee objects to Interrogatory 58 on the grounds that NUREG-0691, " Investigation and Evaluation of Cracking Incidents in Piping in A Pressurized Water Reactors" does not pertain to e,ither steam generators or sensitized Inconel 600, the material used for the 24I-1 steam generator tubes. The in-terrogatory is therefore not germane to the subject matter of this proceeding.

Interrogatory 59 Lescribe any thiosulfate residual remaining in the system as a result of any previous introduction of sodium thiosulfate, or cther sulfur-bearing species resulting in residual sodium thiosulfate, at the start of September, 1981. State the basis for arriving at this determination.

_ _ . . . _ _ . . - . - . , . _ . - _ _ - , - . ~ - ~.__ _ ,... .. - _ .. _ ..-... , .. - ,. ,,,---. - .. ..-. - =.- . --

7_

i s- .

.Reeponse The best estimate of sodium thiosulfate residual in the reactor coolant system at the beginning of September, 1981,-

-is 1 to 2 ppm as thiosulfate. The bases for this estimate are (a) the experimental determination of the sodium thiosulfate concentrations required to produce the specific conductivity increase observed in May, 1981, and (b) the reduction in specific conductivity observed during the use of a resin coated

-precoat filter to process the system in early August, 1981.

See-Reference Document'33, Sections III and IV.

I l.

Interrogatory 60 Do you disagree with the NRC Staff finding at page 8 of its Safety Evaluation that "the specific mechanistic steps in-

-volved in the sulfur-induced stress corrosion cracking phenome-non have not.been clearly established?"

Response

Re,ad.in context, the quoted statement appears to state the specific reduced sulfur species that caused the IGSAC of the Inconel 600 tubes under cold conditions have not been identified. So construed, Licensee does not disagree with the Staff's ctatement.

Interrogatory 61

-Describe and state the basis for determining:

(a) how cracking occurred at the lower portion of the tubes in.the steam generator; and (b) how cracking terminated.

. _ . . - . - . . , . . . _ , . . . . . . , . _ _ , . ~ . - - .-,.,-,,...m...,.,_,.-.- . . _ . -

- ,- o -

,e ,--.,a--,--e..~,oL-w-' e - , 4 m - -,v<w

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Response

An oxidizing potential favoring the existence of in-termediate, aggressive specifies.of sulfur is necessary to produce the IGSAC seen in TMI-1 OTSG tubes. Screening tests using both actual and archive TMI-l'OTSG tubing (Reference Document 45) confirmed that 1) fully oxidized (sulfate) species will not cause IGSAC and 2) application of an oxidizing potential or introduction of air was necessary to cause cracking when 1 ppm thiosulfate was present.

Testing done during development of the hydrogen per-oxide cleanup process (Reference Document 60) revealed that at sufficiently high oxidizing potentials, no intermediate sulfur species can be detected and corrosion of actual OTSG tubing does not occur.

During cooldown of TMI-l in Septeber 1981, oxygen was introduced to the primary coolant either by venting to the vent

header or bp the introduction of oxygenated water during high-t end low-pressure injection system testing (Reference Document 33). Introduction of oxygen in the presence of metastable, ag-gressive sulfur species resulted in a condition conducive to IGSAC.

The environment in the OTSG was dynamic. Oxidizing potential at any point was changing as unit cooldown proceeded, water level changed, and remaining films above the water level dried out. Under such a dynamic condition, the conditions in individual tubes below the upper tubesheet got to a cracking l

(

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-.....,~..5,.....~...s,,, , .,,~ . , _ y .._ .. ...x-.,..-.... , . , _ , - ,

condition long enough to show the widely scattered IGSAC seen in the lower elevation.

The continuing change in conditions also caused ter-mination of the cracking. As conditions proceeded toward more oxidizing ones, the presence of aggressive intermediate sulfur species became less probable. The combination of this chnage in oxidizing conditions with other dynamic changes, such as di-lution of surface films by the bulk OTSG water, caused the cracking to terminate.

Interrogatory 62 Do you disagree with the conclusion of NRC Staff consul-tant Dr. Digby D. MacDonald that a volatile polysulfur species besides thiosulfate must be present in the RCS?

Response

Licensee does not disagree that a volatile polysulfur specie must have been present at the time of the sulfur attack of the FORV*and waste gas system.

Interrogatory 63 Describe all methods you have used to rid the RCS of the polysulfur species identified by Dr. MacDonald.

Response

Gaseous sulfur species, soluble species and deposited sulfur species in the reactor coolant system have been suffi-ciently removed to preclude further sulfur attack. The methods used included ion exchange purification, flushing, venting and

_ - . . _ . . . .. ....m. ,-- .- ._. _ ._. _ __ --.,...,._.....,.........i

D .

purging, and peroxide cleaning followed by ion exchange purifiction. (Reference Documents 59 and 65). See also Section IV.D of TR-008, Rev. 3.

Interrogatory 64 With respect to contaminants other than sulfur (in its various forms and compounds), describe each contaniment within the RCS or steam generators which could cause or contribute to corrosion.

Response

There are no contaminants present in the RCS or steam generators in quantities sufficient to cause or contribute to corrosion.

Interrogatory 65 Describe the sealing mechanism which, along with expan-sion, you claim will " severely reduce the possibility of free

' circulation of reactor coolant in the upper crevice area which could produce an aggressively corrosive environment." Describe all tests, (including a description of which tubes were tested in the TMI-1 steam generator, and their location within the tube bundlef, test data (as defined in 14B, supra), and analyses developed and performed by or relied upon by Licensee to determine this, including a description of the failure mode.

1

Response

The mechanism is summarized in Topical Report 008, Rev. 3, pages 51-53. See also Reference Document 43.

Interrogatory 66 Describe the mechanism which you claim "... limit [s] the possibility of existence of an aggressively corrosive environ-ment due to the presence of small sulfur depcsits." Describe all tests, (including a description of which tubes were tested in the TMI-1 steam generator, and their location with in the tube bundle), test data (as defined in 14B, supra), and l i wm-- a.- ~ . r ;. 7,r, n w-nm -, - - n -,n ,a ,w.w yaen. ,-,g. . . -.m .. ,r,m- m m w.-~ - w .~r~m ,..~- -

o .

analyses developed and performed by or relied upon by Licensee to determine this, including a description of the failure mode.

Response

See Reference Document 43.

Interrogatory 67 Do you claim that cracking in the weld area occurred at heat up phase? State the basis for your answer.

Response

See Licensee's Topical Reoort, 008, Rev. 3, Sections II.B.2 and II.D.2.

Interrogatory 68 Do you claim that cracking in the transition zone occurred at heat up phase? State the basis for your answer.

Response

No. See Response to Interrogatory 67.

Interrogatory 69 Describe and state the basis for determining what precise radiological health and safety consequences, in terms of both radioactive releases into the environment, and exposure to the population, would result from:

(a) a single tube rupture; and (b) a simultaneous rupture in each steam generator.

_ _..__ _ - . . - . . . _ _ ... . . _ . . . _ , , . . . . . , . . . . . . . . . . . . . _ . . - ~ . . , . , . . - - - . . . - . . _ . . ~ . _ . - . _ , - . _ .-

' Response (a) The basis for evaluating the radiological consequences of a single tube rupture, and the.results of the evaluation, are documented in the Three Miles Island Nuclear Station - Unit 1 Final Safety Analysis Report (updated ver-t sion), Volume 8, Section- 14.1.2.10, Pages 14.1.-27 and 27, at-tached.

(b) Because postulation of a rupture in both steam generatore is not a design basis accident under NRC require-ments, idependent assessments of radioactive releases and expo-sures from such an event were not performed in support of the c2be repair program.

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I TMIA INTERROGATORY 69

• * "MI-1&SAR (13] de=c trated t." the s- core s not r urn to e dticality.

Thus cc. Occlabili- is ass ed.

e. Ccn _usions

/ un that .e reac cr tries a es=ains sub .tical.

/The init' 1 blowGis analy wn s has psin a maxd um ther result cower cf loff cercent of ratId peu ; theraf . , no

• 1 dc=acefaili occur./The 5:axi=um te 'rature *ferenciai that ccc- s in the .eam gener ter dcas nct

duce fxcessive tresses, and st gene at integ-ity is maintaiyad. The er a enmenta' doses are within acecpeacle limits.

l 14.1.2.10 S2 erm Geners:Or Ute Failu e r

l l a. Steam Generster Tube Rupture

1) Identification of Accident The environmental effects associated with steam genera:ce tube leskage and subsequent release to the environment are evaluated in the preceding sections. An evaluation has also been perfc med fer the complete severance of a steam generator tute. For this cecerence, activity contained in the ceactor ccclant uculd be released to the secondary system. Sc e of the radicactive noble l gases and iodine would be released to the atmosphere thrcugh the ecndenser air removal system.

2) Analysis and Results .

In analy + g the censecuences of this failure, the following

' secuence of events is assumed to occur (input parametsrs are shown in Table 14.1-20 and a summary of results is given in Table 14.1-(- 21):

a) A double-ended rupture of one steam generator tute occ.:s with unrestricted discharge f.~m each end.

b) The initial leak rata exceeds the normal =akeup to the Reac:ct Coc1 ant System, and system pressure decreases. nc initial operator action is assu=ed, and a low Reacter C001:n:

System pressure trip will ocer .

c) Follcwing reacter trip, the Recc cr Ccclant Systa= pressure L centinues to decrsase until high-press =e injection is actuated. The ecpacity of the high-pressure injection is l sufficient to cocpensate fer the leskage and maintains bc:h 14.1-27

[

UPDATI-1 L 7/82 l

l

_ - . , . - . - . . _ _ _ n ---.- -

TMIA INTERROGATORY NO. 69 TMI-1/ TSAR

^

pressure and volu=e centrcl cf the Reacter Ccciant System.

Thereaftar, the reactcc is assumed to te eccled dcun and 0

depressuri sd at 100 T per heur.

d) After the Reacter Ccolant System te=perature decreases below the pressurs set peint of the main steam line saintf valves, the secondary side will- depressurize slcwly and the main steam safety valves will du=p steam to the at:cspners.

Cecidown centinues with the unaffected s;eam generater at 1008 y/hr until the ta=percture is reduced to 250'T. I Thereafter, ecoldcun to ambient cenditiens is continued using tha decay heat rs=cval system.

e) Tc11 cuing reacter trip, une turbine step valves will close.

Steam line pressure will increase, cpening the steam bypass

, valves to the ccndanser. The bypass valves actuate at a 1cuer pressure than to the steam safety valves. The reacter ecolant that laaks as a result of the tube failure is condensed in the condenser. Caly the fission prcducts that esespe from the condensate are released to the at=csphers.

The first radicactivity release path during this accident is discharged thrcugh the turbine typass to the condenser and then cut the cendensata vacuum pu=p exhaust. A gas-te-liquid partiticn factor of 10** is assu=2d fer the iodine in the cendenser (10]

(11], but ncele gases are assu=ed to be released directly to the at=csphere.

The tctal dose to the bcdy frem all the Xenen and Krypten released is given in Table 14.1-21. Tha correspending dcss to the thyroid is also tabulated.

'"he seccnd radicactive release path is thrcugh the 20in safety valves er thrcugh the emergency feed pump turcine at=cspheric steam exhaust.

The at=cspheric dilutien is calculated using the disparsien factors develcped in Section 2.5.

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TMIA INTERROGATORY NO. 69 l l

l TM~-1/ TSAR TA3LE 14.1-20 (Sheet 1 cf 1)

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TMIA INTERROGATORY NO. 69 .

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. v. e .o Lcw presst:re trip oce::rs at mi t 3 Total depress:r. stien time Of Reacter C0 clan Sy::em ein 04 Rese:Or c:clant leaksce during 3

depresst-1:s icn f 1977

/.ctivity Released to A =0:phere Noble gases curies 16.200 Iodine I-131 dose 0.02 I-101 equivslent curies To 31 Integr:ted Cose a Exclu:1cn Distance Thyrcid Kem 2.13 x 10~

%h01s body Rem 0.155

!JFDATE - 1

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. - . . - - . . _ - . ~ , . _ . . . _ . _ _

4 O Interrogatory 70 Describe and state the basis for Licensee's response to all findings, comments, and recommendations of "GPU's Third Party Review Group."

Response

See Reference Documents 63 and 64.

' Interrogatory 71 Define the following: core tubes; peripheral tubes; aver-age tubes.

Response

A " core tube" is a tube which is located at or naar the center of the OTSG tubesheet.

A " peripheral tube" is located near_or on the periph-ery of the tubesheet.

The " average tube" is located at the point where the tubesheet deflection is approximately equal to 47 percent of

~

the deflection at its center -- approximately 40 inches from the center of the tubesheet.

t II. INTERROGATORIES ADDRESSING THE THIRD PARTY REVIEW GROUP REPORTS Licensee was not a party to the final deliberations of the

. Third Party Review panel (TPR) for preparation of its reports.

The. panel has been disbanded. The answers to Interrogatories T-1 through T-51 are based on Licensee's knowledge of TPR activities and interviews with former members of the TPR. The

a .

representations made below thus reflect Licensee's understanding of factual material cited or conversations held for the purpose of responding to these interrogatories.

Interrogatory'T-1 State the educational and professional qualifications (including a complete list of publications) of each member of Licensee's " Third Party Review Group (TPR)."

Response

The educational and professional qualifications of the TPR members are listed in Attachment 2 of Appendix B of the

"" Report of' Third Party Review of Three Mile Island, Unit 1, Steam Generator Repair" dated February 18, 1983 (hereafter, TPR February 18 Report).

Interrogatory T-2 Describe the precise expertise which each member of the TPR has in the' area of fracture-mechanics.

. Response

! See response to Interrogatory T-1.

~ Interrogatory T-3 I

i Describe the precise expertise which each member of the

-TPR has in the area of stress analysis of steam generator tubes

at nuclear power plants. ,

l-

Response

See response to Interrogatory T-1.

Interrogatory T-4 Identify all those who attended or participated in TPR discussions or meetings, and state the educational and professional qualifications (including a complete list of pub-lications) of each such person. .

Peeponse The secretary of the TPR was the only other individual who attended or " participated" in the TPR discus-sions or meetings in a c'apacity associated with the TPR Charter. The role of the secretary is outlined in the TPR Charter which is Attachment 1 of Appendix B of the TPR Eebruary 18 Report. The secretary from inception through January 1983

-was Edward G. Wallace. A statement of professional and educa-tional qualifications is attached. Following January 1933, there was no person acting in the capacity of TPR secretary.

During all final deliberations by the TPR on formal findings and recommendations, no one other than TPR members was present.

1 TMIA INTERROGATORY NO. T-4 4.

Edward G. Wallace

)

Current Position: Manager, Oyster Creek Expanded Safety Systems Fa'cility Pro- i ject - Responsible for development and implementation of major plant retrofit project.

Previous GPUN Positions:

4 (09 06-83) Manager, Pressurized Water Reactor (PWR) Licensing Responsible for the management of all activities associated with obtaining and maintaining the necessary license; and permits required by federal, state and local regulatory agencies for FWR's within the GPU system. The major efforts involve active representation of company policy and

. programs to the Nuclear Regulatory Commission Staff, review of and concurrence in CPUN program activities perforned to assure the long term licensability of PWR projects, and the review and interpretation of regulatory requirements for

.other GPUN divisions. Served as Chairman, P;&W Owner's Group Steering Comedttee from August,1982 through September,1983.

(11 09-80) Licensing Manager - GPUSC Responsible for obtaining all permits and licenses for new fossil and nuclear GPU projects. Included special assign-ment to collect and analyze studies / reports related to the iMI-2 Accident, and participate in GPU internal studies.

(09 10-78) Lead Project Engineer, FRNS Project Engineering Manage-ment - GPUSC Responsible for the technical supervision and directio:i of the A/E on behalf of GPU. Activities included development and approval of schedules, contracts, design criteria, NSSS interface requirements, and construction interface requirements, and construction interface issues within the nuclear island. Supervised four GPUSC project engineers.

1 (05 09-77) Engineer, Mechanical System, GPUSC Performed system design criteria development activities for fossil and nuclear projects. Developed or reviewed all GPU technical responses to NRC associated with TMI-2 OL -

Licensing review. Special projects included study of decommissioning cost estimates for all GPU nuclear plants for rate making purposes; development of reliability improvement program proposal submitted to FEA.

. . - - ~ . -

Edw^ro G. Wallace Page 2 TIMIA INTERROGATORY NO. T-4 Previous Enployment: ,

(06 04-74) U. S. Navy - Nuclear trained officer, qualified as deck and ,

engineering watch officer on two nuclear submarines. At varying times supervised between 6 and 33 men in the main-tenance, repair and operation of 'all engineering /orepulsicn equipment associated with a nuclear submarine.

Education: B. S. - U. S. Naval Academy - 1969 Naval tt.lclear Pcwer School / Naval Prototype Training -'1970 Graouate Studies - Stevens Institute - 1974-1975 4

l I

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Interrogatory T-5 Identify-all those who acted as employees, agenta, advis-ors, or consultants to the TPR, and state the educational and professional qualifications (including a complete list of pub-

- 11 cations) of each member of Licensee's Third Party Review Gcoup (TPR).

Response

(

The members of the TPR were selected as individuals with expertise in areas relevant to the steam generator repair.

See response to Interrogatory T-1. It was understood that the views expressed by the members were their own and did not represent opinions or conclusions of their affiliated organiza-tions. When in group session, no additional individuals were present in a support capacity for TPR members. The extent to which individual members conferred with others when not in group session is not known.

- Interrogatory T-6 Identify all those responsible for or who contributed to defining the purpose, scope, membership, and operation of the TPR.

6 1

Response

The purpose and charter for the TPR were developed by R. F. Wilson, Vice President, Technical Functions and E. G.

Wallace. The draft charter was commented on by P. R. Clark, 4-then Executive Vice President, before release. The final document released is contained as Appendix B of the TPR k,

i=

. , , l~ ,

a ] .

, c' 4 g

(  %

February 18 Report. Once formed, the TPR operation was }

a directed by.the TPR members. At the suggestion of the TPR, Section IVA of the charter was modified to broaden that aspect of its review.

Interrogatory T-7 Describe procedurally how all interim and final findings, comments, and recommendations were arrived at by the TPR.

Response

"The evaluation was conducted using reviews of perti-nont documents, submittal of written questions to GPU Nuclear, written responses by GPU Nuclear, review of specialty topics by individual members, presentation by cognizant GPU Nuclear or contractor personnel, Review Group meetings, and Executive Ses-sions of the Review Group members only." (Par. 3, pg. 3 of In-terim Report of September 27, 1982). The area specialist would take the lend in each area and draft the initial version of findings and comments. The drafts of all report sections were circulated amongst TPR members with-comments conveyed to the originator or chairman depending on logistics. Interim and f

final findings were reached on the basis of these commente and discussions of the TPR members. Final report production was completed-by the chairman and reflects the collegial process

- used.

Interrogatory T-8 State the basis for the TPR's statement in its February 18, 1983 report, "[S]afe operation of the TMI-l plant after repair of the steam generators will be dependent on ...

(a) [cjompletion of analyses including ...the con-tingency of multiple tube rupture; (b) [t]ranslation of analytical work such as leak before break and multiple tube rupture into useable plant guid-ance, procedures, and training; and (c) a conservative approach of power escalation after completion of repairs."

Response

The TPR's comments in the conclusion section of the TPR February 18 Report were intended merely to flag to thc reader that the conclusions drawn were incomplete at that time since Licensee had not completed its analytical or planning ef-forts. The TPR February 18 Report conclusions were modified in the TPR May 16, 1983 Report (entitled " Report of the Third Party Review of Three Mile Island, Unit 1, Steam Generator Review - Supplement 1", hereafter, TPR May 16 Report) to reflect subsequent Licensee and TPR actions which closed these open issues.

Interrogatory T-9

. State the basis for the TPR's recommendation in its February 18, 1983 report, "(allthough sufficient operating ex-perience with other once-through steam generators (OTSGs) would justify allowing the OD indications less than 40 percent through-wall to remain in service, the ID indications are most probably stress corrosion cracks and should be plugged." De-scribe procedurally how this finding was arrived at.

Response

The occurrence of OD indications in other OTSG's has been the subject of review by other utilities, B&W, and NT.C.

The TPR was made aware of this experience by one of its members. OD cracking has been shown to take place over an ex- 1 tendad period of operation. Flow-induced vibration had been identified as a contributor to the specific failure events.

The experience has shown that the OD cracks are monitorable .

over a period of plant operation without gross rupture.

The ID cracking at TMI-l occurred over a very short period due to IGSAC. Because of the small number of tubes in-volved (60) and the uncertainties that remained at the time the recommendation was made, the TPR conservatively recommended plugging the tubes with less than 40 percent through-wall ID cracks.

Interrogatory T-10 Describe GPU's response to the recommendation referred to

'in'the preceding interrogatory, and identify all those respon-sible for formulating this response.

Resp o_ony Licensee's response is contained in Enclosure 1 Section A Response, of Licensee's letter to the TPR dated April 7, 1983. This letter and Enclosure 1 are included as Appendix A to the May 16 Report, and is Reference Document 63.

R. F. Wilson, Vice President, Technical Functions, GPUN was responsible for the formulation of that response, based'on advice from D. K. Cronoberger, Director Engineering and Design, GPUN, Sterling J. Weems, MPR; and an individual from B&W whose identity cannot be recalled.

Interrogatory T-11 State the basis for T?R's finding in its May 16, 1983 supplement that Licensee's response is satisfactory, and de-scribe procedurally how this finding was arrived at. Include a precise description of any dissenting views expressed by any members of the TPR, its advisors, or consultants, to this find-ing.

Response

Licensee's response noted that a new analysis had been done which showed that for the smaller number of tubes with ID defects to be left in service (17 versus the aarlier estimate of 60) only those tubes which had indications (1) below the 15th support plate and (2) less than 40% through-wall and 2 coils or less in circumferential extent, would be left in service. Additionally, TR-008,Section IX was revised (Rev. 2) and demonstrated that for tnose tubes meeting the above criteria, there was a significant margin between the size of the cracks and the critical crack size that would propagate to failure from fatigue.

The conclusion that Licensee's response was satisfac-tory was reached as discussed in the Background section of the TPR May 16 Report. There were no dissenting views.

o .

Interrogatory T-12 State the basis for the TPR's recommendation in its

> February 18, 1983 report, "[t]ubes within three rows of the e lane region and in the wedge-shaped region at the periphery which have OD indicatious at the 15th support plate or above, should be plugged as has been done in other OTSG's." Describe procedurally how this finding was arrived at.

Response

The recommendation was originally made by B&W to other B&W owners who had experienced OD cracking in the lane region. The TPR was made aware of this recommendation by one of its members and considered it appropriate for TMI-1.

Interrogatory T-13 Describe GPU's response to the recommendation referred to in the preceding interrogatory; and identify all those respon-sible for formulating this response.

Response

Licensee's response is contained in Reference Document 63.

R. F. Wilson, Vice President, Technical Functions, GPUN was respon31ble for the formulation of that response based on input from D. K. Croneberger, Director, Engineering and Design, GPUN; Sterling J. Weems, MPR; and B&W.

Interrogatory T-14 State the basis for TPR's finding in ita Mey 16, 1983 supplement that Licensee's response to the finding described in -

Interrogatory T-12 is satisfactory, and describe procedurally how this finding was arrived at. Include a precise description of any dissenting views expressed by any members of the TPR, its advisors, or consultants, to this finding.

c. .

R,esponse

h. Licensee's respo'nse exceeded the TPR's recommenda-l

< tion. See response to Interrogatory T-13. The conclusion that GPUN's'respense was satisfactory was reached as discussed in the L'ackground section of the TPR May 16 Report.

l-See response to-Interrogatory T-7 for.the procedural metho'd.used. There'were no dissenting views..

k Interrogatory T-15

-State the basis for TPR's finding in its May 16, 1983 supplement that-Licensee's response to Recommendations A.3 and A.4 is satisfactory, and describe procedurally'how this finding was arrived at. Include ^a precise description of any dissent-ing views expressed by any' members ~of the TPR, its advisors, or

' consultants',- to this finding.

Response

'k Licensee' agreed to perform the recommended inspec-tions. TR-008 (Rev. 2),'Section IX.A and Appendix A, and

[- Section III.B'provided the~ basis for the answer'to recommenda-

~

tion-A.3. The response to 'recommendationzA.4 was documented-in TR-008 (Rev. 2),Section II.E.2.

See response to Interrogatory T-7 for the procedural method u' sed. 'ihere were no dissenting views.

~

Interrogatory T-16 Describe all tests,. test data (as defined in 14B, sgpra),

and analyses. developed and performed by or relied upon by the

.TPRito determine the failure scenario described in the February

- 18,-1983 report (Finding B.1). Describe procedurally how this

-finding was-arrived at, and include a precise' description of= ,

any-dissenting views expressed by any members of the TPR, its-

-advisors,;or consultants, to this finding.

Response

The TPR performed no tests or analyses (in the sense of independent mathematical, engineering analyses). The TPR independently reviewed tests, data, and analyses developed and

[

submitted to it by Licensee or its consultants, and utilized l

[ individual member's expertise to verify the assumptions, i

( methods and conclusions contained therein. A full listing of Licensee material provided to the TPR as of February 18 in the bibliography (Appendix D) of the TPR February 18 Report. See also Reference Documents 63 and 66.

See response to Interrogatory T-7 for the procedural method used. There were no dissenting views.

Interrogatory T-17 State the basis for the TPR's recommendation in its February 18, 1983 report, "[wje recommend that GPU Nuclear im-plement corrective measures or verify their existing programs for minimizing ingress of all impurities (not just sulfur) into the reactor coolant system." Describe procedurally how this finding was arrived at.

Response' The TPR discussed the failure analysis scenario to examine the questions of whether other contaminant sources were possible, and whether the scenario accounted for the spectrum of contaminants found on the tubing. On both questions the TPR concluded that Licensee's scenario was the most likely. Howev-er, because of the variety of sources and types of potential r . . . . . . . . . . . . _ . _ ___ __ _

e Q corrodants that exist in a nuclear power plant, the TPR recommended that additional measures be taken to minimize the potential for any impurity introduction to the RCS.

See respones to Interrogatory T-7 for the procedural methods used.

Interrogatory T-18 Describe GPU's response to the recommendation referred to

-in the preceding interrogatory, and identify all those respou-sible for formulating this response.

Response

Licensee's response is contained in Reference Document 63. R. F. Wilson,, Vice President, Technical Functions, GPUN was responsible for that response based on input received from D. K. Croneberger, Director, Engineering and Design, GPUN.

Interrogatory T-19 c' State the basis for TPR's finding in its May 16, 1983 supplement (B.~ Recommendation 1), that "GPU Nuclear actions are considered adequate for safety." Describe procedurally how this finding was arrived at, including a precise description of any dissenting views expressed by any members of the TPR, its advisors, or consultants, to this finding.

Response

Licensee agreed to implement the recommendations of the TPR. See Reference Docuuont 63.

l 1

W The conclusion that this response was satisfactory was reached as discusced in the' Background section of the TPR May 16 Report. There were no dissenting views.

Interrogatory T-20 State the basis for the TPR's decision to recommend addi-tional actions in its May 16, 1983 supplement (Further comments B.1-6) and describe precisely why these actions are unnecessary to insure safe plant operation. Describe procedurally how these "Further comments" were arrived at.

Response

The further comments provided additional items for GPUN consideration and were intended to encourage GPUN to complete actions already underway in a timely manner, or to suggested limits which would make the actions planned more ef-fective. These comments were directed toward plant operability or reliability improvements, rather.than plant safety.

Interrogatory T-21 Describe all tests, test data (as defined in 14B, supra),

and analyses developed and performed by or relied upon by the TPR to determine'the future reliability of the materials in the system.

Response

The TPR performed no tests or analyses (in the sense of independent mathematical, engineering analyses) to determine the future reliability of the materials in the system. The TPR independently reviewed tests, test data, and analyses developed

and submitted by Licensee or its consultants, or utilized the

' expertise of it.divicual members, to verify the validity of as-sumptions, methods and conclusions used therein. A full list of material submitted by Licensee is providad in the bibliogra-phy (Appendix D) of the TPR February 18 Report, and Appendices A and B of the TPR May 16 Report. See also. Refer ~

ence Documents 63 and 66.

Interrogatory T-22 Describe the " minor differences" in findings between the two " independent metallurgical failure analyses performed," and describe their resolution, referred to in the February 18, 1983 report (Finding C.1).

Response

Licensee does not believe a listing of the specific

. minor differences exists. The basis for the differences was determined to be that the two independent failure analysis were performed using different equipment and techniques and were done on different tube samples. The TPR orally received an ex-planation of how these differences were resolved, and was i satisfied with the response. The TPR determined that the dif-ferences were not significant.

Interrogatory T-23 State the basis for the TPR's comment in its February 18,

'1983 report, "[ cracking in the rest of the reactor coolant system)... tend to be very tight and are indeed very difficult

~

to detect." Describe procedurally how this finding was arrived at.

. .o

Response

The TPR's comment reflected the opinion that associ-ated IGSAC in the RCS would be difficult to detect, given that the OTSG tube IGSAC cracks that had not propagated to failure were very tight and difficult to detect without high sensitivi-ty non-destructive evaluation (NDE) techniques. The same NDE techniques used in the OTSG's were not useable for the RCS in-spection.

See response to Interrogatory T-7 for the procedural methods used.

Interrogatory T-24 State the basis for the TPR's comment in its February 18, 1983 report, "GPU Nuclear should remain alert to the possibili-ty that small cracks may, in fact, be present in susceptible components of.the reactor coolant." Describe how the TPR pro-poses that Licensee " remain alert."

Response

The TPR's comment was based en the belief that to some degree, each of the necessary ingredients for IGSAC were present during the OTSG damage period at other locations in the RCS, and non-destructive detection is difficult. Therefore, continued attention to the susceptible components is warranted.

The TPR did not propose a method for Licensee to

" remain' alert."

Interrogatory T-25 State the basis for the TPR's comment in its February 18, 1983 report, "[t]he analysis which led to [the conclusion that steam generator tube defects below a certain size range will

- not' propagate due to flow-induced vibrations] depends on a -

large extrapolations of a limited crack-propagation-rate data base. This makes'it hard to substantiate a firm conclusion."

Describe procedurally how this finding was arrived at.

Response

During the December 7, 8 an'd 9 meetings, the TPR examined'the basis for Licensee's selection of the stress in-tensity threshold value below which fatigue induced propagation would not occur. Licensee noted that the data base it used was primarily MIT laboratory data. The large extrapolation can be seen in the attached curve."Inconnel 600 Threshold Stress In-

' tunsity (MIT Corrosion Laboratory Data)". This curve was ex-cerpted from the TPR bibliography Item 15.3. This was consid-ered to be a tentative basis for the conclusion drawn in TR-008,Section IX, Rev. 1.

See response to Interrogatory T-7_for the TPR proce-dural method used.

t

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .__ _ l

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i y l< c 'f, ,, , a.s v'r . r-TMIA INTERROGATORY NO. T-25 Inconel 800 Threshold Stress Intensity (MIT Corrosion 1.aboratory Data) -

10-3 e FREQUENCY = 5 Hz e R (PMIN/PMAX) = 0.05

• 554 F PURE WATER

• 77 F AIR 10 4 -

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• 808*F, YMI-1 PWR CHEMISTRY ll e Hz AND R SIMILAR TO OTSG

!l TUBE LOADING ll 10-8 ! !I '

r 2 3 5 10 100 THRESHOLD STRESS INTENSITY AKth (M Psh/S 10/18/s2

Interrogatory T-26.

Describe and identify the "large extrapolations" referred to in the' preceding interrogatory.

Response

See response to Interrogatory T-25.

Interrogatory T-27 State. precisely what data was "found to help substantiate GPU's analysis," referred to in the May 16 report.

Response

Licensee identified data in the engineering litera-ture (identified in TR-008 (Rev. 2), Figure IX-4 as Journal of ,

~ Engineering Materials and MIT curves) which covered a portion of the data:needs.

Interrogatory T-28 Describe and identify the extrapolations which are still necessary, as referred to in the May 16 report.

Response

The extrapolations referred to are made from the data used and presented in TR-008 (Rev. 2),Section IX.to the stress intensity' threshold value below which fatigue-induced crack propagation would not occur.

1 i

l Interrogatory T-29 Does the TPR claim that on the basis of this new " data" and extrapolations referred to in the preceding interrogatories, a firm conclusion that steam generator tubing defects below a certain size range will not propagate due to -

flow-induced vibrations, is still hard to substantiate?

(a) If the answer is "no", state the basis for i determining this, particularly in light of the limited data base available for crack propagation rates that has been used in the GPU analysis.

(b) If the answer is "yes", describe the precise safety significance of this finding, and describe GPU's re-sponse to this finding.

Response

No. The TPR auggested that, if practical, simulated vibration loading should be included in the long-term corrosion tests as a conservative response to a limited data base. The TPR concluded that based on the available data and the conser-vative analytical values used, the steam generators could proceed through the hot functional test and operational program proposed by Licensee.

Interrogatory T-30 State the basis for the TPR's comment in its February 18, 1983 report, "a flow-induced vibration type of loading ...

could make a significant non-conservative difference in the results once a crack is initiated." Describe procedurally how this finding was arrived at.

I

O

Response

The TPR comment was focussed on the long-term corro-sion tests Licensee was performing as " lead" tests. In that -

the " lead" tests did not duplicate the vibrational environment the tubes would see in normal operation and the fatigue analysis was in need of revision, the results could over predict tube life if vibration plays a significant role.

See response to Interrogatory T-7 for the procedural methods used.

Interrogatory T-31 Do you agree with GPU Nuclear's conclusion that

' " flow-induced vibrations may not play'any role in propagating

. steam generator cracks? State the basis for your answer.

' Response The quote does not accurately reflect Licensee's con-clusion. The TPR (and Licensee) felt that flow-induced vibra-tion-was an important mechanism of potentially propagating pre-existing cracks and thatiis needed to be understood to de-termine what role it played. Licensee's analysis of

-flow-induced crack propagation concluded that flow-induced vi-bration did not play a significant role in the propagation of small cracks. The TPR reviewed the analysis. Discussions of this issue are found in the TPR February 18 Report, Sections C and.E,' and TPR May.16 Report, Sections C and E.

~

7s Interrogatory'T-32

. State the basis for. recommending long-term corrosion tests which include a simulated flow-induced vibrations loading.

Response

These tests were discussed in a TPR comment, but were uvt included as a TPR recommendation. See also response to In-terrogatory-T-29.

Interrogatory T-33 Describe all long-term corrosion tests which include a simulated' flow-induced. vibrations loading which will be or have been performed by or relied upon by the Licensee. -

Response

Licensee is not aware that any such tests have been or will be.run.

Interrogatory T-34 State the basis for the TPR comment C.4 in the May 16 report, "there is much about reactions between peroxides and system materials which is not understood, so that (in spite of testing) there remains a risk that the process could be detri-mental."

Response

The TPR comment reflected a conservative view that

~

some residual ~ risk to RCS materials would remain because of the complexity of peroxide reactions, and the spectrum of materials and local conditions that exist in the RCS. The TPR did not

(+n . .

feel that the residual risk waa large. Thus the TPR concluded that the peroxide flush was not expected to have an adverse effect on plant safety.

Interrogatory T-35 Does the TPR believe there is any uncertainty as to the residual risks or effects of the clean up process undertaken by Licensee. If.the answer is "yes"-describe the residual risks or effects which remain, or could remain, and describe their safety significance.

Response

See response to Interrogatory T-34.

Interrogatory T-36 Does the TPR claim that the stress levels on the TMI-1 steam generator tubes can-be higher, or the strength of the

' tubes lower, than those in a normal OTSG7 Describe all tests (including a description of which tubes were tested in the TMI-1 steam generater, and their location within the tube bun-die), test data (as defined in 14B, suprn), and analyses developed and performed by or relied upon oy the TPR to deter-

-mine this. Include a definition of a " normal OTSG."

(a) If the answer is "no", state the basis for determining thisi and explain how this reconciles with Finding E.1 in the February 1983 TPR report.

~

(b) If the answer is "yes", describe the precise safety significance of this finding.

-Response l1 The answer is no. This is consistent with Comment Eal, which-expressed only a " minor qualification", as set out in Comment 4 in Section E-16 of the TPR February 18 Report.

~93-i - - -

t The TPR did not define the concept of a " normal" steam generator.

Interrogatory T-37 -

Does the TPR claim that the integrity of the tubes in the TMI-1 steam generator have been reduced to a degree where restart or the subsequent operation of the plant may be influ-enced? Describe all tests (including a description of which tubes were tested in the TMI-1 steam generator, and their loca-tion within the tube bundle), test data (as defined in 14B, supra), and analyses developed and performed by or relied upon by the TPR to determine this.

(a) If the answer is "no", state the basis for determining this, and explain how this reconciles with finding E.1 in the February 1983 TPR report.

(b) If the answer is yes", doncribe how the integri-ty of the tubes in the TMI-1 steam generator have been reduced to influence restart or the subsequent operation of the plant.

Re_sponse The answer is no, as expressly stated in Finding 1, Section E of the TPR February 18 Report. See also, Bibliogra-phy (Appendix D) of the TPR February 18 Report.

i Interrogatory T-38 Does the TPR claim that "the cooldowns starting from the one in April 1979," have subjected the tubes to stresses that are higher than the design stresses?

(a) If "no", state the basis for this statement in light of the TPR comment, "from the informati on received, the levels of these stresses could not be determined with any accu-racy."

(b) If "yes", describe these stress levels, and the effects of these stresses on tne tubes.

Response

No. The TPR believes that the stresses were deter-mined with sufficient accuracy to conclude that the tubes have -

not been subjected to stresses higher than design stresses.

See TPR Report Comment E-1, page 15.

Interrogatory T-39 State the basis for the TPR's commert in its February 18, 1983 raport, "[t]he explosive expansion of the tubes could affect the stress levels, if the process would change the strength or some dimensions of the tubes."

Response

The TPR was simply stating one of the issues that it was going to examine as part of the overall review of the stress analysis of the steam generator.

Interrogatory T-40 Does the TPR claim that the repair process changed the strength and dimensions of the tubes? If the answer is "yes" describe precisely how the strength and dimensions of the tubeu were changed.

Response

The TPR had reviewed Licensee's assessment of both the' strength and dimensional changes that may have occurred as a result of the repair process. The strength was not affected.

The radini dimensions were changed in the kinetic expansion region'but not in a fashion that significantly increased the 4 '

stresses in the tube. The axial dimensions were basically.

unchanged.- The resultant changes were "not expected to affect significantly the streso levels in the tubes in the restart and subsequent operation periods." (TPR February 18 Report Pg. 15).

Interrogatory T-41 Does the TPR claim that the changed strength and diman-siona of the tubes due to the repair process can affect the stress levels in the tubes in the restart and subsequent operation periods?

(a) If'"yes", describe these effects, and state the basis for your answer.

(b) If "no", state the basis for this answer. ,

Response

No. See Response to Interrogatory T-40.

Interrogatory T-42 Dees the TPR conclude that "a corresive environment, and not abnormal stress levels, must have been responsible for the appearance of the cracks," solely on the basis that signif3-cantly higher stress _ levels than in normal OTSG's were not found?

(a) If the answer is "no", what other basis has the TPR relied upon'to reach'its conclusion.

(b) If "yes", state the basis for this statement.

Response

The TPR did not conclude that abnormal stess levels enused the cracks. See TPR February 18 Report Sections B and

-E.

Interrogatory T-43 State-the basis for the TPR's comment in its February 18, 1983 report, "the tubas probably have some small defects that were not detected by the eddy current tests and were not elimi-nated by the repair." Describe procedurally how this finding -

-was arrived at.

Response

The TPR concurred in Licensee's position on the pos-sibility of small undetected defects. See TR-008, Section IX-C.

See responso to Interrogatory T-7 for the procedural method used.

Interrogatory T-44 State the basis for the TPR's comment in its February 18, 1983 report, "[t]hese defects present the potential for leaving the tubes in a weaker condition than in a normal OTSG." De-scribe procedurally how this finding was arrived at.

Response

The TPR comment reflected the need to examine the stability and/or propagation Iate of the small cracks that

-might remain in an in-service tube. Should the cracks propa-gate ac an unacceptable rate and go undetected, the ability of a tube to withstand a design basis load (i.e., main steam line break) could be compromised. See the TPR February 18 Report, g

Section E,. Comments 4 and 5.

?:

See response to Interrogatory T-7 for the procedural method used.

f-l- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ )

_ Interrogatory T-45 Does the TPR claim that the small defects referred to in the two preceding interrogatories could propagate as fatigue cracks if the tubes were subjected to flow induced vibration?

State the basis for your answer, and explain how the limited -

data base available for crack propagation rates that has been used in the GPU analysis affects the conclusion.

Response

No. The TPR comments reflect their interest in the

. free span of the tubes. In that area of the tubes, the TPR examined and did not find fault with extensive analysis by Li-censee. See TR-008 (Rev. 2), Section IX-D, which concluded that flow-induced vibration will not propagate, in an instable manner, small undatected cracks. See also response to Interrogatory T-29.

Interrogatory T-46 State the basis for the TPR's comment in its February 18, 1983 report, 'among the undetected defects there may be some that are large enough to break through to the OD and propagate along-the circumference in a stable manner, with the potential of breaking the tube when the crack becomes unstable." De-scribe procedurally how this finding was arrived at.

Response

The TPR noted that there is a possibility that there are cracks in this category which simply escaped detection due to ECT process random errors.

See response to Interrogatory T-7 for the procedural method used.

b e .

Interrogatory T-47

) Does the TPR claim that the tubes referred to in the pre-ceding interrogatory may break before leak? State the basis for your answer.

i.

Response

No. See the TPR February 18 Report, Section G, Comment 5.

Interrogatory T-48 State the a ' asis for determining that relevant stresses in the free span part of the tubes are axisymmetric throughout the wall?

Response

-Gee TR-008, Section IX-C.

L Interrogatory T-49 Does the TPR claim that stresses in the expancion transi-tion zone are axisymmetric? State the basis for your answer.

Response

Yes. See the TPR February 18 Report, Section E, Comment 5.

'Interrogatgyy T-50 Does the TPR claim that in the transition zcne the tube could break before it leaks? State the basis for your answer.

-Dsscribe.the precise cafety significance of a " controlled leak"

'which would result, including its consequences regarding radio-active releases, and its impact on the stresses within the steam generators during normal and accident conditions.

_. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ . _ _ _ _ _ . - . _ _ _ _ ___ __ _ 1

e ,

Response

No. Supplemental stress analyses of the transition zone wero performed by Licensee. These analysis show an ac-

- ceptable stress level in the transition zone. See Reference Document 29, Sections 2.3-2.4; 66.

Interrogatory T-51 Gtate the basis for the TPR's comment in its February 18, 1983 report, that thare is a " low probability that (a maximum

, total leak rate of 1 lb/hr.] will be obtained in these steam generators." Describe procedurally how this finding was ar-rived at.

Response

The kinetic expansion repair method is similar to a common industrial fabrication process used on a variety of heat exchanger types. Based on the TPR member's experience with the other heat exchanger types, there tends to be some slow change in leak tightness of the joint as opcrating time accumulates.

See response to Interrogatory T-7 for the procedural methods used.

III. LICENSEE'S RESPONSE TO TMIA's REQUEST FOR PRODUCTION OF DOCUMENTS _

TMIA's Request for Production of Documents is so broadly framed as to seek virtually every document which has been gen-erated with respect to the steam generator repair program,

-including material prepared by individuals independent of

t. ... -. . .m .. . . . . . .

o., ,

a Licensee.(the TPR), and information beyond the scope of this proceeding. Thousands of documents are involved. TMIA's re-quest thus is overly broad, unduly burdensome, and seeks infor-mation not relevant to the instant w.tter.

Rather than object at this time and embark on the long &nd arduous task of attempting to argue the specific documents in dispute and the overall burdensomeness of the request, and without waiving its objections, Licensee will make available to TMIA essentially all of its files relating to ti;e tube repair project for TMIA's inspection. Because of their volume, the documents will be placed in a Discovery Room in Middletown, Pennsylvania, in the vicinity of the Visitor's Center for Three

. Mile Island. The reading room will be staffed with personnel who can direct TMIA to the appropriate document or information, and will contain tables and chairs for the perusal and study of documents, as well as microfilm readers and other necessary equipment. The doc'.tments can be examined at the Reading Room, or Licensee will make' copies for TMIA at a nominal charge of 10 cents per page.

Documents containing proprietary information will be made avail able pursuant to applicable proprietary agreement and pro-tective order.

Document production is due on January 29, 1984. 10 C.F.R. 6 2.741(d). However, Licensee will attempt to have the

'6 - -- - -

i . .. ,

documents available for inspection by the rniddle of next week, on or about January 18.

Respectfully submitted, SHAW, PITTMAN, POTTS & TROWBRIDGE f

c_ I

- .q /

Ge'orge F. Trowbridge,'P.C.

Bruce W. Churchill, P.C.

Diane E. Burkley Wilbert Washington, II Counsel for Licensee 1800 M Street, N.W.

Washington, D.C. 20036 (202) 822-1000 Dated: January 13, 1984 9

1 L

2,..

LIST OF REFERENCE DOCUMENTS (For Replies to Interrogatories)

O RIGINATO R' S

' RE F . NO . NATURE OF DOCUMENT UPGD DOC. NO. ORIGINATOR DOC. NO.

• 1. Test Report 58-1017513-00 FW 5054-IR-7 Rev. 0

• 2. Test Report 58-1017505-00 FW 5054-IR-4 Rev. 0

• 3. Test Report 58-1017499-01 FW 50 54-IR-1 Rev. L

• 4. Test Report 58-1017506-00

• FW 5054-IR-2 Rev. 0

• 5. Test Report 58-1017507-00 FW 5054-IR-3 Rev. 0

• 6. Test Report 58-1017510-00 FW 5054-IR-5 Rev. 0

• 7. Test Report 58-1017509-00 FW 5054-IR-6 Rev. 0

• 8. Test Report 58-1017511-00 FW 5054-IR-8 Rev. 0

• 9. Test Report 58-1017518-00 FW 5054-IR-ll Rev. 0

• 10. Test Report 58-1017517-00 FW 50 54-IR-12 Rev. 0

• 'll. Test Report 58-1017512-00 FW 50 54-IR-10 Rev. 0

• 12. Test Report 58-1017508-00 FW 5051-IR-9 Rev. 0

• 13. Procedure 03-1024011-01 B&W-ARC ARC-TP-53 4 Rev. 1

• 14. Procedure 03-1023412-01 B&W-ARC ARC-TP-543 Rev. 1

• 15. Procedure-Cond. 03-1023419-01 B&W-ARC ARC-TP-560 Rev. 1

• 16. Procedure 03-1024013-01 B&W-ARC ARC-TP-526 Rev. 0

• 17. Repair Spec. 44-1020890-01 GPU SP-1101-2 2-006, Rev.

• 18. Final Report 77-1141709-01 B&W-UPGD TDR-007

• 19. Struct. Just. 51-1141584-00 B&W-UPGD NA

• 20. Statist. Calc. 32-1139261-01 B&W-UPGD NA

• 21. Test Report 58-1017555-00 FW 5090-IR-1 Rev. l'

• ' 2 2. Test Report 58-1017556-00 FW 5090-IR-2 Fev. O

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ i

ORT.GINATOR ' S R2F. NO. NATURE OF DOCUMENT UPGD DOC. NC. ORIGINATOR DOC. NO.

• 23. Test Report 77-1135317-00 B&W-LRC RDD.83:5390-03-01'

• 24. Topical Report BAW 10146 B&W-UPGD NA

• 25. Equip. Spec. 08-1134285-02 B&W-UPGD NA

26. Functional Spec. CS ( F) 9 2 / B&W-UPGD NA NSS-5/0770

• 27. Test Report 58-1017488-00 FW 9-69-5049/7/19/82

28. Test Report --

GPUN TDR-417

29. Test Report GPUN TDR-388

30. Test Report GPUN TDR-488

31. Test Repo rt GPUU TDR-421

• 32. Test Report B&W RDD:83:5243-04:01

33. Test Report GPUN TDR-341

• 34. GPUN Specification No. SP-1101-12-039, " Acceptance Criteria for OTSG Repair Tools and Materials. "

• 35. Letter: Stanek to Slear, GPUN-82-359, 10/22/83, Quality Assurance Data Package fo'r First Shipment Explosive Charge Assembly.

• 36. NPGD Quality Assurance Certificate of Confor'.ance, 23-1138065-21, Contract No. 382-7239.

• 37. Manufacturing Specification 27-1137838-01 for Preparing ( l

' -2 3 6 LL, 10/27/82.

• 38. R&D Technical Report, " Chemical Cleaning of Nuclear Parts and Components [ ]," Murphy to Birks, 8/29/75.

• 39. ARC Report RDD:83:5170-01:01, Monter to Caray, "TMI-l [ }

Corrosion Tests," May 10, 1983.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ n

b A

1.

• 40. ARC Report RDD:83:5156-31-01:01, Rovder to Carey, " Assessment of Sampler Used in the Investigation of Various Cleaning, Methods for s Removal of Explosive Expansion Contamination from O.T.S .G Tube Surfaces," March 9, 1983.

• ' 41. B&W Technical Document, " Operating Specification 64-1140812-00 for GPUN Automatic Tube Flushing System.

• 42. Memo: Lamanna and Phillips to Distribution, "TMI-l OTSG Flush,"

PAS-09 March 18, 1983.

• 43. B&W, Test Results, LR:81:5267-05-01, October 26, 1981.

44. GPUN TDR-346, "TMI-l OTSG As-Built Stres s . Analysis. ," App. B.

i

45. J. V. Monter/G.J."Theus, B&W Alliance Research Center, TMI-l OTSG Corrosion Test Program - Final Report," RDD:83:5433-01-01:01, May 9, 1983.

• 46. J.C. Gress, J.H. DeVan, Oak Ridge National Laboratories,

" Behavior of ( ] in sulfer Contaminated Boric Acid I Solutions ," ORNL/TM8544.

f

47. GPUN Specification SP-1101-22-008, Rev. 2, Long Term Corrosion Testing, October 24, 1983.

• . 48. B&W -letter GPUN 83-005, January 6, 1983.

• 49. B&W letter GPUN 83-110, March 10, 1983.

50. GPUN Interoffice Memorandum MT1-ll33, Long Term Corrosion Testing, 7/22/82.

51. GPUN Interoffice Memorandum EM-82-284, Nominal Applied Load for Corrosion Testing, August 6, 1982.

4 6

52. .GPUN Specification SP-1101-28-001, Rev. 2, "Three Mile Island-Unit 1 Primary Water Chemistry," July 19, 1982. ,

53. B&W Interim Report " Examination of TMI-l Third Pulling Sequences OTSG Tubes," B&W Doc RDD:83 :5068-03 :02.

• 54. GPUN IOM MTI-1288 from JAJ to OGS, "IGSAC in [ ] Weld Metal and Existing Plug," April 21, 1983.

55. "EFRI Activities in Support of TMI-l Steam Generator Recovery" Memorandum Report, April 1982, Appendix III.

• 56. M. Rigdon, B&W Lynchburg Research Centar " Evaluation of Tube Samples from TMI-1," Document No. 77-1135317.

57. G.O. Hayner, B&W Lynchburg Research Center, "TM1-1 Second Static Corrosion Test," RDD:83:5087-03:01.

'58. '" Long Term Corrosion Test Program of Nuclear Steam Generator Tubing Samples from TMI Unit 1," First Interim Report, Westinghouse Electric Corporation, October 1983.

59. GPUN TOR 83-506, Final Report on RCS, Hydrogen Peroxide Cleaning.

~ 60. Barnes, Wensky and Lathouse, " Overview of Progross in the TMI-l Tube Cleaning Experiment," Battelle Columbus Laboratories, January 31, 1983.

• 61. GPUN Specification SP-1101-08-001, " Chemical Requirements for

[ ] Flush," Rev. 6, April 11, 1983.

• 62. A.K. _ Agrawal, W.N. Stiegelmeyer and W.E. Berry, "Short Term

/ Cerrosion Evaluation of Kinetically Expanded Tubes in Mockup

' Tube Sheets," Battelle Columbus Laboratories, November 21, 1933.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . -_ n

j.

-t 3 .

)

63 GPUN, Letter: TPR Recommendations and Comments, April 7, 1983.

64. GPUN, IOM, Responses to TPR Recommendations, October 7, 1983.

65. Battelle rinal Report on Examination of Components from the Pressurized Operated Relief Valve and the Waste Gas System on Three Mile Island - Unit 1, July 28, 1983.

66. GPUN letter - EJW-83-0048. TPR Report Supplement 2, December 3, 1983.

UNITED. STATES OF AMERICA ,

NUCLEAR REGULATORY COMMISSION Before The Atomic Safety And Licensing Board ,

In the Matter of ) 4

)

METROPOLITAN EDISON COMPANY ) Docket No. 50-289-OLA

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

Station, Unit No. 1) )

f, I have prepared or assisted in the preparation of the answers to Interrogatories 26,34,35,36,37,38,39, and 40. Said answers are true.and correct to the best of my knowledge and belief. '

~1 L s %m

. Branch Elam Subscribed'and sworn to me this

- / 7 '/r. day of January, 1984.

/ s 1/ $t ,a,. (n 00 umf

.5'tary d Public My Commission Expires: / ,-: c / / ,- ,

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[ Seal]

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UNITED STATES OF AMERICA ,

NUCLEAR REGULATORY COMMISSION Before The Atomic Safety And Licensing Board In the Matter of )

)

METROPOLI'!AN EDISON COMPANY ) Docket No. 50-289-OLA

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

Station, Unit No. 1) )

I have prepared or assisted in the preparation of the answers to Interrogatories 3,7,11,12,.18,19,20,21,22,35,47-57, and 59-68. Said answers are true and correct to the best of my kncewledge and belief.

4 X' s ct <? c c ~ V F. Scott Glac6bbe.

Subscribed and sworn to me this 2 I4 -day of January, 1984.

I h n>, n. /

-t%.n .t Notary Public My Commission Expires: [ofers/"S, / '/I6

/

J

[ Seal]

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.co UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before The Atomic Safety And Licensing Board In tne Matter of )

)

METROPOLITAN EDISON COMPANY ) Docket No. 50-289-OLA

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

. Station, Unit No. 1) )

I have prepared or assisted in the preparation of the answers to Interrogatories'10,13,41,56,69 and 70. Said answers are true and correct to the best of my knowledge and belief.

1 I k L,j k (.4, kil (J %v > d MaryJpng' Graham f

Subscribed and sworn to me this G Nt , day of January, 1984.

N/ a <k ts /

Notary Public My Commission Expires: 4 /d[ 8I, //f'I

[ Seal]

s i

UNITED STATES OF AMERICA '

NUCLEAR REGULATORY COMMISSION Before The Atomic Safety And Licensing Board In the Matter of )

)

METROPOLITAN EDISON COMPANY ) Docket No. 50-289-OLA

) ASLBP 83-491-04-OLA (Three Mile Island Nuclear ) (Steam Generator R>po.ir)

Station, Unit No. 1) )

I'have ptepared or assisted in the preparation of the answers to Interrogatorias 2, 4, 5, 6A, 6B, 8, 25, 42, 43 and 71. Saic, answers are true and correct to the best of my kncwledge and belief.

/

teoiNJ e L.p!Eiann h

f Subscribed and sworn to me this

/;-- day of. January, 1984.

I

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A

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-Notary Public A- .~: L;. a My Commission Expires: f- -- - <

l fl C - : ! D.1?t7

[ Seal]

_m _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .

a i UNITED STATES OF AMERICA NUCLEAP. REGULATORY COMMISSION Before Thc Atomic Safety And Licensing Board In the Matter of )

I METROPOLITAN EDISON COMPANY ) Docket No. 50-289-OLA

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

Station, Unit No. 1) )

I have prepared or assisted in the preparation of the answers to Interrogatories 1, 2, 4, 6B, 8, 14, 15, 16, 17, 43, 44-45, 46 and 53. Said answers are true and correct to the best of my knowledge and belief.

l- 1, 1

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)

nwa ) J'ism George g sm6s ~

/

Subseg bed and sworn to me this

/) - day of January, 1984.

'f;ib.!lt I) n m pn b v'J

-Notary Public d My Commission Expires: ut,( /' / ' '

u

[Seall

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~.

i UNITED STATES OF AMERICA -

NUCLEAR REGULATORY COMMISSION Before The Atomic Safety And Licensing Board In the Matter of )

)

METROPOLITAN EDISON C ~)MPANY ) Docket No. 50-289-OLA

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

- ) (Steam Generator Repair)

Station, Unit No. 1) )

I have prepared or assisted in the preparation of the answers to Interrogatories T-1 through T-51. Said answers are true and correct to the best of my knowledge and belief.

f((qff ( J & ci w w _

Edward G. Wallace Subscribed and sworn to me this

//fi f January, 1984.

n>>>$ A

[otaryPublic My Commission Expires: 30, /N 7

[ Seal]

i O

4 8 l

UNITED STATES OF AMERICA -

NUCLEAR REGULATORY COMMISSION Before The Atomic Safety and Licensing Board In the Matter of )

)

METROPOLITAN EDISON COMPANY ) Docket No. 50-289-OLA

) ASLBP 82-491-04-OLA (Three Mile Island Nuclear ) (Steara Generator Repair)

Station, Unit No. 1) )

I have prepared or assisted in the preparation of the answers to Interrogatories 5, 6A, 9 and 16. Said answers are true and correct to the best of my knowledge and belief.

1 fA j. '/ ken Sterling J. Weems subscribed and sworn to me this

/AC4, day of January, 1984.

YO4'? XXMAA$~

Notary Public My Commission Expires: Jd N7 f

[ Seal}

_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . - _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ . _ _ _ _ _ )

./-

Y UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Beforc the Atomic Safety and Licensing Board In the Matter of )

)

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

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

Station,-Unit No. 1) )

CERTIFICATE OF SERVICE This is to certify that copies of " Licensee's Answer to TMIA's First Set of Interroge. tories and Request for Production of Documents" are being served to all those on the attached Service List by deposit in the United States mail, first class, postage-prepaid, thia 13th day of January, 1984.

BTuce W Churchill, P.C.

Dated: January.13, 1984

_-.______m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ . _ _ _ . . _ _

P 1

l UNITED STATES OF AMERICA

) NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

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

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

Station, Unit No. 1) )

SERVICE LIST Sheldon J. Wolfe Atomic Safety and Licensing Administrative Judge Board Panel Chairman, Atomic Safety and U.S. Nuclear Regulatory Commission Licensing Board Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Docketing and Service Section (3)

Washington, D.C. 20555 office of the Secretary U.S. Nuclear Regulatory Commission Dr. David L. Hetrick Washington, D.C. 20555 Administrative Judge

~

Atomic Safety and Licensing Board Joanne Doroshow, Esq.

Professor of Nuclear Engineering Louise Bradford University of Arizona Three Mile Island Alert, Inc.

Tucson, Arizona 85271 315 Peffer Street Harrisburg, Pennsylvania 17102 Dr. James C. Lamb, III Administrative Judge Jane Lee Atomic Safety and Licensing Board 183 Valley Road 313 Woodhaven Road Etters, Pennsylvania 17319 Chapel Hill, North Carolina 27514 Norman Aamodt Richard J. Rawson, Esq. R. D. 5, Box 428 i Mary E. Wagner, Esq. Coatesville, Pennsylvania 19320 Office of Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Comission Washington, D.C. 20555

. - . _ _ _ . _ . _ _ _ . _ _