ML20085D777
| ML20085D777 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 07/26/1983 |
| From: | Gallo J, Goldfein M COMMONWEALTH EDISON CO., ISHAM, LINCOLN & BEALE |
| To: | |
| Shared Package | |
| ML20085D770 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8307290105 | |
| Download: ML20085D777 (20) | |
Text
Edison - 7/26/83 9
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In The Matter of )
)
CCMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454 OL
) 50-455 OL
-(Byron Nuclear Power Station, )
Units 1 & 2) )
APPLICANT'S REPLY TO INTERVENOR ROCKFORD LEAGUE OF WOMEN VOTERS' PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW REGARDING STEAM GENERATOR TUBE INTEGRITY ,
As authorized by Stipulation among the parties and by Memorandum of Board Approval of Time Extension, Common-wealth Edison Company ("Ediso'n" or " Applicant") files the following reply to Intervenor Rockford League of Women Voters' (" League" or "Intervenor") Proposed Findings of Fact and Conclusions of Law Regarding Steam Generator Tube Integrity, which were filed on July 1, 1983.
Introduction The proposed findings of fact and conclusions of law on League Contention 22 and DAARE/ SAFE Contention 9 (c) -- Steam Generator Tube Integrity -- submitted by the League are organized as follows: types of tube degradation; I
steam generator design, flow-induced vibration; the loose-parts monitoring system; pre-service and in-service in-spections; water chemistry; tube rupture events; unresolved 8307290105 830726 PDR ADOCK 05000454 O PDR
safety issue A-3; and conclusion. Applicant's reply to the League's proposed findings on the steam generator issue are organized according to the above categories. Applicant's reply findings demonutrate that the proposed findings the League seeks to have this Board adopt are replete with misstatements, obvious misunderstandings concerning the evidentiary record, and wholly unsupported assertions.
Accordingly, Applicant submits that the findings discussed below should be rejected by the Board.
Types of Tube Degradation The League on page 8 (unnumbered) of its Opinion, citing Finding 17, states that " Denting will occur at Byron Unit 1. It will take from 3-10 years for initiation, depending on how well the water chemistry guidelines and condenser inservice inspection program are adhered to."
First, Mr. McCracken stated in his prepared testimony that Byron Unit 1 "could have minor denting within three to ten years" (emphasis added) rather than that denting "will" occur within three to ten years. (McCracken, NRC Staff Prepared Testimony at 6, ff. Tr. 4473.) Second, in response to extensive cross examination by the Board, Mr. McCracken explained that with proper secondary system control at Byron, it would take approximately three times longer (ten years, rather than three yearc) for initiation of denting.
(McCracken, Tr. 4772.) Further, Mr. McCracken emphasized that the minor denting that would occur would not affect steam generator tube integrity. For example, he stated:
t_
O
" Denting is, in fact, self-limiting in most cases. A steam generator which has minor denting can continue to operate for extended periods of time, There's no reason to anticipate that if you have minor denting in a unit, in that time frame that you cannot continue to run it, because, as discutsed in some of the other testimony, when you dent a tube, j ou are not affecting the primary system integrity. You are simply squeezing in on the tube a little bit. You cause some increase in stress in the tube. But it would take an enormous amount of denting to have that amount of stress affect its corrosion performance. Also, the amount of denting is not--does not restrict the flow in the reactor coolant system, so although it's something that you don't particularly like to have, it is not a problem which addresses the integrity or the ability of the unit to perform its function and protect public health and safety."
(pl. at 4773.)
After a technical explanation of the denting phenomenon ([d.
at 4774-76), Mr. McCracken went on to state:
[F] rom a public health and safety concern, I would not anticipate tha't denting would be a problem over the lifetime of the Byron plant. They will have it. It will be an operational concern. But I would not anticipate that denting at this plant, unless they callously disregard what they are committing to as far as their chemistry, would ever become a public health and safety problem; i.e.,
something that would initiate tube failures.
. . . [T]he denting that will occur will be to a minor extent, and not enough to force inner-rolled tubes, as we were talking about Surry, where you physically deform an already deformed tube and put excessive stresses into it. (Id. at 4777-78.) ~-
Design Features (Flow-Induced Vibratic i.f On page 12 of the opinion, Intervenor, citing Finding 45, states that "ninety percent or 100% main feed L
i flow through the main nozzle is found to contribute I
much more significantly to wear function than operation at lower power levels." Further, the whole discussion on page 12 appears directed to questioning the efficacy of a 90-10 flow split, verses a 70-30 split. However, as stated in Applicant's Proposed Findings of Fact and Conclusions of Law, submitted June 7, 1983, at page 10, Applicant's witness, Mr. Timmons, acknowledged that if there were no tube expansion, a split equivalent to a 70-30 split at Krsko (which translates to approximately a 75-25 split for Byron) would be needed to significantly reduce flow-induced vibration. (Timmons, Tr.
6016.) However, Mr. Timmons pointed out that both the model testing on the effect of tube expansion with different flow velocities as well as instrumentation before and after expansion of a tube at Krsko reveal that tubes have lower vibration levels with a 90-10 split and tube expansion than they would with the acceptable condition of a 70-30 split and no tube expansion. (Timmons, Tr. 6230.) Intervenor, in fact, recognized this conclusion on page 18 of its Opinion, citing Finding 217.
On page 15, Intervenor states that Westinghouse conducted accelerated corrosion testing to assess the i I
effects of the reduced tube-to-tube hole clearance on the i potential for denting in expanded tubes, but also remarks that Applicant did not provide evidence on this matter at the hearing, and it is difficult to simulate aged degradation.
L.
i Intervenor goes on to acknowledge, however, that according to Westinghouse, the results of this testing indicate that the potential for denting is not increased for tubes expanded at the baffle intersections. Obviously, the prepared testimony of Mr. Timmons, which states that the Accelerated potential for denting is not increased for tubes expanded at the baffle intersections, constitutes evidence on that matter. Perhaps, Intervenor means that the underlying data was not introduced into evidence. However, Mr. Timmons was subject to cross-examination which provided ample opportunity to have such data introduced into evidence. In addition, Intervenor provides no citation in the record for its assertion relating to the difficulty of simulating aged degradation. ~
Although Intervenor's assertions on page 15 of the Opinion, citing Findings 64 and 65, that it is not known exactly which of the 100 tubes identified as candidates for expansion will be selected is correct, it should be pointed out that preliminary identifications have been performed; it is only the final selection as to exactly which tubes will be expanded that is not yet completed. (Timmons, Tr.
6038.) Further, Mr. Timmons testified that the exact number and locr. tion of the tubes to be expanded is the only design detail remaining with respect to the proposed modification.
(pd. at 6306.)
On page 16 of the Opinion, Intervenor, citing Findings 69-71, discusses a discrepancy in Mr. Timmons'
testimony with respect to the necessity of plugging expanded tubes over the lifetime of the plant. A complete reading of the record, however, clearly discloses Mr. Timmons' belief that Westinghouse's proprietary tube vibration data dem-onstrate that with a 90-10 flow split, none of the approx-imately 100 expanded tubes in each steam generator will reach the 40% tube wall degradation level requiring plugging as a result of flow-induced vibration over the life of the plant. (Timmons, Tr. 6198-99, 6202, 6264-65.) Mr. Timmons explained that his initial statement that a small number of expanded tubes might have to be plugged over the life of the plant was based on a consideration of Westinghouse data involving bypassing flow that did not include the effects of tube expansion. When he revi~ewed the data involving the effects of flow reduction in combination with expansion, he corrected his earlier statement. (Id,. at 6201-02, 6264-65.)
On page 16 of the opinion, Intervenor asserts, citing Finding 72, that Timmons did not know how many more hours it would take for the tube with 6% degradation removed from Krsko in May to reach 40% degradation given the same power level ranges for the period January-May of 1982. This assertion leaves the impression that there was no way to calculate the number of additional hours it would take for such a tube to reach 40% degradation. Contrary to this implication, Timmons stated only that he did not calculate a tube-wear rate, but that he was sure somebody did. He then went on to discuss factors that are considered in correlating t
tube wear with vibration. (Timmons, Tr. 5978-79.) Timmons also testified that the rate of through-wall penetration actually decreases with time and as wear increases. (pd.
at 5980-81.) Moreover, it should be emphasized that Mr.
Timmons testified that neither the tubes that are candidates for expansion due to flow-induced vibration nor the tubes not requiring expansion are expected to experience 40% wear through the tube wall over the life of the plant. (pd. at 6198-99, 6202, 6265.)
On pages 19 and 20 (unnumbered) of its opinion, Intervenor questions the empirical basis for Dr. Rajan's conclusion that based upon the Staff's preliminary review of the proposed modifications, the objective of minimizing tube degradation associated with flow-induced vibration will be accomplished by those modifications. The record reveals that the NRC Staff has been aware of the research into the tube vibration matter for almost 18 months, and a number of Staff members have had over a dozen meetings with Westinghouse, at which there has been an exchange of information. Mr.
Rajan stated that the review of the modification to address the flow-induced vibration phenomenon has been one of the most detailed reviews he has observed in his nine or ten years _with the NRC. (Rajan, Tr. 4681-82.) In addition, the review by Argonne National Laboratory, retained by the NRC Staff as consultants in this matter, has been coextensive with the Staff review. The Staff has furnished Argonne
all pertinent Westinghouse information submitted to the Staff, and has been in communication with Argonne regarding the modification. Argonne advised the Staff that vibration levels that could be expected with expanded tubes and reduced flow through the main feed nozzle would result in tube wear that would not reach 40% over the 40-year life of the plant. A written report will be sent by Argonne to the NRC Staff upon completion of its review. (Rajan, Tr. 6328, 6330-37.)
Loose Parts Monitoring System Intervenor states in the Opinion on page 20 (un-numbered) that the " advanced equipment" necessary for visual inspections, i.e., fiber optics and/or radio camera devices and for acoustic monitoring may only be at the disposal of Westinghouse and not utilities, such as Edison. Intervenor has not supplied nor is Applicant able to find any support in the record for this assertion.
Intervenor also states at page 21 (unnumbered) of the Opinion, citing Finding 103, that written tool and material inventory control procedures are in draft form and may or may not be in place before fuel load. However, Mr.
Blomgren stated that the intention is that they be complete before fuel load. As maintenance is performed on the ,
steam generators over time, new procedures will be developed to encompass specific activities, which would be incorporated into the procedure at that time. (Blomgren,
Tr. 4257-58.) In other words, the procedures as they exist now are expected to be completed before fuel load. Any newly instituted procedures based on later-discovered exigencies would be added as they arise.
Preservice and Inservice Inspections At page 23 of the Opinion, following the statement that the maximum permissible leak rate corresponds to a .43-inch maximum allowable crack length, Intervenor avers, ,
without cit.ation, "[Y]et, the Ginna rupture was a single tube rupture at 700 gpm." Applicant cannot infer the basis on which Intervenor ties the Ginna event, which was caused by tube wear (Fletcher, Applicant Prepared Testimony at 14, ff. Tr. 5908), to the leak-be. fore-break principle, which is related to degradation in the form of stress-corrosion cracking. (Patel, Applicant Prepared Testimony at 12-13,
, ff. Tr. 4126.) A large leakage event resulting from stress-corrosion cracking has not occured for several years.
(Fletcher, Applicant Prepared Testimony at 11-13, ff. Tr.
5908.)
Water Chemistry On page 34 of the Opinion, Intervenor, citing Findings 164 and 166, refers to the ruptures at Point Beach and Surry. It should be noted that these events occurred in 1975 and 1976 respectively, and, as Mr. Fletcher testified, the implementation of AVT water chemistry, rigorous secondary
side water chemistry control, periodic inspections and design changes since 1976 are expected to prevent recurrences of these events. (Fletcher, Applicant Prepared Testimony at 11-13, ff. Tr. 5908.)
On page 36 of the Opinion, Intervenor, citing Finding 176, states that a complete sample of the chemical contained in the tube support plate crevice has never been taken, and there is no empirical evidence on the concentration of corrodents at the tube support plate crevice or the minimum concentration of corrodents necessary at the crevice for corrosion to occur or to rapidly propagate. However, Dr. Wootten testified that the geometry of the crevice is, by design, nominally 12 mills. When denting occurs, that crevice is packed with magnet'ite, an iron oxide with a very small porosity. Dr. Wootten stated that taking a solutica from that packed body in an operating steam generator would be very difficult, if not impossible. (Wootten, Tr. at 4177.) In any event, the rigorous AVT water chemistry control program that will be implemented at Byron will enhance the long-term integrity of the reactor coolant pressure boundary by minimizing the formation of corrosion products delivered to the steam generator. (Wootten, Applicant Prepared Testimony at 16, ff. Tr. 4126; Fletcher, Applicant Prepared Testimony at 7-9, ff. Tr. 5908.)
On page 37 of the Opinion, citing the recommen-dation of Mr. Bridenbaugh (Finding 179), Intervenor recommends
that Byron operating chemistry procedures be reviewed by an independent body to assure that proper procedures are in place and that the procedure is " identified, followed and complied with." Applicant submits that (1) there is neither an NRC requirement nor an evidentiary basis for requiring an independent review and (2) an independent review is not necessary. As noted by the Board, the record establishes that Mr. Bridenbaugh's recommendation is not made for technical reasons, but rather for incentive c arposes (Judge Smith, Tr.
6469.) Contrary to the implication in Intervenor's recommen-dation, Applicant possesses the expertise as well as a powerful incentive to ensure that the procedures are properly in place; if any shutdown is incurred resulting from a procedure that is not properly in place, Applicant will suffer significant economic loss, both in down time and replacement costs. Further, as Mr. Bridenbaugh himself testified, he has no reason to believe that Commonwealth Edison's procedures will not be completed as necessary and contain the necessary high level of water chemistry procedures implementation. (Bridenbaugh, Tr. 6465.) Accordingly, Intervenor's recommendation finds no support in the record and should be disregarded.
Tube Rupture Events There is similarly no support for Intervenor's statement on page 38 of the opinion, citing Mr. Bridenbaugh's
' testimony (Finding 181), that there is an increased prob-ability that accidents will be initiated by tube failures during normal operation and an increased likelihood that accidents not now considered in the safety analysis may occur as a result of the steam generator tube degradation after some period of operation. As Intervenor recognizes in Finding 182 (not cited in its Opinion section), on cross examination, Mr. Bridenbaugh admitted that he performed no calculations to support either assertion. (Bridenbaugh, Tr.
6475.) Mr. Bridenbaugh also admitted that he has performed no independent calculations or analyses to ascertain the probability or radiological consequences of multiple tube failures in concurrent tube rupture and LOCA events. (pl. ,
at 6476.)
On pages 38-39 of the Opinion, citing Finding 183, Intervenor notes that Mr. Fletcher's conclusion that steam generator tube degradation at Byron should not be a safety concern and that tube rupture should not occur, even under the conditions of MSLB's or LOCA's, is based in part on " limited operational experience of steam generators with AVT water chemistry." Applicant is at a loss to understand Intervenor's judgment that there has only been " limited" operational use with AVT water chemistry inasmuch as it.
acknowledges the industry conversion to AVT in 1974 (page 34 l
of the opinion). In addition, no citation is provided to
)
'the record to demonstrate that Mr. Fletcher's conclusions i
. _ _ , _ ~ _ . . - . _ .. -- -"
are indeed supported only by " limited" bases (other than to Mr. Fletcher's own testimony which clearly does not support Intervenor's assertions).
Intervenor also questions Mr. Fletcher ls above conclusions by stating that a Byron-specific probability risk analysis considering tube ruptures and LOCA's has not been done. Mr. Hitchler, Manager of Probabilistic Risk Assessment with the Nuclear Safety Department of Westinghouse, testified that Westinghouse has provided generic analyses to show the impact on peak clad temperature when multiple tube ruptures in a large LOCA event are postulated. (Hitchler, Tr. 6127.) He explained that Byron-specific analyses had not been performed because the evaluation that has been performed in the past has shown a small impact on the overall LOCA analyses for multiple tube rupture. That is, analyses have been performed indicating that failure of from five to eight tubes during a large break LOCA would not produce a significant change in the results. (pd. at 6127-28.) Mr.
Hitchler also testified that Byron-specific accident analyses have been performed that consider multiple-tube rupture accidents (id. at 6122-25) and a number of analyses have been submitted to the NRC as part of the Westinghouse Owners Group program for development of emtrgency procedures and guidelines with respect to tube ruptures during steam break and other combined events. (Id. at 6123.) These are generic guidelines that have been forwarded to the Byron
I operating staff for implementation according to the specific instruments at that plant. (pd. at 6124.)
-Overall, the record reveals firm support for Mr.
Hitchler's unrefuted quantitative assessment, discussed in ,
Applicant's Findings 223-227, that "the frequency of multiple tube ruptures combined or as a consequence of (i) large LOCA events is 5x10-7; (ii) small LOCA and transient events with normal pressure differentials is 2x10-5; and (iii) main steam /feedwater line break events is 3x10-5 . . . The frequencies of occurrence for these accidents are beyond the range of probabilities established generally for design basis accidents. . . .
Thus, as Applicant and the Staff aver, the frequency of postulated tube ruptures combined or as a consequence of transient conditions and accident conditions such as MSLB's and LOCA's is calculated to be extremely low over the 40-year life of the plant. (Applicant's proposed findings on steam generator tube integrity, pg. 21. See Hitchler, Applicant Prepared Testimony at 5-8, ff. Tr.
5908.)
Intervenor, on page 39 of its Opinion, citing Finding 187, states:
"The consequences of a large MSLB inside con-tainment could be adversely affected by such an event. Calculations have shown that provided there are no additional events such as operator error or safety valve failures, containment integrity should be effected [ sic], the core should remain covered and cooled due to the addition of emergency core cooling, and there should be ample water supply available for long term cooling.
However, if operator error and/or safety valve failure cEanges the accident sequence, the MSLB could be adverselv affected and subsecuent tube ruptures are possible. (Finding 187)" (Emphasis added.)
However, the underlined portions are not included in Inter-venor's Finding 187, are not included in the portion of the record cited in the Finding, nor in the prepared testimony of Mr.
Marsh (Marsh, NRC Staff Prepared Testimony at 4, ff. Tr.
4473.) Accordingly, these portions of the statements should be disregarded by the Board inasmuch as they represent an improper attempt to supplement the record. The same is true with respect to the following statement found on page 41 of the Opinion:
However, these calculations assume no operator error, inadvertent valve failures and/or rapid depressurization of the secondary side occur and do not lead to further tube ruptures and/or a MFLB.
That statement does not appear in the corresponding Findings (194 and 195) nor in the portions of the record cited by those findings. The same is true with respect to the statement on page 40 of the Opinion that the low probability assigned to MSLB, MFLB and LUCA accidents must be multiplied by the number of reactor years for the industry as a whole to determine an industry wide probability.
Similarly, Intervenor's statement on page 42 of
-7 the opinion that the 10 per year estimate for the frequency c f tube rupture events in combination with accidents that are predicted to result in severe core damage contains a
" great degree of uncertainty" is not supported in the record. Once again, the finding (number 198) on which the assertion is based contains no such statement. In fact, the record reveals that these figures have an 80% confidence level with a range of 2.5 and a 90% confidence level with a 4
range of 5. (Hitchler, Tr. 6235-36.)
i e
4
Unresolved Safety Issue A-3 Intervenor's statement with respect to the Un-resolved Safety Issue set forth on pages 46-47 of the opinion does not comport with the record. As stated in the NRC Staff's Proposed Steam Generator Findings at pages 25-26 of that Opinion:
The NRC Staff regards the Westinghouse steam generator tube integrity unresolved safety issue A-3 as effectively resolved through a combination of remedial measures, such as those identified above, operational precautions, such as routine in-service inspection and surveillance, and preventative tube plugging. This is borne out by the discernible downward trend in the amount of tube plugging necessitated at domestic nuclear plants over the past several years since the inception of a number of the described precautions.
The Staff anticipates formal documentation of the resolution report of USI-A-3 in approximately July 1983.
Intervenor's Conclusion In its conclusion on page 48 of the Opinion, Intervenor's suggest, citing no authority, that because Byron may be the first operational plant, it should be a licensing condition that Applicant instrument the Byron Steam generators with accelerometers, and develop a special inspection program. However, Mr. Timmons testified that Commanche Peak Unit 1, not Byron, would be the first D-4 plant in operation. (Timmons, Tr. 6058, 6065.) He also testified that Westinghouse intends to install accelerometers in the first plant, whichever that plant is. (Id.) Accordingly,
this issue is merely a matter of timing that need not be made a licensing condition. It is simply a procedural question that should be left to the appropriate administration of the NRC Staff. With respect to implementing'an inspec-tion program other than that to which Byron has committed pursuant to the regulatory guide, nothing in either the record or Intervenor's over 90 pages of opinion and findings supports such a requirement.
Also in the conclusion, on page 49 of the opinion, Intervenor suggests a finding that the Board has failed to satisfy General Design Criteria 30 and 31 Criterion 30--
Quality of reactor coolant pressure boundary--was not litigated as part of League Contention 22 and DAARE/ SAFE Contention 9(c) on steam generator tube integrity. (Applicant also notes that criterion 16, quoted by Intervenor, is outside the purview of the steam generator contention.)
With respect to Criterion 31, the record, as discussed in Applicant's proposed steam generator findings, amply demonstrates that Criterion 31 has been satisfied. Applicant submits that the remainder of findings stated in the conclusion to Intervenor's Opinion are not supported by the record and, accordingly, should not be adopted by the Board.
In sum, the findings and conclusions that Inter-venor proposes in support of its steam generator tube integrity contention that have been discussed above are not supported by substantial evidence in the record. Accordingly,
- d
~ . ?!
J Applicant respectfully requests that these proposed findings of fact and conclusions of law submitted by Intervenor not ,
be adopted by the Board. J
, Respectfully submitted, n>
g foseph Gallo l' f Michael R. Goldfein On Behalf of Commonwealth Edison Co.
I S H A M. , LINCOLN & BEALE 1120 Connecticut Avenue, N.W.
Suite 840 Washington, D.C. 20036 (202) 833-9730 i ISHAM, LINCOLN & BEALE Three First National Plaza Chicago, Illinois 60602 (312)S58-7500 n
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454 OL
) 50-455 OL (Byron Nuclear Power Station )
Units 1 & 2) )
CERTIFICATE OF SERVICE The undersigned, one of the attorneys for Common-wealth Edison Company, certifies that he filed the original and two copies of the attached " APPLICANT'S REPLY TO INTERVENOR ROCKFORD LEAGUE OF WOMEN VOTERS' PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW REGARDING STEAM GENERATOR TUBE INTEGRITY" with the Secretary of the Nuclear Regulatory Commission and served a copy of the same on each of the persons at the addresses shown on the attached service list. Service on the Secretary and all parties, unless otherwise indicated, was made by deposit in the U.S. Mail, first-class postage prepaid, this 26th day of July, 1983.
N .l y One of the attorneys for Commonwealth Edison Company Of Counscl Joseph Gallo Michael Goldfein ISHAM, LINCOLN & BEALE 1120 Connecticut Avenue, N.W.
Suite 840 Washington, D.C. 20036 Three First National Plaza Suite 5200 Chicago, Illinois 60602 Dated: July 26, 1983 1
SERVICE LIST COMMONWEALTH EDISON COMPANY -- Byron Station Docket Nos. 50-454 and 50-455 Mr. Ivan W. Smith Secretary Administrative Judge and Chairman Attn: Chief, Docketing and Atomic Safety and Licensing Service Section Board U.S. Nuclear Regulatory Commission Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Ms. Betty Johnson 1907 Stratford Lane Rockford, Illinois 61107 Dr. Richard F. Cole Atomic Safety and Licensing Ms. Diane Chavez Doard Panel SAFE U.S. Nuclear Regulatory Commission Washington, D.C. 20555 326 North Avon Street Rockford, Illinois 61103 Atomic Safety and Licensing Dr. Bruce von Zellen Board Panel U.S. Nuclear Regulatory Commission Department of Biological Sciences Washington, D.C. 20555 - Northern Illinois University.
DeKalb, Illinois 60115 Chief Hearing Counsel Office of the Executive Joseph Gallo, Esq.
Legal Director Isham, Lincoln & Beale U.S. Nuclear Regulatory Commission Suite 840 Washington, D.C. 20555 1120 Connecticut Ave., N.W.
Washington, D.C. 20036 Dr. A. Dixon Callihan Union Carbide Corporation Douglass W. Cassel, Jr.
P.O. Box Y Jane Whicher Oak Ridge, Tennessee BPI 37830 Suite 1300 Mr. Steven C. Goldberg 109 N. Dearborn Ms. Mitzi A. Young Chicago, Illinois 60602 Office of the Executive Legal Ms. Patricia Morrison Director U.S. Nuclear Regulatory Commission 5568 Thunderidge Drive Washington, D.C. Rockford, Illinois 61107 20555 Atomic Safety and Licensing Mr. David Thomas Appeal Board Panel 77 South Wacker U.S. Nuclear Regulatory Commission Chicago, Illinois 60621 Wachington, D.C. 20555
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