ML20009A967
| ML20009A967 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 07/10/1981 |
| From: | Lessy R NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | JOINT INTERVENORS - CALLAWAY |
| References | |
| NUDOCS 8107140558 | |
| Download: ML20009A967 (38) | |
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!LN UNITED STATES OF AMERICA 4 '/(( /Qp z
N NUCLEAR REGULATORY COMMISSION U%gOI A.
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BEFORE THE ATOMIC SAFETY AND LICENSING BOAR 98. %
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Docket Nos. STN 0:48 UNION ELECTRIC COMPANY
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STN 50-486 (Callaway Plant, Units 1 and 2)
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ANSWERS OF THE NRC STAFF TO JOINT INTERVEN0RS FIRST SET OF INTERROGATORIES AND REQUESTS FOR DOCUMENTS The NRC Staff herewith files its responses to Joint Intervenors First Set of Interrogatories and First Request for Production of Documents to the NRC Staff.
1.
Two stop work orders were issued on June 9, 1977.
(a) Stop work order #8 stated, "all structual concrete pour operations - no concrete to be placed in any forms until an investigation program is initiated.
Fill concrete with no embedments can continue" stop work order #9 stated, " Effective June 9,1977 until further notice the material control department will make no further issues to the field of embedments supplied under Bechtel P.O.10466-C-131-2.
(b) To prevent further embedments from being cast in concrete prior to completing an investigation.
2.
Yes.
During the review of the embedded plate matter between April and June 1980 the NRC Office of Inspection and Enforcement did accept four deviations from the.AWS Code D.1.1.
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(a) Those exceptions are identified in NRC Report 80-14, page 7 and FSAR section 3.8.3.6.4.3.
(b) The reason these exceptions are permissible is found in attachment 3 to NRC Report 80-14.
(c) NRC Report 80-14 found the four deviations to the AWS code to be permissible.
Further review by the NRC will be performed during the FSAR review.
3.
An audit of the Cives Steel Company was not performed by the NRC; however, audit reports developed by Union Electric Co. were reviewed by the NRC.
4.
The width of the crack mentioned on p. 20, entry 13a in NRC Report 78-01 in the " plant" north wall of the control building was one-sixteenth (1/16) inch at the widest place.
5.
15 hairline cracks.
6.
See Reports NCR 2-2081-C-A; 2-2173-C-A attached.
7.
The NRC did not state that such cracks are a " recurring problem."
NRC Report 78-01 states that "[T]he NCR [non-confcrmance report] prepared by Daniel states that the above-mentioned cracks were documelted in the l
NCR for information and to indicate a recurring problem."
8.
(a)
It is the NRC, Office of Inspection & Enforcerent's, view that the cracks described are caused by a combination of voldme Change (shrinkage) & restraint. Drying shrinkage of concrete is one of the principal causes of cracking. Snrinkage is an inherent characteristic of hydraulic cement concrete.
If shrinkage were allowed to occur without any restraint, concrete would not crack. Therefore, the combination of shrinkage and restraint can cause cracking (ref: ACI Report 224, Control
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of Cracking in Concrete Structure, Title #69-69).
Such shrinkage cracking is routinely accounted for in the design of reinforced concrete.
(b)
The "other cracks" described are caused for the same reasans as described above in (a).
9.
None, but the bulk of such shrinkage cracks should occur within the first week after pouring, although the subject concrete structural members are designed and constructed to minimize cracking. The structural elements are also designed to function with any such additional cracking.
These additional cracks will not compromise the structural integrity of the structure to function as designed.
10.
Insignificant increases could occur however, decreases could also occur with loading. Neither would have an adverse effect on the structural integrity of the system.
11.
No, there was no reason to as it was structurally insignificant.
12.
Yes.
13.
Simply, NCR 2-2173-C-A replaced the previously written NCR 2-2081-C-A by incorporating and updating the information.
14.
As previously stated in response to interrogatories No. 8, 9 and 10 which identifies the cause of the subject crack it is our assess-ment that the crack does not affect the design basis or the structural integrity of the structure.
15.
(a) Re NRC Report 77-06/Circumferential Concrete Crack in Reactor Cavity Moat Area: Length of crack is estimated to be 65', (270*
circumference, radius of 14').
1b (b) Depth of crack is depth of the M4Y" X 13 embedded struc-tural steel member or approximately 4".
3 (c) The width of crack was observed (noted in NCR #2-0631-C-A) to be approximately k" but the entire crack was removed and repaired.
(d) Proximity of the crack to embedded structural steel was 4" outward at top surface (construction Joint).
16.
Document on reactor cavity moat area circumferential crack is NCR #2-0631-C-A prepared by J.A. Petsche, Daniel International on 5/9/77.
The document is a nonconformance report of the crack described in No.15.
17.
Content of above document; A Nonconformance Report (NCR) with section as follows:
(1) Area of item, (2) Controlling Document (3)
Description of Nonconformance, (4) Recommended Disposition and Basis for Recommendation, (5) Cause of Honconformance and Action t: Prevent Recurrence, (6) Action taken to Control Nonconformance, (7) Approvals for Action Taken.
18.
(a) Name, title of all persons participating in the investiga-tion are identified on the above (Question 16) NCR including:
J.A. Petrickt.
Daniel, C.L. Miller, SNUPPS' Civil Group Supervisor; P.E. Diviak, Bechel, Project Engineer; Mr. White, SHUPPS, QA Manager; C. McFarland; C.B. Blieserres (signatures somewhat illegible).
l (b) The crack in question was repaired and no longer exists.
19.
This interrogatory misquotes the NRC Report No. 483/77-06, I
- p. 21:
required by NRC - " is incorrect, should be NCR.
(a) NRC inspector is A.A. Varela.
i (b) This question suffers from the misquote.
(c) Other than the documents identified above, the NRC has not retained these documents, they may be available from the licensee.
(d) This also suffers from the misquote; The repair of the nonconforming condition fulfilled NRC requirements identified in 10 C.F.R. 50 Appendix B.
Criterion XV; and Bechtel's Engineering Specification C-103.
(e) The NRC inspector physically inspected the property and adequately restored nonconforming conditions and verified through review of documentary evidence that 10 C.F.R. 50 Appendix B Criterion XV, XVI and XVII were met. The inspector augmented his documentary review for evaluation of the nonconformance disposition by interviews with respon-sible QC personnel.
20.
See Diagram Detail A of NRC 2-0631-C-A (attached).
21.
This question was answered in No.19(c).
22.
There is no 10 C.F.R., Part 50, Appendix 3.
23.
NCR #2-0631-C-A identifies cause to be:
related to shrinkage l
of C6&I liner plate in the moat area relating to the welding in this area.
24.
Document which establishes reporting standard for cracks in l
permanent concrete: Bechtel Power Corp. letter dated 3/29/77 Number BLSE l
4194 filed C-103, " Evaluation of Cracks in Structural Concrete," in reference to Daniel's #1119 dated 1/25/77 and Union-1246 dated 2/10/77.
25.
Documents attached (see No. 24).
26.
See answer to No. 6; the NRC did not state that it was a recur-ring problem; however, there have been approximately 17.
27.
See NCRS 2-1973-C (1/19/78); 2-SN-2611-C (10/9/80), 2-SN-3593-C (2/25/81); see also responses 31, 33, 37-38, el sea. The use of the word I
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" problem" at Callaway in this interrogatory is argumentative; a characte ization in this proceeding to which the Staff does not agree.
28.
Due to a clerical error, the response to this item is not avail-able today. The response will be mailed under separate cover in the coming weeks.
29.
The NRC inspector concurs in NCR #2-0653-C-A Cause of Nonconformance as primarily being:
insufficient vibration of the toe of the first concrete lift.
30.
Yes, the Staff has not made a detailed study of the differences in congestion by specific area.
In general, congestion of the type identified above occurs only above the tendon access gallery.
31.
This is answered in No. 29.
(a) Union Electric actions, subsequent to NCR 2-0856-C-A, to minimize recurrence of concrete imperfections as those in Tendon Access Gallery include:
" Closer supervision over vibration crews was maintained in order to minimize recurrence." Training sessions of the vibrator crews were also conducted. These actions had the effect of minimizing such defects.
(b) See i.nswer to (a) above.
(c) The cctions described in (a) fulfilled the requirements of 10 C.F.R. 50, Appendix B, Criterion XVI.
- 32. The relevant requirements relating to placement of concrete are contained in approved industry codes and standards and NRC Regulatory Guides committed to by Licensee and found acceptable by the NRC, as iden-tified in SAR. Any modifications in concrete placerrent could also be
e contained in Section 3.8 of the FSAR, a document which is public1v avail-able.
33.
(a)-(d)
The inadequacies in the procedural requirements had no discernible impact on QA or safety, but in any event was negligible.
Licensee responded to this Citation by revising WP-109 and QCP-109 both concerning concrete placement and pour to incorporate procedural changes consistent with explicit requirements of engineering specification C-103 Section 10.1. These documents would be available from Union Electric.
These revisions were dated November 3 and 7, 1977.
34.
The repairs of honeycombing in the Tendon Access Gallery did conform to the repair procedure detailed in NCR 2-0856-C-A and as speci-fied in the engineering specification C-103, Section 15 on in-process repair of concrete.
35-36. Mobility did not hamper repairs, for the repairs were to an exterior surface. The details of removing unsound concrete surrounding the main steel and in back of base plates were accomplished by one man (skilled laborer) and specialized treatment for replacement with grout was accomplished using acceptable industry techniques identified in engineering specification C-103, Section 15 on in process repair of con-crete.
37-38.
In process testing had not been done for the limited period 8/17-9/1/77, but subsequent to that time procedures for physical compres-sive strenth testing have been implemented. As a result, stop work order No.14 was changed to a start work order in December 1977. Because of material certifications and subsequent users tests, there is however no
evidence that the non-physically compressive strength material in the period described did not conform to manufacturer's specifications.
39.
Yes.
40.
The NRC conclusion that repair of voids in the Tendon Access Gallery has no adverse safety implications rests on quality control l
verification during the repairs that the repairs were adequate.
41.
The NRC inspector who reviewed the Soniscope Report of Wiss, 1
Janey, Elstner & Associates is A.A. Varela, l
42-43 See " Objections Of The NRC Staff To Joint Intervenors' First Set of Interrogatories."
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44.
This specific data would only be available through the
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Permittee at this time. Similar information may, however, be included in Revision 1 of the Final Report of the licensee regarding this matter, which is expected this summer, and which will be made public.
l 45.
No.
46.
The cause of honeycombing in the dome was subsidence of the l
concrete around the reinforcing steel at the unformed surface during placement.
47.
"rioneycombing" is not discussed in IE Report 80-30. The dome
" imperfections" in the licensee's estimation, were caused by subsidence of concrete around the reinforcing steel at the unformed surface during placement. The I&E staff concurs with this reasoning. Relative weight and probability is not germane to the dome imperfections because their actual extent of occurrence must be identified by the licensee as re-quired in IE Report 80-30.
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48.
NCR 2SN-2790-C was not prepared by the NRC; therefore the question of the basis for statements made therein should be addressed to the Permittee.
l 49-50.
Information regarding the applicable design specifications of the Permittee should be furnished by the Permittee.
l 51.
If imperfections are identified and properly repaired or other-vise evaluated, there will be no change in the dome's capubility to meet t
the original design criteria.
l 52.
The quoted sentence pertains only to the original four areas of l
concrete imperfection referenced in IE Report 80-30. Generally, the
" complex nature" referenced were associated with temporary construction I
aids (i.e. walkways, conveyors, and related supports).
53.
The term " voids" is not used in IE Report 80-30, but as described previously in these responses, there is no basis to causally associated the two matters, a
i 54.
This is to be addressed in UE Final Report Revision 1, which is due this Summer. The Staff has no basis to believe that imperfections of this type exist elsewhere.
55.
The testing methods employed by the licensee will be addressed in their Final Report, Revision 1, which the NRC Staff will then review.
Thus, supplementation of this response must await the completion of these two events.
56.
The use of ice to control mix temperatures during hot weather l
is acceptable and is recognized by the concrete industry (Portland Cement i
Association and American Concrete Institute). The use of ice can be confirmed through review of site batch plant records.
57.
The " actions" referred to in IE Report 80-30 includes the licensee's destructive and nondestructive testing programs on the dome.
The results of the testing will be included in the Final Report, Revision 1, when it is submitted.
58.
The material observed to be flaking off was troweled concrete mortar adjacent to a repaired area and is of no consequence.
59.
The item has not been formally closed in an NRC inspection report; such a report closing this item is imminent and will be made publicly avaflable.
60.
Documents are attached.
61.
Inspections of reinforcement steel were conducted between December 13, 1977 and January 6,1978 during which Union Electric and Becdhtel first discussed interpretation of concrete cover requirements for reinforcing steel. This item was identified in NRC Report 77-11, page 10 and Report 78-01, page 8-9 (item 3a(8)).
62.
(a) The reinforcing steel referred to in NRC Report 77-11 was the horizontal steel on the outer face of the reactor building wall 3rd concrete lift.
(b) Concrete cover requirements are set forth in the following codes and standards:
(1) ASME Section III, Division 2, article CC-3534.1 (cover).
reinforcement).
(3) ACI 349-76, Section 7.14 (concrete protection for reinforcement).
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(c) The following enclosed documents set forth the NRC Staff's views on concrete cover requirements:
(1) NRC Report 78-01, enclosure 1 (minutes of meeting on January 23,1978, page 7 of 11 (item 4 of summary and positions).
(2) NRC memorandum from I. Sihweil to 0. parr dated March 6, 1978.
(3) NRC letter 0. Parr to SNUPPS Licensee's dated March 13, 1978 (item 2 above as enclosure).
(4) HRC memorandum 0. parr to R. mattson dated July 12, 1978.
(5) NRC memorandum I. Sihweil to 0. parr dated July 28, 1978.
The above documents (1) through (5) are being provided.
(d) The following documents set forth the licensee's and contractor's interpretation of concrete cover requirements:
(1) NRC Pvport 78-01, enclosure 1 (mintues of meeting on January 23,1978, page 6 of 11.
(2) Letter dated February 13, 1978 from N. Petrick (SNUPPS) to E. Case (NRC).
l The above documents (1) and (2) are also being provided.
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63.
The NRC interpretation of concrete cover requirements prevailed at 340 degrees azimuth, however, as it was determined that the local area of reduced concrete cover would have no adverse impact on the structure
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and Union Electric Co. was permitted to leave the reinforcing steel as placed.
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64.
(a) Specification C-112 Rev. 9. Section 7.10 (tolerances) states, "unless noted or approved otherwise, the reinforcing steel shall be placed within the tolerances given in Section 7.3.2 of ACI 318 but in no case shall be cover be reduced by more than one-third of the specified Cover."
(b) The NRC position is that the one-third reduction in con-crete cover should not be utilized as a general provision of the con-struction specifications.
It's the NRC position that 2 inch minimum cover is a " minimum" to assure corrosion control. However, in the local areas below the sixth lift where the concrete cover was reduced below the 2 inch minimum it is the NRC Staff's view that it will not adversely affect the structural integrity of the containment structure, and is therefore acceptable.
65.
An ASME Section III, Division 2 code interpretation was re-quested of Brown & Root, Inc. on June 22, 1976 regarding tolerances on placement of reinforcing steel. The inquiry indicated tha article CC-3534 contained requirements on minimum concrete cover and that article l
CC-4342 required the designer to specify tolerances an placement of re-1 inforcement steel.
The following interpretation was requested:
"Is it reasonable to assume that minimum concrete cover nay be reduced as long as p7 acing tolerances in construction specifications are not exceeded? Paragraph 7.3.2 of the commentary to ACI 318 appears to suggest that such reductions are permissible provided that coverage is not reduced by more than one-third the original specified amount."
i.
The pruposed reply from the working group of ASME stated, "It is the opinion of the committee that minimum concrete cover... are absolute and may not be reduced as a function of placement tolerance... minimum cover given in the applicable specifications should include a margin for minus placement tolerances so that the minimum values of CC-3534 are not violated."
On October 21, 1977 a formal code interpretation (III-2-77-13) was issued which stated, "The minimum concrete cover require..:.cs of CC-3534 are not to be reduced as a function of placement tolerance.
In relation to the minimum spacing given in CC-3534, only plus toleranc>. cre permitted.
As a practical matter, minimum cover given in the applicable specifications should include a margin for minus placement tolerances so that minimum values of CC-3534 are not violated." A code change was not necessary per the above interpretation. The ASME code interpretation is being provided as an attachment to this response.
66.
NRC Report 78-01, enclosure 1, page 7 of 11 required the minimum cover of 2 inches be obtained by the 6th lift of the reactor building.
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l In order to accomplish this, it necessitated using the 5th lift as a transition area (i.e. a gradual shifting of the vertical steel inward through the 5th lift so that the horizontal steel could be tied to the outside of the vertical bars and maintain the minimum cover requirements).
67.
(a) Union Electric did comply with the NRC position taken on minimum cover requirements and revised the placement tolerances on re-inforcing bars to be "minus 0," "plus 1 " for exterior walls. This was reflected in FSAR submittal, Section 3.8.1.6.2.3.
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(b) No documents as to Union Electric's evaluation are avail-able from the NRC.
68.
The 23 NCRs are as listed in Attachment B.
The Attachment provides the designation, nonconforming condition, disposition where known, and date of the inspection which identified the nonconforming condition (data from NCR forms). This listing is substantially complete but is being updated and any additional information will be provided.
69.
The third lift was a relatively complex area of the reactor containment wall, containing large and small penetrations, buttresses, tendons and other items.
70.
Ten is not the correct number. This information should be directly available from the source, from the permittee in response to Joint Intervenors' Interrogatory 81.
71.
(a) The location and size of the areas were determined not to be significant in terms of adverse effects on the structure and, there-fore, the extent of the areas was not documented by specific location or size.
(b) NRC Report 78-01, enclosure 1 (minutes of meeting on January 22,1978) summarize the discussion on reinforcement with less l
l than 2 inches as specified on placement drawings (see item 2, p. 6 of 11 of referenced document).
This referenced document is being provided as an attachment.
(c) NRC Report 78-01, enclosure 1 provides the NRC Stafrf position regarding minimum cover requirements.
P. 7 of 11 states, "The staff position that the commitment of a 2 inch minimum concrete cover for the concrete containment as made in Section CC-3533.1 of Appendix C to
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BC-TOP-5 for #6 through #18 reinforcing steel to control design and con-struction. The value is a minimum, meaning the absolute minimum cover to assure corrosion control in the actual construction shall not be less than 2 inches.
The staff expects that by wall lift #6 all reinforcing in sizes #6 through #18 will meet this requirement."
The above referenced document has been provided as an attachment to these responses.
72.
(a) See answer to No. 71(a) above.
(b) NRC Report 78-01, enclosure 1 summarizes discussion of concrete cover of 12 to 13 inches which was greater than that permitted (see page 6 of 11 of referenced document).
(c) HRC Report 78-01, enclosure 1 provides the NRC position regarding maximum concrete cover requirements.
P. 8 of 11 states, The staff considers that the commitment of a depth of not more than t/5 to reinforcing steel that is considered face reinforcement as made in Section CC-3534 of Appendix C to BC-TOP-5 to contgrol design and construction.
The value is a maximum as rounded to the next whole inch, meaning the absolute maximum depth in order to provide surface crack control for the concrete containment. The staff expects that by wall lift #6 all face reinforcing will meet this requirement. The staff will consider special cases on this requirement where necessary wall blockouts may require local variations to the maximum depth to face reinforcing.
The above referenced document has been provided as an attachment to this response.
73.
The precise location and number of variations from CC-3533.1 of Appendix C to BC-TOP 5 for reinforcement beneath the sixth lift is not available at this time; however, based on observations of reinforcement installed in the third and subsequent lif ts it was determined that the
deviations identified regarding concrete cover would have no adverse effect on the safety of the structure to perform under design conditions.
74.
(a) See NCR 2-2055-C-A attached hereto.
(b) The nonconforming bars will not impair the integrity of the structure.
Based on the relatively small number of such exceedence and that the amount did not exceed 1", the Staff believes that these nonconforming bars have no adverse safety significance.
75.
Union Electric, Rechtel, and SNUPPS objected to the require-ments of a minimum concrete cover of two inches over reinforcing steel on the outer face of the reactor containrent. The Staff does not know whether these are similar individual or joint objections.
(a)(b)&(c) The basis for their objection, details and reason set forth are contained in SHUPPS letter to E. Case dated February 13, 1978.
The above referenced document has been provided as an attachment to this response.
76.
Yes. Union Electric, Bechtel and SHUPPS did object to the requirements of a maximum concrete cover on the fice reinforcing steel of I
ten inches. The Staff does not know whether these are similar, indivi-dual or joint objections.
l (a)(b)&(c) The bases for their objections, details and reason set forth are contained in SNUPPS letter to E. Case dated February 13, 1978.
The above referenced document has been provided as an attachment to this response.
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77.
The NRC Staff does not believe that the reduction in cover, as observed, has reduced or compromised these properties.
78.
The NRC Staff does not believe that exceedance in cover, as observed, has reduced or comprised the structural system's ability to control cracking.
79.
For buildings other than the reactor building, UE was in com-pliance with the provisions of ACI 318-71.
81.
Essentially the same questions were posed in a letter to Mr. Gerald Phillip dated February 8,1981. The NRC response to the questions is attached he'ceto as Appendix B.
In addition as to81(b),
t'.ie NRC persons having firsthand knowledge of the allegations are James McCarten, Investigator, Region III; Robert Burton, Investigator, Region III; William Hansen, Callaway Sr. Resident Inspector; James Foster, Investigator, Region III, William Key, Inspector, Region III.
82.
(a) NUREG-0017 was published in April 1976. Regulatory Guide 1.21 was originally issued as Safety Guide 21 on December 29, 1971.
Regulatory Guide 1.112 was published in April 1976.
(b) NUREG-0017 has not been revised. Safety Guide 21 was issued and published as Regulatory Guide 1.21 in June 1974.
Regulatory Guide 1.112 was revised in May 1977.
(c) All documents were published upon completion in final form.
(d) Revisions to all documents were published upon completion infinal form.
(e)-(f) See " Objections Of The NRC Staff To Joint Intervenors' First Set Of Interrogatories."
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83.
fdCLEAR FACILITIES HAVING BEEN ISSUED AN OPERATING LICENSE AFTER JUNE 1, 1974 DATE LICENSED UNIT LOCATION OF ISSUE Arkansas 2 Russellville, AK 07/18/78 Beaver Valley 1 Shippingport, PA 01/18/76 Browns Ferry 2 Decatur, AL 06/28/74 Browns Ferry 3 Decatur, AL 07/02/76 Brunswick 1 Southport, NC 09/08/76 Brunswick 2 Southport, NC 12/27/74 Calvert Cliffs 1 Lusby, MD 07/31/74 Calvert Cliffs 2 Lusby, MD 08/13/76 Cook 1 Bridgman, MI 10/25/74 Cook 2 Bridgman, Ill 12/23/77 Crystal River 3 Red Level, FL 12/03/76 Davis-Besse 1 Oak Harbor. 0H 04/22/77 Farley 1 Dothan, AL 06/25/77 Farley 2 Dothan, AL 10/23/80 Fitzpatrick Scriba, NY 10/17/74 Haddam Neck Haddam Neck, CN 12/27/74 Hatch 1 Baxley, GA 08/06/74 Baxley, GA 06/13/78 Hatch 2 Indian Point 3 Buchanan, NY 12/12/75 Millstone 2 Waterford, CN 08/01/75 Nine Mile Point 1 Scriba, NY 12/26/74 North Anna 1 Mineral, VA 11/26/77 North Anna 2 Mineral, VA 08/21/80 Oconee 3 Seneca, SC 07/19/74 Peach Bottom 3 Peach Bottom, PA 07/02/74 Prairie Island 2 Red Wing, MN 10/29/74 Rancho Seco 1 Clay Station, CA 08/16/74 Salem 1 Salem, NH 08/13/76 Sequoyah 1 Daisy, TN 02/29/80 l
St. Lucie 1 Fort Pierce, FL 03/01/76 i
Three Mile Island 2 Middletown, PA 02/08/78 l
Trojan Prescott, OR 11/21/75 84.
There will be no change in the fuel fission products.
In the comparison of Hf absorber activation products versus AG-In-Cd activation products, the hafnium isotopes that are produced either have relatively short half lives (and thus do not provide a long-term storage problem) or are stable. Both the hafnium and the AG-In-Cd are clad in stainless steel, and, therefore, are not susceptible to corrosion.
Even if the l
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cladding were perforated, a Hf cortosion problem would not be expected, as evidenced by fact that hafnium is used unclad in naval reactor appli-cations.
85.
Indian Point 1. Yankee Rowe, and Shippingport have utilized hafnium rods.
86-87. See " Objections Of The NRC Staff To Joint Intervenors' First Set of Interrogatories."
88.
No.
89.
The effluent release monitor set-points are based on the maximum permissible concentrations in 10 C.F.R. Part 20, Appendix 8. Table II,
' Column 2 for releases to the hydrosphere.
For releases to the atmosphere, the noble gas monitor set points are based on an annual dose of 500 mren to the total body and 3000 mrem to the skin (implied dose levels in 10 C.F.R. 20, Appendix B. Table II, Column 1).
The EPA Uranium Fuel Cycle Standard, 40 C.F.R. Part 190, is an ambient dose standard that limited individual exposure from all fuel cycle sources (except mining activities and radon plus its daughters).
90.
The PWR-GALE Code calculates that there will be approximately 390 curies /yr of tritium released in the liquid effluent.
This figure is determined by the PWR-GALE computer code which is a mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents from pressurized water reactors.
l The PWR-GALE does not calculate tritium release fractions due to fission and activation.
The code calculates the annual quantity of tritium available for release using a functional relationship derived from measured j
liquid and vapor tritium releases at operating PWRs and the integrated l
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thermal power output during the calendar year in which the releases occur.
The relationship expresses the total tritium as a function of power output.
The quantity of tritium released through the liquid pathway is based on the calculated volume of liquid released, excluding secondary system wastes, with a primary coolant concentration of tritium of 1.0 uCi/ml up to a maximum of 50% of the total quantity of trit um calculated to be f
available for release.
The estimate of 41o curies / year is the number that is reported 91.
This value was reported by in the SHUPPS Final Safety Analysis Report.
the Applicant as the results obtained from running their version of the NRC finds that 390 curies of tritium per year are estimated PWR-GALE Code.
to be released by running the NRC version of the PWR-GALE Code.
The Staff is not aware of any such study; if you are aware of 92.
such a study, it is requested that yota bring it to the attention of the NRC Staff, in this proceeding at this t.ime.
Dissolved noble gases relesed in the liquid are estimated 93.
(a) to be approximately 0.1 Ci/yr.
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(b) Gaseous effluents include:
Nuclide Ci/yr Kr-83m 3
Kr-85m 33 Kr-85 260 Kr-87 8
Kr-88 48 Kr-89 all Xe-131m 16 Xe-133m 66 Xe-133 3500 Xe-135m 41 Xe-135 130 Xe-137 42 Xe-138 2
94.
The PWR-GALE Code calculates that there will be 1000 C1/yr released in the gaseous effluent. The PWR-GALE Code does not calculate tritium release fractions due to fission and activation.
i 95.
For short term accidents requiring consideration pursuant to 10 C.F.R. Parts 50 and 100 of the NRC regulations, i.e. design basis acci-l l
dents, the atmospheric dispersion characteristics of solid particulates l
are taken to be the same as those of gaseous materials.
j 96.
The foregoing is equivalent to assuming that the particulates are of cufficiently small size as not to be significantly subject to driving forces that would tend to separate them from a gaseous plume.
Atmospheric dispersion models displaying those characterit;Je, therefore, are those described in the NRC Regulatory Guide 1.145, a document which is publicly available.
97.
In Section 12.2 of the Staff's Standard Review Plan,Section IV l
" Evaluation Findings" state that "The source terms used to develop these airborne concentration values are comparable to estimates by othe appli-cants with similar designs, and are acceptable." Since Callaway is using
Fort Calhoun Operating Data ::r their expected average airborne radio-activity concentrations (Table 12.2-10), and since Fort Calhoun is a PWR, as is Callaway, then it would appear that very relevant experience is being used by Callaway in their source term evaluation for potential airborne radioactivity. These source terms are reviewed by the Staff and would be the bases for Staff acceptance.
98.
Reg. Guide 1.11 rev.1 authored by the NRC Staff states, "For conservative estimates of radioactive decay, an overall half-life of 2.26 days is acceptable for short-lived noble gases and of 8 days for all iodines released to the atmosphere. Alternatively, the actual half-life of each radionuclide may be used."
Tne NRC Staff considers this choice of 2.26 days and 8 days to be conservative (i.e. an overestimate of the actual circumstance) and there-fore acceptable.
99.
During refueling operations reactor coolant mixes with water in the transfer canal and spent fuel pool, thus adding itium to the spent fuel pool. After the plant is in operation, any make-up water added to the pool may also contain tritium.
After the first several years of l
operation, the total tritium release rate from the plant is approximately equal to the amount entering the primary coolant. Thus, significant increases in the amount of tritium in the spent fuel pool after the first several years of plant operation are not expected.
100. The Staff is not aware of any plans for the Applicant to expand his spent fuel pool capacity using high density spent fuel storage racks.
Therefore we have not reviewed a Callaway submittal for this expansion and the consequences commensurate with the change in the rate of evapor-
.' ion of the spent fuel pool water and subsequent change in airborne radioactivity. However, from past experience, the consequences of such expansion has resulted in a minimal increase in the 85Kr release with an inconsequential impact upon occupational or population doses.
101. The temparary storage identified above is actually the minimum time required before the spent fuel would be transferred to the spent fuel shipping cask and then offsite. This is a conservative assumption for the handling and transferring of fuel due to the short time for decay of fission products. However, the spent fuel pool, cooling system, criticality analyses and purification system design parameters are all based on the fuel being permanently stored in the spent fuel pool. This is the most conservative assumption for the design of these systems and tne criticality analyses. Therefore, this interrogatory is responded to in the FSAR since the design parameters are based on indefinite fuel storage.
102. This question is also covered by the same reasoning as 101, since the evaporation rates shown in Table 9.1-4 of the FSAR are based on Therefore, an indefinite storage pericd rather than the temporary storage.
I the evaporation rate does not have to be controlled since the evaporation rates given are for the worst possible cases and permanent storage is 4
assumed.
Tritium is monitored at the release points either at the ventila-103.
l tion exhausts, if it is airborne, or in the liquid radioactive waste streams if it is water born.
However, these would take into consideration all sources of tritium vithin the plant not just releases from the spent fuel pool.
l
104. The presumption of " increased fuel rod failure caused by lor.g-term storage in spent fuel pools" is without substantive basis. However, if aj what ever means the fuel rod cladding were breached, thus exposing the pellets to spent fuel pool coolant, the leach rates would vary from about 2
2 g fuel /cm.d for Cs to 2 X 10-9 g fuel /cm.d for Pu, depending 3 X 10-0 l
somewhat on the purity of the water and length of tine of fuel exposure.
These estimates are based on information provided in the open literature.
For further information please refer to the following reports:
i 1.
A.D. Mitchell, J.H. Good, and V.C.A. Vaughn, " Leaching l
of Irradioted Light Water Reactor Fuel in a Simulated Post Accident Environment," 0Rf4L/TM-7546, May 1981.
2.
Y.B. Katayama, " Spent LWR Fuel Leach Tests," PHL-2782, April 1979.
105. The NRC Staff d.1s not make quantitative estimates of the length of time spent fuel rods can be stored in spent fuei pools before increased comparting would occur due to structural degradation. The reason is that signi.icsnt degradation would not be expected during the lifetime of the plant (usually 30 to 40 years). Moreover, if compaction of an assembly were to somehow occur, there wou'd De a decrease not to increase in criti-l 1
f cality.
106. Because the cladding is in a very benign environment in the spent fuel pool (as compared with an in-reactor environment) in terms of temperature, pressure, irradiation fluence, and hydraulic loading, the percentage of fuel cladding that would be expected to fail over a one to thirty year period is essentially zero.
107. The NRC does not ordi :rily require a specific monitoring frequency for isotopes in the spent fuel pool. Operating and chemical l
monitoring frequency of the spent fuel pool cleanup system is established by the utility and reviewed by the NRC to assume it meets the requirements of title 10 C.F.R 20.1(c) as it relates to maintaining radiation exposures as low as reasonably achievable.
108. The spent fuel pool is not normally discharged to the liquid waste system or any other discharge path to the environment.
109. This matter will be determined once the Staff examines the bases for Joint Intervenors' contentions as requested in the Staff's interrogatories. From that, it can be discerned which, if any, of Joint Interenors' contentions will survive summary disposition. As to the first proceeding, construction defects, it is presently anticipted that the individuals identified in response to interrogatory 111 will form the bases of the Staff's witnesses. This interrogatory will be updated once Joint Intervenors' responses to Staff and Permittee interrogatories is reviewed.
110.
(a) No.
(b) The basis for this observation was comparing the number of items of noncompliance and inspectors' hours in the selected uonths of July 1, 1980 - August 31, 1980 compared with prior two month periods.
In the later period, there were a few more items of noncompliance, and this was the basis for the observation that there had been an increase.
- However, the numbers should be normalized to accurately reflect any differences in l
the number of inspections during the relevant time period, as well as the addition of a resident inspector. When that is done, the more accurate reading of the trend is that there was no discernible increase, in fact l
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l '
there was a small (0.5%) decrease during the compared periods. The docu-ments reviewed were the inspection reports during the relevant periods.
(c)
(1) At "the time of the SALP meeting, the Permittees" QA Staff consisted of 8 personnel, which figure includes 3 consultants.
l This figure does not include the Permittee's corporate QA personnel.
(ii)
" Average" may be defined many ways, but 8 would be an approximate " average."
(iii) There are no documents specifying the minimally acceptable size. The licensee may also delegate this authority (but not responsibility) to consultants or subcontractors. However, the licensee is expected to meet the requirements set forth in 10 C.F.R. Part 50, Appendix B.
(d)
(1)
Infraction = 10 points, deficiency = 2 points.
(ii) The point system can be used as one of the tools for comparison with other facilities at approximately the same stage of con-struction considering the number of inspections and inspection man-hours during an appraisal period. There is no set number as to defining un-acceptable performance.
(e)
(i) The materials involved were four inch (4") pipe spools used for the steam generator No. 4 blowdown line to the flash tank.
(ii) The vendor was DRAVO Corporation, Pipe Fabrication Division,1115 Gillman Avenue, Marietta, OH 45750.
(iii) Material was delivered on March 22, 1979.
(iv) See Report 50-483/80-04.
(v) See Report 50-483/80-04. Any additional documents are in the possession of the licensee.
(f) Permittee personnel had been reducu from approximately 8 to approximately 6.
However, these persons ~ere replaced by 3 consultants.
This is permissible, see subpart (c) above. As to the technical areas of the replacement consultants, this information should be available from the Penaittee.
111. The NRC Staff counsel, Roy P. Lessy, and the Project Manager, Gordon Edison, are routinely involved in such responses.
In addition, see affidavits attached.
In addition, all docunants requested have been enclosed. With respect to Joint Intervenors' separate "First Request For Production Of Documents To NRC" document 3 has been provided.
As to document 4, it is requested that you provide the NRC Staff with the illegible copy that you have, and we will endeavor to obtain a legible copy.
Based on the identi-fying information you submitted, the document could not be located. As to document request 1 no such rainutes or transcripts as requested exist.
As to document request 2, the only report that has been prepared is the one that Joint Intervenors state is already in their possession. There are no NRC documents for category. There may be NCR's (e.g. Daniel prepared documents) which should be obtained from the licensee.
l l
l l
Attorney Making Objections:
Roy P. Lessyi dr.
l Deputy Assistant Chief Hearing Counsel Dated at Bethesda, Maryland this 10th day of July, 1981.
e UNITED STATES OF AffERICA NUCLEAR REGULATORY C0ft11SSION l
B_EF_0RE THE ATOMIC SAFETY AND LICENSIIM BOARD I
In the Matter of UNION ELECTRIC COMPANY
)
Docket Nos. STN 50-483 1
STN 50-486 (Callaway Plant, Units 1 and 2) i i
l AFFIDAVIT OF WILLIAM A. HANSEN NOW COMES W. A. HANSEN AND BEING DULY SWORN, DEPOSES AND SAYS AS FOLLOWS:
1.
I AM EMPLOYED BY THE U. S. NUCLEAR REGULATORY COMMISSION AS RESIDENT INSPECTOR.
2.
I AM DULY AUTHORIZED TO ANSWER QUESTIONS 1-10(F), 81 AND 110 AND I HEREBY CERTIFY THAT THE ANSWER GIVEN IS T, RUE TO THC BEjT OF)(Y
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Subscribed and sworn to before me this M day of W4 7 f j
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NUCLEAR REGULATORY COMMISSION i
BEFORE THE ATOMIC SAF2TY AND LICENSING BOARD 1
In the Matter of UNION ELECTRIC COMPANY
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Docket Nos. STN 50-483 i
(Callaway Plant, Units 1 and 2) l STN 50-486
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AFFIDAVIT OF F. C. HAWKINS l
Now comes F. C. Hawkins and being duly sworn, deposes and says as follows:
1.
I au employed by the U.S. Nuclear Regulatory Commission as ReactorInspector(R-III).
2.
I an duly authorized to answer Questions 44-59 and I hereby certify that the answer given is true to the best of ny knowledge, d
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F. C. Hawkins Subscribed and sworn to before me this / 3" day of July,1981.
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My Commission expires:
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE AT0 HIC SAFETY AND LICENSING BOARD In the Matter of
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UNION ELECTRIC COMPANY Docket Nos. STN 50-483
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STN 50-486 (Callaway Plant, Units 1 and 2)
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AFFIDAVIT OF JAMES E. FOSTER Now comes James E. Foster, and being duly sworn, deposes and says as follows:
1.
I am employed by the U.S. Nuclear Regulatory Commission as Investigator RIII.
2.
I am duly authorized to answer Questions 68, 69, 70, 80, and 81 and I hereby certify that the answer given is true to the best of my knowledge, M
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.es E. Foster Subscribed and sworn t., before this /# ay of July,1981.
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My Commission expires
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UNITED STATES OF At! ERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of UNION ELECTRIC COMPANY
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Docket Nos. STN 50-483
)
STN 50-486 (Callaway Plant, Units 1 and 2)
)
AFFIDAVIT OF EUGENE J. GALLAGHER Now comes Eugene J. Gallagher, and being duly sworn, deposes and says as follows:
1.
I am employed by the U.S. Nuclear Regulatory Commission as Senior Civil Engineer, Office of Inspection and Enforcement, Reactor Engineering Branch, Mechanical, Structural and Metallurgical Section, Division of Resident and Regional Reactor Inspection.
2.
I am duly authorized to answer Questions 1-14, 26, 27, 60-67, and 71-79 and I hereby certify that the answer given is true to the best l
of my knowledge.
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Eugene 1. GalTagher V
Subscribed and sworn to before me this / 6 rday of July,1981.
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UNITED STATES OF AMERICA nn NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOAR
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In the Matter of Docket Hos. STN 50-483 STN 50-486 UNION ELECTRIC COMPANY (Callaway Plant. Units 1 and 2)
AFFIDAVIT OF ANTHONY A. VARELA_
Now comes Anthony Verela, and being duly sworn, deposes a follows:
I am employed by the U.S. Nuclear Regulatory Connission as 1.
Reactor Inspector, IE-1.
I am duly authorized to answer Questions 15-25 and 29-41, 2.
best of my and I hereby certify that the answer given is true to the knowledge.
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i Subscribed and sworn to before ne this /dFday of July,1981.
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My Connission expires//
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O UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE AT0f1IC SAFETY AND LICENSING BOARD In the Matter of UNION ELECTRIC COMPAllY Docket Nos. STN 50-483 STN 50-486 (Callaway Plant, Unit; 1 and 2)
IDAVIT OF CHARLES MILLER How comes Charles Miller and being duly sworn, deposes and says as follows:
1.
I am employed by the U.S. Nuclear Regulatory Commission as riuclear Engineer, Effluent Treatment Systems Branch.
2.
I am duly authorized to answer Interrogatories 82, 83, 89, 90, 91, 93, 94 and 108 and I hereby certify that the answer given is true to the best of qy knowledge.
G Charles Miller Subscribed and sworn to before me this day of July,1981.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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UNION ELECTRIC COMPANY Docket Nos. STN 50-483 STN 50-486 (Callaway Plant, Units 1 and 2)
AFFIDAVIT OF FRANK CONGEL Now comes Frank Congel and being duly sworn, deposes and says as follows:
1.
I am employed by the U.S. Nuclear Regulatory Connission as Lpc t See t...,
Leoke,2>L,loyiai 2.
I am duly authorized to answer Interrogatories 88, 89, 97, 98 and 103 and I hereby certify that the answer given is true to the best of nly knowledge.
w Frank Congel
/
Subscribed and sworn to before me this day of July,1981.
Ttti.)0 CYwL Katary Public
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~7!/ !? 9 My Comission expires:
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I UNITED STATES OF #iERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of UNION ELECTRIC COMPANY Docket Nos. STN 50-483 STN 50-486 (Callaway Plant. Units 1 ard 2)
AFFIDAVIT OF SEYMOUR BLOCK Now comes Seymour Block and being duly sworn, deposes and says as follows:
1.
I am employed by the U.S. Nuclear Regulatory Comission as Senior Health Physicist.
2.
I am duly authorized to answer Interrogatory No.100 and I hereby certify that the answer given is true to the best of Iqy knowledge.
litM /
2 Seymodr Block Subscribed and sworn to before me this day of July,1981.
l A$ 0 Terxmu Rotary Public l-
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My Comission expires:
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j UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of UNION ELECTRIC COMPANY Docket Nos. STN 50-483 STN 50-486 (Callaway Plant, Units 1 and 2)
CERTIFICATE OF SERVICE I hereby certify that copies of "ANSUERS OF THE NRC STAFF TO JOINT INTERVENORS FIRST SET OF INTERROGATORIES AND REQUESTS FOR DOCUMENTS" in the above-captioned proceeding have been served on the followino by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Comission's internal mail system, this 10th day of vuly,1981:
James P. Gleason, Esq., Chainnan Barbara Shull Atomic Safety and LicenQ ng Board Lenore Loeb 513 Gilmoure Drive League of Women Voters of Missouri Silver Spring, MD 20901 2138 Woodson Road St. Louis, MO 63114 Mr. Glenn 0. Bright
- Atomic Safety and Licensing Board Mar.iorie Reilly U.S. Nuclear Regulatory Commission Energy Chairman of the League of Washington, DC 20555 Women Voters of Univ. City, MO l
7065 Pershing Avenue Dr. Jerry R. Kline*
University City, M0 63130 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Gerald Charnoff, Esq.
Washington, DC 20555 Thomas A. Baxter, Esq.**
l Shaw, Pittman, Potts & Trowbridge Mr. John G. Reed 1800 M Street, N.W.
Rt. 1 Washington, DC 20036 Kingdom City, MO 65262 Dan I. Bolef 1
Treva J. Hearne President, Board of Directors Assistant General Counsel for the Coalition for the Environment, Missouri Public Service Commission St. Louis Region P.O. Box 360 6267 Delmar Boulevard Jefferson City, MO 65101 University City, M0 63130
- Served with enclosures.
l l
. Donald Bollinger, Member Rose Levering, Member Missourians for Safe Energy Crawdad Alliance 6267 Delmar Boulevard 7370a Dale Avenue University City, MO 63130 St. Louis, MD 63117 Mr. Fred Luekey Presiding Judge, Montgomery County Rural Route Rhineland, MD 65069 Mayor Howard Steffen Chamois, M0 65024 Professor William H. Miller Mr. Earl Brown Missouri Kansas Section, School District Superintendent American Nuclear Society P.O. Box 9 Department of Nuclear Engineering Kingdom City, N 65262 1026 Engineering Building University of Missouri Mr. Samuel J. Birk Colunbia, MO 65211 R.R. #1, Box 243 Morrison, MO 65061 nr. Harold Lottman Presiding Judge, Dasconade County Robert G. Wright Rt. 1 Associate Judge, Eastern District Owensville, MO 65066
' County Court, Callaway County, Missouri Eric A. Eisen, Esq.
Route #1 Birch, Horton, Bittner and Monroe l
Fulton, M0 65251 Suite 1100 1140 Connecticut Avenue, N.W.
l Atonic Safety and Licerising Washington, DC 20036 Board Panel
- U.S. Nuclear Regulatory Commission Docketing and Service Section*
Washington, DC 20555 Office of the Secretary U.S. Nuclear Regulatory Commission Atonic Safety and Licer. sing Washington, DC 20555 Appeal Board
- U.S. Nuclear Regulatory Commission Washington, DC 20555 Kenneth M. Chackes**
Chackes and Hoare Attorney for Joint Intervenors 314 N. Broadway St. Louis, Missouri 63102
>V
~Roy P. Lessy
- Served with enclosures.
Deputy Assistant Chief Hearing Counsel
UNITED STATES OF AllERICA NUCLEAR REGULATORY COMMISSION BEFORE THE AT0f1IC SAFETY AND LICENSING BOARD In the flatter of
)
UNION ELECTRIC C0l1PANY Docket Nos. STN 50-483
)
STN 50-486 (Callaway Plant, Units 1 and 2)
)
AFFIDAVIT OF 14ICHAEL TOKAR Now comes Michael Tokar and being duly sworn, deposes and says as follows:
1.
I am employed by the U.S. Nuclear Regulatory Comission in the reactor fuels section, Core Performance Branch.
2.
I am duly authorized to answer Interrogatories 84, 85, 104, 105 'and 106
, and I hereby certify that the answer given is true to the vest of my knowledge.
h4 MX a
' Michael Tokar Subscribed and sworn to before me
. thi s /c / /i day of July, 1981.
(
nds 'I D rQEr u _)
Notary Public d)
My Commission expires: sun /
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