ML19346A211
| ML19346A211 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/03/1981 |
| From: | Chackes K CHACKES & HOARE, COALITION FOR THE ENVIRONMENT, ST.LOUIS REGION |
| To: | NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| References | |
| ISSUANCES-OL, NUDOCS 8106050408 | |
| Download: ML19346A211 (39) | |
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UNITED STATES OF AMERCA 1 NUCLEAR REGULATORY COMM'.SSION '),
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD $
M' wy9 In the Matter of )
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UNION ELECTRIC COMPANY ) Docket No. STN 50-48 3-OL *' /
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JOINT INTERVENORS' FIRST SET yy 0 I .
C OF INTERROGATORIES TO NCR STAFF 9 v.s. g s N ,,,s
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Joint Intervenors Coalition for the Environmen Louis '
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Region, Missourians for Safe Energy and Crawdad Alliance rdquesti tnat the atrached interrogatories be answered fully, in writing, and under oath by members of the NRC Staff who have personal know-l ledge therecf or are the closest to having personal knowledge thereof.
Unioss otherwise stated all of the attached interrogatories relate to Union Electric Company's Callawa'r Plant Unit One. Nuclear Regulatory Commission is sometimes abbreviated "NRC."
" Identify" or " Identification", when used with reference to a document or documents requires, in addition to whatever information is specifically requested, identification of the NRC document number (if applicable), a statement of the date of the document, the general nature and description of the subject matter and contents of the d'o'cunienti, 'and the name (s) of the person (s) who prepared the document.
" Identify" or " Identification", when used with respect to a person or persons requires, in addition to whatever information is specifically requested, a statement of the full name, current or I last known address and telephone number, employment and position as 1
of the time period (s) relevant to the subject interrogatory, and
//
8106050
1 current employment and position.
' Respectfully submitted, CHACKES AND HOARE ,
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Kenneth M. Chackes Attorneys for Joint Intervenors 314 North Broadway St. Louis, MO 63102 314/241-7961 i
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1 THE FOLLOWING INTERROGATORIES RELATE TO CONTENTION ONE, I.
SUBSTANDARD REINFORCED CONCRETE CONSTRUCTION, A. EM3EDDED PLATES.
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- 1. Were one or more Stop Work Orders issued on June 9,1977 suspending the use of safety related embedments in concrete pours pending an inspection of all the safety related embedments onsite?
If the answer is in the affirmative, please state:
(a) The exact terms of each Stop Work Order; (b) The reasons why each Stop ' Work Order was issued.
- 2. During the process of evaluating af ter June 9, 1977 whether the embedded plates presented a safety-significant problem, did the NRC Staff determine or approve Union Electric's determination that some exceptions to structural welding code standards would be permissible? If so, please state fully:
(a) What exception (s) would be permis::ible; j
(b) For each exception in (a), the reason (s) for the determination that the exception was permissible; (c) The identity and location of any documents in your possession relating to the determination that exceptions were permissible.
- 3. State whether an NRC inspector conducted an audit of the Cives Steel Company with regard to the fabrication of the embedments for the Callaway Plant. If the answer is affirmativa, identify the
- document (s ) relating to such an audit.
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THC FOL* OWING INTERROGATORIES RELATE TO CONTENTION ONE, I.
SUBSTANDARD REINFORCED CONCRETE, 3. CRACKS IN CONCRETE.
. 4. Regarding Nuclear Regulatory Commission (NRC) Report i .
No. 50-4C3/78-01, what was the width of ths crack mentioned on p.
j 20, entry 13a, in the " plant" north wall of the control Building?
- 5. Regarding. Nuclear Regulatory Commission (NRC) Ra'po rt No .
50-483/78-01, state the nunber of the "other cracks" referred to on page 20 of this report.
- 6. S tate the location, length, width and shape of each crack counted in the answer to the preceding interrogatory.
- 7. State the reason or reasons why the cracks referred to on page 20 of NRC Report No. 50-480/78-01 are described- as a " recurring
! problem," that is, why are such cracks recurrent?
- 8. What, in the opinion of the' NRC St.aff, is the cause of (a) the twelve foot long crack described on page 20, entry 13a of NRC Report No. 50-483/78-01;
! (b) the "other cracks" described on page 20, entry 13a l
l of NRC Report No. 50-483/78-01. Each crack should be separately addressed if there are different causes foz different cracks.
- 9. What reason or reasons does the NRC Staff have to believe
! that concrete such as that described on page 20, entry 13a of NRC Report No. 50-483/78-01 will not develop additional' cracks?
- 10. What reason (s) does the NRC Staff have to believe that the cracks described o'n page 20, entry 13a of the NRC Report No.
50-483/78-01 will not increase in size. .
- 11. Have any employees or agents of the NRC measured any of i
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the cracks described on page 20, entry 13a of NRC Report No. 50-483/78-01 subsequent to the filing of Report No. 50-483/78-017 If so, state the following:
(a) identify all documents in the possession or control of the NRC Staf f in which there is reference to the subsequently measured cracks; (b) identify personnel who made the subsequent measurements; (c) summarize the NRC Staff's assessment of the subsequent measurements .
- 12. Are the cracks described on page 20, entry 13a of NRC Report No. 50-483/78-01 still accessible to visual or instrument inspection? ,
- 13. Regarding Nuclear Regulatory Commission (NRC) Report No.
50-483/78-03, what is meant on page 3 by the statement that NCR 2-2081-C- A was " superceded" by NCR 2-217 3-C-A7 .
- 14. State NRC Staff's assessment of the safety signifi-icance of the twelve foot crack in the Control Building wall, referred to on page 3 of NRC Report No. 50-483/78-03 and on page i
! 20 of the NRC Report No. 50-483/78-01.
- 15. Regarding Nuclear Regulatory Commission (NRC) Report No.
I 50-483/77-06, and the circumferential concrete crack in the Reactor Containment Building referred to on pp. 20-21,. state the 6,llowing:
(a) the length of the crack; (b) the depth of the crack; (c) the width of the crack; (d) the proximity of the crack to reinfor,cing steel and other embedded materials.
- 16. Identify any documents in the control or possession of
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the NRC which make reference. to the circumferential crack e in the r sactor cavity moat area, approximately 42 inches from cavity liner and extending through 270 degrees are, referred to on pp. 20-21 of NRC Report No. 50-483/77-06.
- 17. Su:.imarize separately the content of each document listed
- in answer to the preceding interrogatory.
- 18. This interrogatory applies to the following sentence in NRC Report No.- 5 0-4 8 3/77-06, pp. 20-21: "It was reported by licensee by telephone on May 10, 1977 . . that an investigation- had been initiated to determine the safety related significance of the crack
! (in the reactor cavity moat area]" (emphasis added) :
(a) State the names and titles of all persons involved ,
in the inver- tigation here referred to; l
(b) Summarize the result of the investigation here referred to.
- 19. ,This interrogatory applies to- the following sentence in
. l NRC Report No. 50-483/77-06, p. 21: " The inspector visited the subject area on June 28, 1977, and observed that, in accordance with the coaclusion of the investigation and af ter repairs _ required -by <
the NRC had been completed and accepted by QC personnel, work had f
j progressed to the extent that physical inspection of the repair was not possible" (Emphases added) :
, _ (a) Identify by name and title the inspector referred to; l (b) Descr'oe in detail the " repairs required by the NRC";
i (c) Identify all documents which relate to " accept (ance] ,
of the repair] by QC personnel",
i- (d) State whether Union Electric's continuation of work had progressed to such an extent that physical i'nspection was not possible violated any NRC D
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t requirements, any quality control procedures or any work procedures, and if the answer is affirmative, s tate the requirements and procedures violated.
4 (e) State the reason (s) why quality assurance and safety ,
assurance would not be adversely affected by the Commission's lack of opportunity to physically ' inspect the repair.
- 20. Describe in detail the repairs made to the circumferential crack in the reactor cavity moat area, as further described in NRC Report No. 50-483/77-06, pp. 20-21.
- 21. Identify all documents which relate to acceptance of the repair of the circumferential crack in the reactor cavity moat area by quality control personnel.
- 22. State whether, in the NRC Staff's opinion, the repair of the circumferential concrete crack in the reactor cavity moat area, documented in NRC Report No. 5 0-4 8 3/77-06, pp. 20-21, meets the quality assurance criteria of 10 C.F.R. Part 50 Appendix 3.
S tate the bas s for the answer to this interrogatory.
- 23. State the cause of the circ'umferential crack in the reactor cavity moat area.
- 24. Identify all documents which establish a reporting
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standard for cracks in permanent concrete' at the Callaway Plant.
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- 25. Summarize the reporting standards set forth in the I
documents listed in the answer to the preceding interrogatory.
- 26. How many recurrences of cracks, as anticipated by NRC l
Report No. 50-483/78-01, page 20, have occurred since the date of
- that report? ,
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- 27. Provide nonconformance report (UCR) numbers for,any cracks counted in the answer to the preceding interrogatory.
(a) Describe any procedures , modifications of ' materials ,
modifications in design, implemented subsequent to NRC Report No. 50-483/78-01, which are intended to detect, lessen or eliminate the problem of concrete cracks at the Callaway Plant.
(b) State the dates on which items listed in the answer to paragraph (a) of this interrogatory were , effective.
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THE FC* *.OWING INTERROGATOR 7ES P.2LA~E TO CONTENTION ONE , I.
SU3 STANDARD REINFORCED CONCRETE CONSTRUCTION, C. HONEYCOMB ING , 1.
REACTOR BASE MAT.
- 28. Regarding NRC , Report No. 50-483/78-02, p. 4, and the statement that technical specification C-103 (Q) was revised to differentiate major from minor defects which require approval prior to repair:
(a) summarize the differentiation between major and minor defects, as established by the revision of C-103(Q);
(b) state whether major and minor concrete defects were undif ferentiated for purposes of repair approval prior to the January 23, 1978 revision of C-103 (Q) . i
- 29. S tate whether, in the NRC Staf f's opinion, the honeycombing in the base mat resulted, i'h whole or in part, from the congestion of trumplate wall dowels , main steel, rebar supports , or form ties, or any or all of these singly or in combination, so that adequate vibration of the concrete mat was hampered. )
- 30. Are there any areas of the base ' mat which are less marked than other areas by congestion as described in the preceding interrog- :
l atory, and, if the answer is affirmative, describe the differences in conges tion by specific area.
- 31. State the NRC Staf f's conclusion as to the cause of honeycombing in the tendon access gallery. concrete, attributing relative weight and probability to each cause if more than ene cause is named. Also state:
(a) Nhat actions were takcr. by Union Electric subsequent to NCR 2-0856-C-A to prevent the recurrence of voids
in concrete such as those 'in the tendon access gallery, as described in NCR 2-0856-C-A;
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(b) Whether the NRC Staff believes the preventive actions described'in the answer to paragraph (a) 'of this-l-
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l interrogatory are adequate; (c) The basis for the NRC Staff',s conclusion as to the i
adequacy of preventive a.ctions- named in the , answer i
to paragraph (a) of this interrogatory.
- 32. Have NRC requirements relating to the placement.of concrete in Category I concrete structures been changed or modified
' during construction at Callaway Unit 1, and, if the answer is affirmative, state: .
(a)~ the nature of the change or modification; f
(b) the reason for the change or modification; (c) the effective date of the change or modification; i
l l (d) identify the document (s)' effecting the change or modification.
l l 33. The following interrogatory applies to the : following.
sentence in NRC Report No. 5 0-4 8 31 7 7-07, p. 4: "WP-109. . . does not identify either ' vibration' , ' consolidation' , or ' densification, '
clarificatbn of the procedural requirements is required."
- (a) State the implications for safety and quality assurance in the lack of clarification of procedural requirements I as herein referred to; (b) State whether the clarification herein referred to has been provided; ,
(c) Summarize separately the nature of each clarification 9
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which has been provided for vibration, consolidation, and/or densification, respectively; (d) ' Identify the documents containing the clarification ,
referr'ed to 'in the answer to the preceding paragraph of this interrogatory, .and the ef fective date of each clarification.
- 34. Did repairs of the honecombing in the tendon access' gallery conform to the repair procedure suggested by Daniel Inter-i national, as described in NCR 2-0856-C-A, and if the answer is negative:
(a) State. the nature of deviations from the suggested procedure; -
(b) State the reason (s) for deviations from the suggested procedure; (c) . Identify all documents relating 'to repairs and summarize i
l their contents .
- 35. Regarding NRC Report No. 50-483/77-06, p. 22, state f
whether recair of the honeycombing therein referred to was hampered f by limited mobility of work crews due to the trumplate wall-dowels, the main steel, rebar supports and form ties.
- 36. If the answer to the preceding interrogatory is in the Megative, describe the repair -procedures utilized and the reason (s) .
shy work crew and equipment mobility was not an impediment.in these repairs.
- 37. Regarding the statement in NRC Report No. 50-463/77-07,
- p. 13, that " dry-pack grout was not being tested as required. . .
. Specification C-191. failed to include such a . tes t,"
be caus e . . . .
l . s tate whether the untested' dry-pack grout herein referred to, or t !
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the repairs involving this grout, have been tes ted subsequent to NRC Report No. 50-433/77-07.
- 38. If the answer to the preceding interrogatory is -in the affirmative, identify all documents which refer to test of the dry-pack grout,. and summarize separately the contents of each such document.
If the answer is in the negative, state the reason why no tests have been performed.
- 39. Does the NRC S taff believe that af ter repairs of voids in the tendon access gallery, the condition of the base -slab has no adverse safety implications?
- 40. State the bases for the NRC Staff's conclusion as to the preceding interrogatory.
- 41. Regarding the report on the soniscopic study of the base slab by Wiss, Janey, Es tner and Associates , Inc., dated August 1, 1977, identify the NRC inspector, referred to on page 13 of NRC Report #50-483/77-07, who inspected this report.
- 42. S tate whether the NRC has evidence of honeycombing at nuclear power plants other than Callaway where Bechtel Power Corp.
was architect / engineer, contractor or subcontractor. If affirmative ,
identify the other plants and provide the following information,.
separately for each plant:
(a) S ta te the da te (s ) when concrete was poured in which honeycombing was found; (b) Identify all NRC reports which pertain to the honeycombing.
- 43. State whe ther the NRC has evidence of honeycombing at nuclear power plants other than Callawny wher'e Daniel International If af firmative , identify the was contractor or subcontractor.
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other plants and provide the following information, separately for each plant:
(a) State the,date(s) when the concrete was poured -in which honeycombing was found.
(b) Identify all NRC Reports which pertain .to the ,
honeycombing.
1
THE FOLLOWING INTERROGATORIES RELATE TO CONTENTION ONE, I.
SUBSTANDARD REINFORCED CONCRETE ONSTRUCTION, C. HONEYCOMBING,
- 2. REACTOR BUILDING DOME.
- 44. State separately for each void or imperfection which has occurred in the Reactor Building dome, including but not necessarily limited to the seven areas of imperfection referred to in NRC Report '
No. 50-483/80-30, pp. 3-4, the depth,' width, length, location and shape of each such imperfection.
- 45. State whether, in the opinion of the. NRC Staff, the im-perfections and voids in the concrete of the dome, described in NRC Report 50-483/80-30, pp. 3-4, are attributable to the same cause (s) as the voids in the base mat.
- 46. State the bases for the conclusion in 'the preceding interrogatory.
- 47. State the NRC Staff's conclusion as to the cause(s) of ,
honeycombing in the Reactor Building dome described in NRC Report No. 50-483/80-30, pp. 3-4, attibuting relative weight and probability to each cause if more than one cause is named.
- 48. What is the basis in NCR 2SN-2790-C for specifying the imperfections in the concrete of the dome unreportable?
- 49. Identify the design specifications governing the thickness of the exterior walls of Callaway Unit 1 Reactor Building dome.
- 50. How thick are the exterior walls of the dome according to the design specifications listed in the answer to the preceding interrogatory.
- 51. In the opinion of the NRC Staff, how extensive can honey-i combing be before the integrity of the containment building dome of Callaway Uni't 1 is compromised?
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- 52. This interrogatory pertains to the following sentence in l NRC Report No. 50-483/80-30, p. 4: "(L} icensee personnel attributed l i
the occurrence of the imperfections to the complex nature of those portions of the dome slab where the imperfections had occurred".
(emphasis added) . State separately for each portion of the dome where an imperfection occurred the nature of the " complexity" here referred to. .
- 53. Why have actions taken by Union Electric subsequent to e the discovery of voids in the tendon access gallery, and designed to prevent recurrence of voids in concrete, proved inadequate to prevent voids in the dome?
- 54. S tate any and all reasons the NRC Staf f has to believe that imperfections in the concrete of the dome are limited to areas identified in NRC Report No. 50-483/80-30, pp. 3-4, and identify all documents and tests which form a basis for this conclusion.
- 55. By what testing methods d'id the NRC Staff determine I
the extent of imperfections in the concrete of the dome? )
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- 56. State whether ice was used in lieu of water in the Reactor Building dome concrete mix. If so, provide th'e following !
additional information:
(a) State whether the use of ice in lieu of water in concret f mix violates. any procedures , regulations, or requiremen-applicable to construction at Callaway Unit I?
(b) Identify r il procedures, regulations, or requirements which form a basis for the answer to paragraph (a) l 1
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I of this interrogatory, and summarize separately the content of each item listed.
- 57. This interrogatory applies to the following sentence in NRC Report No. 50-483/80-30, p. 5: "The licensee has committed to and undertaken actions to address the reactor dome concrete imper-l fection issue." (emphasis added) l (a) State separately the nature of each action referred l
l to; i (b) State the date of each action undertaken or anticipated; ,
'. c ) State the extent to which each action listed in the answer to paragraph (a) of this interrogatory has resolved the dome concrete imperfection issue.
- 53. Regarding NRC Report No. 50-483/80-27, p. 21, state the car.se of " flaking" on a concrete repair therein referred to.
59, S tate whether the matter of flaking concrete, referred to in the preceding interrogatory, has been closed by subsequent-inspection, anu if the answer is affirmative, identify and state the number of the closing Nuclear Regulatory Commission Report.
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THE FOLLOWING INTERROGATORIES RELATE TO CONTENTION ONE, I.
SUBSTANDARD REINFORCED CONCRETE CONSTRUCTION, D. CONCRETE COVER.
- 60. Identify by number, date and relevant pages all i
documents which pertain, in whole or in part, to nonconformance with l
! concrete cover requirements through the' fif th lif t of the Reactor ;
Building and as to each report listed, further state:
i (a) the nature of the nonconformance; f
i (b) the location of the . nonconformance; (c) the date on which conformance was achieved or explain other disposition of the nonconformance; I
(d) identify documents which verify conformance.
- 61. State the date on which Union Electri'c first communicated with the NRC Staff regarding the NRC interpretation of concrete
! cover requirements with regard to the Reactor Building wall at -
Callaway Unit 1, and further degcribe in detail the communication.
- 62. This interrogatory pertains to the :lollowing sentence in NRC Report 50-483/77-11, p. 4: " [At 340 degrees azimuth of the third lif t of the Reactor Building] concrete cover was less than that required by NRC interpretation of the concrete cover requirements, but within the concrete cover requirements as interpreted by the licensee and contractors." (emphasis added) :
(a) Describe the relationship and location of the reinforcing steel and concrete cover in dispute;-
(b) Identify all documents settir.g forth the " concrete cover requirements" referred to in the above sentence, and summarize separately the contents of each
- document; (c) Identify all documents setting forth the NRC's j ..
. .. . .. interpretation of the concrete cover requirements I
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referred to in the above sencence, and summarize separately the contents of each document; (d) Identify all documents ' setting forth the licensee's
.and contractors' interpretation of the above-mentioned concrete cover requirements, and summarize separately the contents of each document.
- 63. Regarding the conflict between the NRC's interpretation of concrete cover requirements at 340 degrees azimuth and the licensee / contractors' iaterpretation, as indicated -in NRC Report' -
No. 50-483/77-11, p. 4, state which interpretation' prevailed at 340 degrees azimuth.
- 64. This interrogatory applies to the following sentence in s NRC Report 50-483/77-11, p. 10: "Bechtel Power " Corporation personnel repeated that their interpretation of the cover requirements was that the bwo-inch cover requirement can be reducec to an absolute -
minimum of an inch and one third per a provision of the specifications which allows reduction of the specified cover by one-third" . (emphasis added):
\ (a) Identify the provision which allows reduction of the specified cover by one-third, and summarize its contents ;
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(b) State whether the NRC Staff believes the one-third reduction can be utilized without adverse safety implicatio ns , and the basis for the Staff's belief.
- 65. Provide identifying information regarding "a draf t Code case" involving the matter of concrete cover, as referred to on page 10 of NRC Report No. 50-483/77-11; and state whether the Code L
change under consideration was implemented.
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- 66. This interrogacory applies to the following sentence in NRC Report No. 50-483/77-11, p. 11: "[A] two-inch minimum concrete cover will be required for the sixth and subsequent lif ts, utilizing the ' fif th lif t as a transition area." Explain what is meant by utilizing the fif th lif t as a tranrition area.
- 67. NRC Report No. 50-483/77-11, at p. 11, indicates that
' Union Electric is evaluating" the requirement that the company comply with concrete cover requirements at the sixth and subsequent lif ts:
(a) State the outcome of Union Electric's evaluation; (b) Identify documents which pertain to Union Electric's evaluation, and summarize separately the contents of each document. .
- 68. Identify, and provide report numbers and dates, for all nonconformance reports pertaining, in whole or in part, to concrete placement in the third lif t area, including but not necessarily limited to the 23 reports mentioned in NRC Report No. 50-483/77-10, at p. 19, and summarize separately the contents of each report.
- 69. Insofar as NRC Report No. 50-483/77-10 mentions 23 noncon-formance reports on the third lif t concrete pour and describes this as "an unusually large number," explain any and all factors which could be considered to have contributed to such a large number of l repo.rts . _
- 70. Identify the ten nonconformance reports pertaining to the third lif t of the Reactor Building wall which were still outstanding the evening of November 21, 1977.
- 71. This interrogatory applies to information in item 3. a. (8),
page 8-9 of NRC Report No. 50-483/78-01, where "several areas" of se
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concrete covar are mentioned as . "less than two inches as specified on pl:2 cement drawings."
(a) Describe the' location and size of each of the "several areas" and describe the location of rebar 'within each area; (b) Summarize the discussion at the January 23, 1978 meeting between Union Electric and the NRC in Bethesda, Maryland, as to the areas of nonconfozming concrete cover described in item 3.a. (8) ;
(c) S tate the decision reached at- the January 23,:1978 meeting as to whether the two inch cover required by Bechtel Topical Report BC-TOP ,5, Section CC-3533.1 of Appendix C could be reduced by one-third pu. suant-to specification No. C-112.
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- 71. This interrogatory applies to information in item 3.a. (7),
j page 8, of NRC Report No. 50-483/78-01, involving "several areas" of " concrete cover. .
, of 12 to 13 inches which appeared to be more )
than' permitt9d. "
(a) Describe the location and size of the "several areas" herein referred to; (b) Summarize the discussion in the January 23, 1978 meeting between Union Electric and the NRC in Bethesda, Maryland, as to the areas of nonconformance described 1 in item 3.a.(7);
(c) State the. decision which was reached in the January 23, 1978 meeting as to the areas in which concrete cover exceeded maximum allowed by BC-TOP-5, Section
.bC-3534 of Appendix C. 1
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- 73. Regarding cariaticns baneath the six:n lift of :22 Isac:or Building from the two-inch concrete cover requirement established by.
Section CC-3 5 33.1 of Appendix C to 3C-TOP-3 for = 5 -through =13 reinforcing steel, state the following:
(a) The number of variations- from Section CC-3533.1 of Appendix C to BC-TOP-5 for #6 through #18 reinforcing steel, beneath the sixt.%-lift;
- 03) The location of each such variation; (c) The safety implications of such variations when considered cumulatively.
- 74. This interrogatory applies' to a 1 1/2 r nch placement tolerance described in .!RC Report No. 50-483/78~01, item 3.b.(14),
- p. 11, and to :;CR 2-205 5-C-A which was "dispositioned 'use as is' by Bechtel prior to concrete placeme.7 t":
(a) S tate the basis for Bechtel's "use as is" disposition in NCR 2-2055-C-A; .
(b) State the NRC Staf f's conclusion as to the safety implications of the "use as is" disposition of'NCR 2-2055-C-A, and the basis for this conclusion.
- 75. S tate whether Union :lectric, Bechtel Power Corporation, Danial Inter. ational Corporation, or SNUPPS , alone or in concert, have cbjected to the requirement at the Callaway Plant of a minimum concrete cover of two inches over reinforcing steel on the outer face of the reactor containment with no placement tolerance on that minimum dimension at or above the sixth lift, and if. the answer is af firmative,* further state:
(a) The bases for objection to the requirement; l - , _
.02)_.ihe Details of any alternative requirement proposed
(c) -The reasons se: forth by the. proposing party for adoption of any alternative requirement named in answer to the preceding paragraph of this interrogatory.
- 76. S tate whether Union Electric, Bechtel Power Corporation, Daniel International Corporation, or SNUPPS, alone or in concert, have objected to the requirement at the Callaway Plant .of a maximum concrete cover on face reinforcing ~ steel 'of ten inches- at or above the sixth lif t, and if the answer is affirmative, further state:
(a) The bases for objection to the requirement; '
(b) The details of any alternative requirement proposed by any of the above parties;
- (c) The reasons set forth by the proposing party for 7
adoption of any alternative requirement named in answer i .
to the preceding paragraph of this interrogatory.
l l
- 77. State to what extent the NRC Staf f believes that a 1
I reduction in the two-inch minimum concrete cover on reinforcing steel in the lower five lif ts of the Callaway Reactor Building exterior wall may have reduced or compromised the following properties of the structural sys tem:
(a) Protection against corrosion of the steel if exposed l
to weather;
~
~ ~ (b)
~
Protection against excessive heat; (c) Assurance of adequate bond for rebar development.
- 78. S tate to what extent the NRC Staff believes that exceeding the ten-inch maximum concrete cover on reinforcing steel below the
- sixth .lif t of the Reactor Building may have reduced or compromised the structural , system's ability to control cracking.
' l. .
r l 79. State whether Union Electric =ade any changes in its minimum and maximum concrete cover requirements or procedures for i
reinforcing steel in outside faces of buildings other than the i
Reactor Building following the NRC meeting of January 23, 1978. If 1
l the answer is affirmative, cite specific changes and identify relevant documents.
- 80. For the 23 NCRs regarding the third lif t referred to in NRC Report No. 50-483/77-10, page 19, state on what da ,' and what hour each was resolved or closed out.
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l l TF.E FOLLOWING INTERROGATORIES RELATE TO CONTENTION ONE, II. SUBSTANDARD '
l-
! PIPING, A. MATERIAL MANUFACTURING
- 81. With respect to NRC Report No. 483/80-10 and the allegations , inves tigations , and inspections upon which it is based, provide the following information: . I 1
(a) When, by whom and by what medium were ; the subjact )
allegations communicated to the NRC?
(b) Identify by. name and- position all employees of the NRC having firsthand knowledge of ti.e allegations.
(c) With respect to the telephone conversations between.
the alleger and the NRC Region III office on or' about October 11, 1979 and on or about April 11, 1980, state.
(separately for each conversation) whether the NRC has the following items: I 1
(i) a tape or other electronic recording;
-(ii) a transcript of the conversation (s);
(iii) notes taken by NRC employee (s) of the contents of the conversation (s) .
If the answer to any part of this interrogatory ((c) (i) , (ii) or i I
(iii)) is in the negative, state for such part whether the NRC ever t
had such an item and explain why the NRC no longer has it.
If the answer to any part of this interrogatory ((c) (i), (ii)
~
or (iii) ) is affirmative, provide the contents of the subject item..
(d) Near the bottom of page 4 of NRC Report No. 50-483/80-lc the report states 'that a total of 15 welds were encompassed. Exhibit A, . however, refers to four spool pieces, i.e. , S 0 01, S 0 02, S 0 07, and j
- - , - . . - - - , , . , . , . . , _ , , , , , , , , ~ , , , , , , . , ,
,- ,,,- , .. ,- ,_.~, .., ...,. .--,.. ,.- . ,,--, ,. ,. n., , , , . . - - - - , - , , - . , - . , , - .,
l' Intorrogatory 81, continu3d
' S 008, and twenty-six welds (6 field welds, 11 circumferential factory welds, and 9 longitudina.
factory welds) . Provide the following information:
(i) State whether only 15 of ~ 26 ' welds were inspected and if so,, explain why.
(ii) Identify exactly which welds were inspected and which welds were not inspected in the .
inspections made on March 26 and 27,1980.
(iii) Describe in detail, what, if any, effort was made to contact the alleger concerning the location of the alleged crack.
(e) On Exhibit A to NRC Report No. 5 0~-4 8 3/ 8 0-10, is a hand-written note with a line drawn to S 002 which says: " Seam inspected by WLK". Provide the following.
information with respect to that reference:
(i) State the name of "WLK" and his or her employment position on the date of the inspection, (ii) When was the subject inspection made.
(iii) Describe how the subject inspection was made, stating specifically whether the inspector looked into the end of S 0 02 and examined the
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inside of the seam weld.
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(f) Exhibit B, attached to NRC Report No. 50-483/80-10, contains a nonconformance report (NCR) and a deficiency report (DR). The NCR refers to an " overlap" while
' the DR states that " the vendor's longitudinal weld. . .
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'* s a. o =, e s
+ - - - - . - ,- ..,y .,*ws. .-w
- -- -3 y y -=y- -. ,,,-g-e, - <,s-.,,,w w. -e-p.---c- ,,.-, e= w - -= w%, y.,
'IntOrroga tory 81, continuud is not fused uniformly into the plate surface as required by Material Spec. SA 3 58, Para' 5. 2. 3" .
(i) Are the NCR and the DR referring .to 'the same defect in the quoted portion set out above?
(ii) Would " overlap" be a violation of Material Spec. SA 358, Para. 5.2.3?
(g) Exhibit B, attached to NCR Report No. 50-483/80-10, at pages 7 and 8, indicates' that NCP No. 2S N- 05 01-P was downgraded to a DR based on a letter from Mr. J.L. -
Turdera, Bechtel's Project Engineering Manager. In paragraph (a) of his. letter Mr. Turdera deals with the maximum allowable reinforcement and cites ASME Section III but ignores ASME .Section II, . listed as a
" Controlling Document" in the NCR. .
(i) Did Mr. Turdera properly ignore Section II i
in his discussion of maximum reinforcebent when it was listed in the " Controlling Documents' If affirmative, please explain why it was i
proper.
l (ii) Why didn' t tP NRC Resident Inspector question the fact that Mr. Turdera ignored Section II when.he (the inspector) received these documents on April 14, 1980?
(iii) In the investigation that occurred on March 26 and 27, 1980, why did the investigators not have these " Controlling Documents"? .
(h) In Exhibit B, page 8, paragraph (a), Mr. Turdera cites ASME Section III para. NC 44 26.2. Ar ticle
~ - . . . .
[
ntsrrogatory 81, cont'd.
I NC 4000 from which this is taken is titidd Fabrication ,
I and Installation and is not a material specification. -
Section 'III para. NC 2561, Required Examination, states:
" pipe made in accordance with.. . . SA 358. . . shall be treated as material" . This .would make Section II SA 358 the basic controlling document governing longi-tudinal seam welds in SA 358' pipe. Both the NCR (Exhibit B, page 7) and the DR (Exhibit B , page 1) agree in this and cite Section II SA 358. Was Mr.
Turdera correct in. citing NC 4426.27 If affirmative, please give code references that substantiate this.
(i) Was the pipe in 2-EP-01-S 0 02 single or double welded?
Identify the documents which verify this.
(j) In paragraph (b) , of Exhibit B, page 8, the letter e
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s tstes : "The SA 358 material specification -references the ASME Code,Section III, paragraph UW-51(b)". .The letter also cites: " paragraph UW-51(b) of Section VIII of the ASME Code" . Are these references correct? If l I
I not, please provide correct references.
(k) Exhibit B, page 8, paragraph (b) deals with " overlap", .
a defect detected by visual examination. . Was Mr.
~ ~
Turdera correct in failing to deal with " overlap" as j a violation of Section II SA 358 para. 5.2.3? If he was correct explain why.
(1) Why wasn't the overlap in spool piece 2-EP-01-S 002 removed prior to f abricating the spool piece?
=e
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4 Interrogatory 81, cont' d.
(m) How were the radiographs of the longitudinal w' elds in spool No. 2-2P-01-S '0 02 correctly read and accepted
- with weld overlap?
(n) In paragraph (c) of Exhibit B, page-8,.several persons-listed d:here, " agreed that the pipe' meets -the code requirements and does not fall under NCR category".
l (i) Was that determination correct?
l (ii) Was that determination based on the -information given in paragraphs (a) and (b) ? -
(iii) ~If not, upon what was that determination based? -
(iv) Did any of those individuals read the subject -
NCR and consider the requirements of Section l II EA 3587 l (v) Was this letter ,a basis for the NRC report saying that, "no items of noncompliance were ' identified?
(vi) Are there cases where ASME code . violations in ,
Class II piping do not fall under the .4CR category? If yes, please explain.
(o) In NRC Report No. 50-483/80-10, the fif th paragraph l
states: "The remaining reinforcement (" fall through")
in the pipe was measured and found to be within ASME f
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welding code tolerances."
I (i) Wnat examination method was used in making this measurement?
(ii) At what intervals and over what length of the f pipe were these measurements made?
,iii)
( Was the reinforcement measured back to. the
. 45 degree elbow?
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. . . . , - , - . . . , . . . . , , . - , - . . . - - , , . . , - - , . ~ . , . . , - . . , , , . - ~ . _ , . , . . . . . - . . . ~ , . . - .
f.. .
Interrogatory 81, cont' d.
(iv) Identify and provide the contents of the inspection report, if any, used to document i
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the measurements.
(v) What section and paragraph of the ASME code was used to establish.. the allowable amount of reinforcement?
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(p) Exhibit 3, page 3 is an NCR describing a minimum wall violation. The cause of the nonconformance is given as, " ovality in pipe" .
(i) After the ovality was noticed was the pipe checked for conformance to ASME Section II SA 358 para. 15.1.2?
(ii) Was this measurement documented? If so, identify the subject document.
(iii) At what intervals was the pipe measured?
(iv) Did the pipe conform to SA 358 para. 15.1.27 (q) In Exhibit B, page 5 is an Inspection Report regarding the wall thickness of the edge preparation. I t lis ts
)
the ASME code as the controlling code and as the inspection standard. The NCR, Exhibit B , page 3, does not list the ASME code as a controlling document.
~ ~ ' '
(i) Should the NCR, Exhibit B , page 3, list the A3ME code as a controlling document? If not, explain why.
l (ii) Do 'Bechtel specifications take precedence over l
ASME specifications?
(r) ,In Exhibit B , on p. age 3 under the heading, " Recommended
,,,,,, Disposition and Basis for Recommendation" is the
~
Interrogatory 81, continued s tatement , "Bechtel to determine that min. wall of
. 814 will meet design criterias" ; and on page 4, Exhibit B, the statement is made that, " the calculated minimum wall thickness for 10" BCB. . . is .795".
(i) What design criteria were used to determine this minimum wall?
(ii) What paragraphs of ASME Section III were used in this determination? .
(iii) What paragraphs of ASME Section III were used in recalulating the minimum wall?
(s) Who was the vendor of the spool piece which was the subject of NRC Report No. 50-483/80-107
( t) Who manuf actured and supplied the' pipe to the vendor?
(u) When was the pipe manufactured?
(v) Were the radiographs of the longitudinal welds in spool No. 2-3P-01-S 0 02 checked by the NRC? If they were, what were their findings?
l (w) When was each of the following first notified, 1
formally or informally, of the subject allegations?
(i) Daniel International; ,
_ , . (ii) Union. Electric; (iii) Bechtel; (iv) SNUPPS.
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l THE FOLLOWING INTERROGATORIES RELATE TO CONTENTION TWO.
- 82. State with respect to each of the following publications the answers to the following:
Publications: .
PWR GALE Code, NUREG 0 017 Regulatory Guide 1.21 Regulatory Guide 1.112 (a) When was the document first published?
(b) When was ay revision published?
I (c) When was the document completed in final form (before being published) ;
! (d) When was any revision completed in final form (before
! being published) ;
I (e) Lis t each nuclear f acility from which data were i
cc,llected in the preparation of each document and/or revision and state with respect to each facility the i
f number of Effective Full Power Days.
(f) State whether any revision of the document is presently being worked on and if so, the identity of each nuclear facility from which data have been or will be collected and how long each such facility has been in operation.
- 83. State the identity of each nuclear facility which has been' licensed for operation since the date a document or its revision has been completed as. contemplated in the preceding interrogatory.
- 84. If hafnium is used in the control rods instead of silver, i
indium and cadmium, what changes in the fission, activation, and 9
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,-.e,,y ,, ,r.---,w., , , w- m-< , . ~ - y
corrosion products might be expected?
- 85. At what Westinghouse reactors has there been experience
~
with hafnium control rods?
- 86. S tate the derivation of the maximum permissible -concen-trations (MPC) listed in 10 CFR 20 Appen' dix B Table II and describe the relationship of Appendix B to 10 CFR 20 Sections 105 and 106. -
- 87. Idntify e document in which the relationship described in answer ' to the preceding interrogatory is explained.
- 88. Would a person who drinks two liters a day of water, each liter contaminated with the MPC listed in 10 CFR 20 Appendix B, Table II for one i,sotope, receive a radiation dose equivalent of 500 millirems per year? .
- 89. State whether an NRC-licensed facility's liquid or gaseous ef fluent calculations and monitor alarm set points will be based upon 10 CFR 20 Appendix B MPC's or upon 40 CFR 190.- .
- 90. What will be the amount (curies- per year) of tritium released each year in the liquid effluent. How is this figure derived?
(a) How much of this figure will be producsd by fission?
(b) How much by activation?
(c) of the amount produced by fission, describe fully the liquid pathway whereby the tritium will be released into liquid effluent.
- 91. State the derivation of the estimate of 410 curies of tritium as the average release per year in the liquid ef fluent of a 1000-megawatt pressurized water reactor using zirconium-alloy-clad fuel rods. Include an account of the amount of boric acid estimated to be used per ' year in the reactor vessel. ;
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_ .. _ _ _ . . . . _ _ _ - . . _ _ _ . - . _ -__ . . ~ _ - _ _ _ , _ . . - . _ _ _ _ _. _ _ . -
I.
- 92. Has any estimate or study been prepared which concludes that the operation of the Callaway Plant or a plant like We Callevay Plant could produce in edible fish in the Missouri River (or any river) an amount of radionuclides that could be dangerous to the health of a person eating the fish? If so, identify the study or estimate by title, author and date and state who now has such report in his possession or custody.
- 93. Are there any noble gases at all such as xenon-127, xenon-13 krypton-81 or krypton-85 released from a pressurized water reactor such as Callaway Plant Unit One? If so, state which gases and the l
l amount releaced in terms of curies per year in (a) the liquid effluent and (b) gaseous emissions. ,
- 94. What will be the amount (curies per year) of tritium released each year in gaseous emissions? How is tMs figure derived?
(a) How much of this figure will be produced by fission?
l (b) How much by activation?
l
, (c) of the amount produced by fission, describe fully the l
pathway whereby the tritium is released to the
! environment.
- 95. State the dispersion characteristics of solid particulates
! which may be released from a pressurized water reactor such as
_Callaway _ Unit one during a short term accident.
- 96. Describe the model assumed in answering the previous interrogatory.
- 97. Regarding SNUPPS FSAR, Section 12. 2. 2, state the NRC Staff's opinion of the correctness of Union Electric's using data l on in-plant radioactive contamination obtained at the Fort Calhoun nuclear plant to estimate in-plant contamination at Callaway. State l .
, , _ . . _ _ . , . . . . , , , . _ , . . . , _ . . , , ~ , . - . . . . . . , _ . _ , , , . , . _ , ~ . . .
l ti.e bases for the answer provided.
- 98. Regarding SNUPPS, Callaway site Addendum: p. 2.3-66, state the NRC Staff's opinion of the correctness of the choice of 2.26 days and 8 days as the half-lives used in the PUFF model calculations.
1 State the bases for the answer provided.
- 99. Will there be any tritium in the spent fuel pool? If
- o, how much will be added (curies per year) in each year of the normal expected or planned operation of the Callaway Plant? Will such additions increase as the plant gets older? If so, estimate l the amount or rate of increase.
100. Re ferring to the parameters and assumptions listed-i under Table 12. 2-11 (SNUPPS, FSAR, Vol. 9) , state whether the NRC S taff believes longer fuel storage and compacting of fuel assemblies in the spent fuel pool will change the estimated rate of evaporation and airborne radioactive concentrations.
101. References to temporary storage of spent fuel are made in Sectio ns 9.1.4.2.1 (p. 9.1-2 8), 9 .1. 4. 2. 3 - (p . 9.1-3 8) , 9.1.4.2.3.1.
Phase V (p . 9.1-4 6 ) and 12.2.1.7 (p . 12. 2-5) (SNUPPS , FSAR Vo ls . 6 and 9). State what changes in design parameters will be required if storage time is increased beyond the estimated temporary storage time in the spent fuel pools.
102. Referring to Table 9.1-4 (SNUPPS FSAR Vol. 6) " Fuel Pool Cooling and Cleanup System Design Parameters" S tate how evaporation rates can be controlled and whether any changes in the evaporation rate are to be expected due to longer storage time of spent fuel.
i 103. State whether the NRC requires any monitoring of the am'ount of tritium being formed when neutrons from the spent fuel react with boron in the spent fuel pool.
l
e 104. State the expected uranium, transuranic, and fission product leaching rates from exposed fuel pellets resulting from increased fuel rod cladding f ailure caused by long-term storage in spent fuel pools.
105. State the time periods that' the NRC Staff estimates spent fuel rods can be stored in spent fuel pools before degradation' of cladding and structural parts and increased compacting provide an increased risk of a criticality incident. -
106. State what percentage of fual cladding is expected to f ail af ter one year's storage in the spent fuel pool; af ter five years' storage; a
~
thirty years ' storage.
107. State how frequently the NRC requirds water of the spent fuel pool to be ronitored and for what specific radioactive isotopes it is to be monitorad.
108. State which radioactive isotopes are expected to be reler--d to the liquid waste effluent from the spent fuel pool.
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THE FOLLOWING ARE GENERAL INTERROGATORIIS RELATING TO ALL CONTENTIONS OF THE JOINT INTERVENORS.
109. Identify all expert witnesses that' are expected to testity for the NRC Staff at each hearing on Joint Intervenors' Conte Ations in this matter, and state separately for each person identified:
(a) the subject matter on which the expert is expected to tes tify; (b) the substance of the facts and opinions to which the expert is expected to testify; (c) a summary of the grounds for each opinion.
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t 9
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- f. .
2 110. This interrogatory pertains to NRC Report Co. 50-483/
80-29, the enclosure referenced therein (SALP Evaluation), and all other SALP Evaluations, if any. ,
(a) Have SALP Evaluations been conducted for the Callaway Plant' by the NRC prior to the one referenced above?
If so, identify all documents reflecting such SALP Evaluations and NRC Reports pertaining thereto.
- 02) With regard to the statement on p. 2, 13.a. of 50-4 8 3/80-29, " the current trend of noncompliance !
is on an increase," provide- the bases for the state-ment including, but not limited to, identification of all documents reviewed in reaching the conclusion indicated and a statement of how the increase was measured and in relation to what time period the increase was measured.
With regard to the statement on p. 3, 54, " the minimal (c) size of the QA staff has contributed to the delay of l
1 identification of problems at the Callaway site,"
state (i) What is the size of the QA staff.
(ii)
What is the average size of the QA staff at other nuclear power plants currencly under cons truction.
(iii)
What is the minimally. acceptable size of
(
the QA staff for a plant such as Callaway l
Unit one and identify the document (s) upon which your. answer to this aragraph is based.
-.-x . , . _ _ . , , _
.intorregatory 110, con t' d (d) With regard to the point system atilized on the second page of the Licensee Performance Evaluation (the enclosure with 50-463/80-29):
(i) Explain how points are astigned to different areas of noncompliance.
(ii) Expl,ain the meaning assigned to the total points assessed, including an explanation and description of the scale used for comparison and a statement of the amount of total points that would result in a conclusion of inade-cuate performance.
(e) With regard to the statement on the third page of the Licensee Performance Evaluation ( the enclosure with 50-483/80-29) , "Because of observed problems at the Callaway site involving audit of vendors and receiving inspection in wnich quality assurance i
controls were not understood by inspecting groups. ..
state to what " observed problems" the statement refers, including the following information, separately l
for each problem:
(
(i) Identify the materials involved and their l
functions within the plant.
(ii) Identify the vendor of the mater .als.
(iii) The date of delivery to Callaway site.
(iv) The date on which the problem was discovered
- and explain in detail, iracluding identification 1
of personnel involved, how the prcblem was discovered.
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I.. orrogatory 110, cont' d (v) Identify all documents which pertain to the problem.
(f) With regard to the " dilution of staff controlling construct}.on activities" referred to on the third page of the Licensee Perfoi.mance Evaluation (the -
enclosure with 50-483/80-29), provide the bases for the indication that the staf'f has been diluted, including the varying sizes of the staff over the period in question and state in what specific areas of the ataff there has been dilution.
111. Identify, separately for each of the above interrogatories and of the subparts thereof, the person (s) providing the answer.
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