ML19259C466

Submits Draft of Proposed ECCS Findings Which Reflect Record as of Close of Feb 1979 Hearing.Certificate of Svc Encl

ML19259C466
Person / Time
Site:	Black Fox
Issue date:	05/09/1979
From:	Davis L NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Wolfe S Atomic Safety and Licensing Board Panel
References
NUDOCS 7906220130
Download: ML19259C466 (31)

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PUDLIC DOCUAIENT R003f i

Sheldon J. Wolfe, Esq.

Mr. Frederick J. Shon, Member Atomic Safety and Licensing Board Atomic Safety and Licensing Board V. S. fluclear Regulatory Commission U. S. fluclear Regulatory Commission Washington, D. C.

20555 Washington, D. C.

20555 Dr. Paul W. Purdom Director, Environmental Studies Group Drexel University 32nd and Chestnut Street 7

_ Philadelphia, Pennsylvania 19104 In the Matter of y 'y [ Ashe PUBLIC SERVICE COMPAfiY OF OKLAHOMA, h,

c-2 (G* 'G7pg g r-ASSOCIATED ELECTRIC COOPERATIVE, INC. Af1D WESTERfl FARMERS ELECTRIC COOPERATIVE, If1C.

cs (Black Fox Station, Units 1 and 2) c Gentlemen:

In motions dated April 16, 1979 and April 26, 1979, the f4RC Staff requested and received extensions of time to delay the filing of its Proposed Findings of Fact and Conclusions of Law on ECCS matters. As those pleadings indicate, the time was needed for the Staff to consult and conclude its on-going talks with General Electric about various aspects of the Staff's generic review of GE's Appendix K model.

Although we had completed our Proposed Findings based on the evidentiary record in the proceeding at the original due date, three matters re-lating to ECCS considerations had developed which made it unclear to Staff Counsel whether the information reflected in those findings remained accurate. The first matter relates to recent information the Staff had received concerning the application of the Leibniz Rule (for differentiation under the integral sign with moving boundaries) in the Safe and Reflood Codes in the GE ECCS evaluation model used for Black Fox.

Since it was unclear whether this new information affected our prior evaluation, the Staff began discussions of the matter with GE. While GE has provided information indicating its belief that this matter will have a small effect on calculated peak clad temperature, they are still in the process of providing information so that the Staff can determine whether or not the effect may be sufficient to constitute a "significant change" under Part II of Appendix K.

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ihe second matter concerns magnitude of the heat transfer coefficient calculation used to determine the pressure transient in the Safe Code.

The original GE ECCS model used h = 4 BTU /HR - FT2 _ of but the Staff found this magnitude to be unacceptable.

While the Reflood Code used to calculate part of the reflood transient does not use the h = 4 figure, General Electric continues to use the h = 4 figure in its Safe Code.

For this reason, the propriety of the use of h = 4 in the Safe analysis is currently being investigated by both GE and the Staff.

The third matter relates to our TLTA testimony.

As stated in the supplemental testimony of Mr. Wayne Hodges, while the Staff felt that sufficient conservatisms existed in the current evaluation model to offset the cumulative effects of TLTA phenomena, we felt that a quanti-fication of those effects should be provided to confirm the accuracy

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of these predictions.

We outlined generally in Mr. Hodgds testimony the nature of the work we felt would be desirable to provide needed confirmation. Hodges Testimony at 6 fol. Tr. 8033. Also in February the Staff sent GE a rather detailed outline amplifying the general guidelines set forth in the testimony.

However, it is our understanding that GE has not started on this work, but rather believes an alternative course of test work to be adequate.

Due to the constraints on Staff and GE manpower resulting from other activities, we have not yet been able to meet to resolve whether GE's alternative approach will satisfactorily provide the confirmation the Staff believes is necessary.

Because we cannot ascertain at this time whether in fact any modification of our prior testimony will be needed and because we are not sure exactly how long it will take to resolve these items, we are herewith filing a draft of our proposed ECCS findings which reflect the record as it stood at the close of the hearing in February of this year.1/

When we have resolved our present problems with GE, we will submit to the Board any necessary amendments of the enclosed draft proposed findings.

We anticipate resolution of the matter within the next month or two.

On a monthly basis, we will keep the Board informed of progress in this matter.

Sincerely, l

> v,.

L. Dow Davis Counsel for NRC Staff cc:

See Page 2.g} jj} _1/ Pagination has been changed so that to insert our draft ECCS Findings in our previous submission, one should remove pages 31 to 40 and insert the enclosed pages 17-40.

cc: Jeseph Gallo, Esq. Mrs. Ilene H. Younghein Michael I. Miller, Esq. Mrs. Carrie Dickerson i Mr. Clyde Wisner Andrew T. Dalton, Jr., Esq. Atomic Safety & Licensing Appeal Board Atomic Safety & Licensing Board Panel Docketing and Service Section i Lawrence Burrell Mr. Gerald F. Diddle Mr. Vaughn L. Conrad Joseph R. Farris, Esq. Alan P. Bielawski Mr. Maynard Human Mr. T. N. Ewing Dr. M. J. Robinson 2282 114 h

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) PUBLIC SERVICE COMPANY OF OKLAHOMA, Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC. ) STN 50-557 AND ) WESTERN FARMERS ELECTRIC COOPERATIVE, INC. (Black Fox Station, Units 1 and 2) ) CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S FINDINGS OF FACT ON ECCS i MATTERS", dated May 9, 1979, in the above-captioned proceeding, have F been served on the following, by deposit in the United States mail, first class, or, as indicated by an asterisk through deposit in the I Nuclear Regulatory Commission's internal mail system, this 9th day of l May, 1979:

• Sheldon J. Wolfe, Esq.

Michael I. Miller, Esq. Atomic Safety and Licensing Board Isham, Lincoln & Beale One 1st National Plaza U.S. Nuclear Regulatory Commission te 240 Washington, D. C. 20555 gg llinois 60606

• Mr. Frederick J. Shon, Member Mrs. Carrie Dickerson Atomic Mafety and Licensing Board U.S. Nuclear Regulatory Commission Citizens Action for Safe Energy, Inc.

x9 Washington, D. C. 20555 a mo lahoma 74107 Dr. Paul W. Purdom M e W sner [t; Director, Environmental Studies Group Drexel University 9g Public Affairs Officer 32nd and Chestnut Street Y{0 Philadelphia, Pennsylvania 19104 Joseph Gallo, Esq. Arlington, Texas 76011 Isham, Lincoln & Beale 1050 17th Street, N.W. Andrew T. Dalton, Jr., Esq. Washington, D. C. 20036 Attorney at Law 1437 South Main Street, Room 302 Mrs. Ilene H. Younghein Tulsa, Oklahoma 74119 3900 Cashion Place Oklahoma City, Oklahoma 73112 k 2282 115

,

• Atomic Safety and Licensing Mr. T. N. Ewing Appeal Board Acting Director U.S. Nuclear Regulatory Commission Black Fox Station Nuclear Project Washington, D. C.

20555 Public Service Company of Oklahoma P. O. Box 201

• Docketing and Service Section Tulsa, Oklahoma 74102 Office of the Secretary of the Commission Dr. M. J. Robinson

{ U.S. Nuclear Regulatory Commission Black & Veatch i Washington, D. C. 20555 P.O. Box 8405 l

• Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission i

Washington, D. C. 20555 Lawrence Burrell i Route 1, Box 197 Fairview, Oklahoma 73737 Mr. Gerald F. Diddle General Manager Associated Electric Cooperative, Inc. P.O. Box 754 Springfield, Missouri 65801 i Mr. Vaughn L. Conrad Public Service Company of Oklahoma P.O. Box 201 Tulsa, Oklahoma 74102 Joseph R. Farris, Esc. Robert Franden, Esq. 1 Green, Feldman, llall & Woodard i 816 Enterprise Building xX /{J e _ c 7 Tulsa, Oklahoma 74103 L. Dow Davis Alan P. Bielawski Counsel for NRC Staff Isham, Lincoln & Beale One First National Plaza Suite 4200 Chicago, Illinois 60603 Mr. Maynard Human General Manager Western Farmers Coop., Inc. P. O. Box 429 P Anadarko, Oklahoma 73005 22g2 llf

. 22. 2. Emergency Core Cooling System (ECCS) Intervenors' Contention 2 on ECCS, as originally phrased, was as follows: Intervenors contend that the Applicants have not adequately demonstrated that the Emergency Core Cooling System for Black Fox 1 and 2 meets the requirements of 10 C.F.R. Part 50, Appendix K. 23. Following a motion for summary disposition, on the basis of sworn ~ affidavits from the Staff and the Applicants that the ECCS calculations for Black Fox had been performed, that the Plant complied with Appen-dix K, that full-scale testing of the ECCS was not needed to meet the requirements of Appendix K,1/ and that no other matters of fact or arguments raised by the Intervenors in their affidavit opposing summary disposition revealed.a genuine issue of material fact to be litigated (September 8,1978, Summary Disposition Order at 8-15), the Licensing Board reduced the ECCS issue in controversy to the following questions: Board Question 2-1 concerned the separation of thermal and hydraulic effects in the core spray distribution, Board Question 2-2 concerned any possible new ECCS code errors, and Board Question 2-3 concerned TAP-16 and the core spray distribution tests. The following is a discussion of the evidence heard on these issues. J_/ Comments, Rulemaking Hearino on_ Acceptance Criteria for ECCS Systems for Light Water Reactors, CLI-73-39, 6 AEC 1085,1127 (December 1973). 2282 117

22. 2. Emergency Core Coolina System (ECCS) Intervenors' Contention 2 on ECCS, as originally phrased, was as follows: Intervenors contend that the Applicants have not adequately demonstrated that the Emergency Core Cooling System for Black Fox 1 and 2 meets the requirements of 10 C.F.R. Part 50, ? Appendix K. 23. Following a motion for summary disposition, on the basis of sworn affidavits from the Staff and the Applicants that the ECCS calculations ~ for Black Fox had been performed, that the Plant complied with Appen-dix K, that full-scale testing of the ECCS was not needed to meet the requirements of Appendix K,I/ and that no other matters of fact or arguments raised by the Intervenors in their affidavit opposing summary disposition revealed.a genuine issue of material fact to be litigated (September 8,1978, Summary Disposition Order at 8-15), the Licensing Board reduced the ECCS issue in controversy to the fol. lowing questions: Board Question 2-1 concerned the separation of thermal and hydraulic effects in tne core spray distribution, Board Question 2-2 concerned any possible new ECCS code errors, and Board Question 2-3 concerned TAP-16 and the core spray distribution tests. The following is a discussion of the evidence heard on these issues. If Comments, Rulemaking Hearing on Acceptance Criter_ia f_or ECCS _ Systems for Light Water Reactors, CLI-73-39, 6 AEC 1085,1127 (December 1973). 2282 118

24 a. Separability of Thermal and Hydraulic Effects in ECCS Core Spray Distribution Board Question 2-1 reads: Do doubts exist about the independence and separability of thermal and hydraulic effects in the specific calcu-lations used to demonstrate compliance of Black Fox Station with Appendix K? ~~ Since Board Question 2-1 and Board Question 2-3 ("What bearing has TAP A-16 upon the Black Fox ECCS evaluation?") are both concerned with core spray distribution tests and the separability of thermal and hydraulic effects, the two will be combined in the following discussion. 25. Concerns over the adequacy of ECCS core spray distribution arose in 1974 when data from a Swedish BWR with overhead vertical spray nozzles showed that the presence of steam and pressure had an unanticipated effect on the trajectory of water from the system's core spray system. These tests showed that contraction of the High Pressure Core Spray (HPCS) cone could occur in steam, Tr. 6347, 6372, thus having an uncalculated effect upon the cooling geometry of the ECCS system. Although other tests in air had shown the GESSAR-238 core spray calcu-lations to be acceptable, Applicants' vendor, General Electric, com-mitted itself to confirmatory testing to demonstrate the adequacy of the GESSAR (and thus the BFS) 238 reactor's core spray distribution system in steam. Frahm testimony at 2-2, 2-5 fol. Tr. 6484; Minor 2282 119 testimony at 2-3, 4 fol. Tr. 6598; Levine testimony at 1 to 2 fol. Tr. 6323. It was this proposed core spray testing that became the subject of Task Action Plan 16 which is entitled, " Steam Effects on Core Spray Distribution." 26. Since fomer full-scale distribution tests had been in air (Levine at 2 fol. Tr. 6323), the core of the controversy between the Applicants and Intervenors was whether confirmatory testing of the core spray distribu-tion in the GE BWRs would show the same results in a steam environment as it did in the former full-scale tests in air at atmospheric pressure. Intervenors maintained that it would be necessary to test the core spray system in a 360 full-scale steam model or an operating reactor in order to accurately gauge the effects of steam on spray distribution. Applicants, however, felt that full-scale testing was impracticable due to the absence of a readily available steam supply of sufficient pro-portions needed to simulate a full-scale core, and that the actual reactor conditions could be realistically modeled through a combination of air and steam tests in a less than 360 configuration. Thus it became apparent that since the testing was not to be done in a full 360 steam mode, the adequacy of General Electric and Applicants' plans to confirm the accuracy of present core spray distribution depended upon whether themal or steam effects from one part of the reactor were likely to influence spray distribution oh another part of the reactor. 2282 120

_ 19 _ n 27. Testimony at the hearings showed that two factors might affect the BFS high-pressure core spray distribution system and thus the minimum amount of coolant which would be available from the core spray cooling 7 system in the event of a loss-of-coolant accident (LOCA) at BFS. These were hydraulic effects (the interaction between water streams coming from two different core spray nozzles) and thermal effects (heat trans-fer between subsaturated ECCS water and the steam environment in the core). Tr. 6517. 28. General Electric's expert witness, Mr. ' Aaron Levine, maintained that since large water droplet sprays from the BFS high pressure core spray system were much less affected by the steam environment than the more usual small droplet sprays and since thermal exchanges between the spray and steam were all completed a very short distance from the spray nozzles (saturation was said to occur approximately five diameters from the nozzle), all thermal interaction between spray flows and steam would occur prior to this point and the spray flow would be controlled by only hydraulic parameters (interaction between spray flows from adjacent nozzles) whose effects are well known. Thereafter, it would be possible to model the hydraulic and steam portions of spray flow independently, thus obviating the need for full-scale 360 steam test-ing. Levine at 2 to 3 fol. Tr. 6323. The Staff role in this contro-versy was to exercise its supervisory function in reviewing the test methodology, assumptions, and results to ensure in its own mind that 2282 121 the proposed tests would satisfy the requirements of Task Actior. Plan 16. Frahm at 2-2 fol. Tr. 6484. 29. The tests contemplated by General Electric to show the acceptability of current core spray computations, peak clad temp.eratures and compliance with Appendix K to 10 C.F.R. Part 50 in the event of a LOCA were con-ceptually complex. First, the program proposed by GE called for tests of a single GE 238 reactor core spray nozzle (of same design as BFS) at reactor design conditions in a steam atmosphere. Tr. 6398.

Next, after cbserving the core spray distribution of one nozzle in steam, a model and simulator nozzle of the single nozzle in steam would be developed in air.

Tr. 6400. Following the modeling of a steam nozzle in air, full-scale 360 testing of the 70 or so HPCS simulated nozzles will be conducted in air. By the use of superposition calculations and interaction factors derived from comparison of the test results to the calculation, a final calculated steam distribution representative of the expected 360 fullscale results in steam will be obtained. Tr. 6402. The results of this test modeling and calculational methodology will then be compared with full-scale tests of six spray nozzles (a 30 sector of a GE 238 reactor) in steam to confirm that the spray distri-bution obtained by ' testing in steam and air are, in fact, representa-tive of the whole GE 238 core in steam. Levine testimony at 2 to 3 fol. Tr. 6323, Minor at 2-5 fol. Tr. 6598; Tr. 6403-6404. 2282 122 30. Testimony from the Intervenors generally contested the assumption made by General Electric that thermal effects will be separable from hydrau-lic effects and felt that the tests would not confirm the adequacy of present core spray distribution patterns. While the Intervenors pre-sented no specific facts, tests or calculations to show the inaccuracy of any of the information presented by the Applicants (presumably because their witness was not an expert in thermo-hydraulics (Tr. 6599-6601),6604) they did seek to impeach the Applicants' expert witness by claiming the non-separability of thennal and hydraulic effects (Minor testimony at 2-2 to 2 fol. Tr. 6598, Tr. 6388-89) and by generally trying to impeach GE with concerns the NRC Staff had expressed about some of the methods and assumptions to be used in the verification process. Tr. 6357-6361; 6439. However, none of their testimony revealed any specific facts to show that Appendix X was not complied with because of core spray distribution (Tr. 6602-6603) nor did this Licensing Board perceive any facts from its own review of the record which would show non-compliance with Appendix K or question the efficacy of GE's verification program tests. Accordingly, since the weight of the evidence showed that (1) BFS complies with Appendix K (2) that the proposed core spray testing may very well show the ade-quacy of present core spray distribution (Tr. 6397, 6448), (3) that verification tests are to be completed before the plant is operable (Tr. 6353) and (4) that there are no technological reasons why the design of the plant or the core spray cannot be modified if the 2282 123 verification program shows problems with core spray distribution (Tr. 6357; Levine at 4 fol. Tr. 6323), this Licensing Board believes that a reasonable assurance exists that the plant can be safely constructed and later operated without danger to the public because of core spray distribution in the ECCS system. 31. b. ECCS Code Errors As stated above, Intervenors' Original Contention 2 maintained that BFS would not comply with Appendix K to 10 C.F.R. Part 50. However, follow-ing the submission of affidavits by the NRC Staff and Applicants to the effect that the BFS would conply with the provisions of Appendix K, Part 50, and in response to Intervenors' assertion in their affidavit opposing summary disposition that numerous ECCS errors exist, the Board ordered testimony on Board Question 2-2, which reads as follows: What "recently discovered" errors may exist in GE ECCS evaluation codes? Are there any errors other than those set forth in the SER at pp. 6-10 of Appendix A? The recently discovered errors alluded to in the Board questions were those reported to the NRC by General Electric in 1977, whicn were corrected and later summarized in the BFS SER issued in June of 1977. 32. Testimony by the Applicants' witness, Mr. Aaron Levine of General Electric, showed that while from time to time new calculations or 2282 124 assessments of ECCS codes and calculations had been made, Aaron Levine at 1-3 fol. Tr. 6323, passim,-the cumulative effect of these changes was a decrease in peak cladding temperatures. Levine at 3. To aid in the improvement and upgrading of the GESSAR-238 ECCS code, the General Electric Company had put into effect a program to standardize ECCS inputs and verify that those figures are correct for the particular plant being analyzed. Levine at 1-2 fol. Tr. 6323. With the GE ECCS reverification program 80 percent complete, Mr. Levine testified that refinements and/or corrections to GE BWR-6 ECCS codes had resulted in a net reduction in peak clad temperatures. Tr. 6453. Mr. Levine said that presently GE knew of no other ECCS errors and had a high degree of confidence that any other changes resulting from the reverification program would not result in an increase in the peak cladding tempera-ture associated with the design basis accident. Levine at 3 fol. Tr. 6323. In general, Staff testimony on ECCS errors was in accord with the Applicants' testimony that all known ECCS errors were corrected. Shearon at 2-3 fol. Tr. 6484. However, testimony by Hodges at 1 fol. Tr. 8033 discussed the potential for code errors based upon recent TLTA data. As stated in Mr. Hodge's testimony, a comparison of TLTA data with ECCS code analyses may result in code changes due to modeling e rro rs. 2282 125 33. Additional testimony from the Staff showed that under the NRC's License Contractor Vendor Inspection Program, the NRC Staff had recently con-ducted a series of inspections of nuclear vendors (including the BFS vendor, GE) to verify that adequate quality assurance procedures were established and implemented in order to minimize the likelihood that ECCS safety analysis computer code errors would go undetected. Brickley at 2-13 fol. Tr. 6484. While the inspection in question was a rather general one which dealt primarily with the adequacy of procedures and their implementation, no specific errors in the code used for the GESSAR-238 analysis were identified during this inspection. Brickley at 2-14,15 fol. Tr. 6484; Tr. 6507. 34. While testimony by the Intervenors' witness, Mr. Minor, pointed out that there had been ECCS code errors in the past (i.e., those docu-mented in the SER) and that in their opinion there would be additional errors uncovered in the future, Minor at 2-6 to 2-7 fol. Tr. 6598, they were not able to furnish the Licensing Board with any specific examples not known to the NRC, the Licensing Board or the parties. Thus, while this Licensing Board has no doubt that refinements and even improve-ments in ECCS codes will occur in the future (as the Applicants' testi-mony revealed), we cannot, on the basis of the record before us, state that any significant additional ECCS errors are extant for BFS because: 2282 126

1. Applicants testified that at present they are unaware of any ECCS code errors, supra. 2. The Staff is unaware of any additional errors, other than those which may be disclosed by comparison with TLTA data. 3. The Staff has reviewed the Applicant's ECCS submission and found the BFS ECCS system to conform to Appendix K. 4. Intervenors' witness was not an expert in the area, Tr. 6599-6601, 6604, nor was able to identify any specific uncorrected ECCS errors. Tr. 6602-6603. 5. In any event, based on evidence of past ECCS code corrections, any code errors discovered in the future could be corrected for BFS by the operating license stage. See Levine at 1-3 fel. Tr. 6323. 35. c. Low Pressure Coolant Injection Diversion (LPCI System Diversion Pursuant to its duty to keep the Licensing Board informed of new information,l/ the Staff advised the Board and parties that a BWR-6 calculation recently performed by General Electric (Allen's Creek 1/ Virginia Electric Power Company (North Anna Power Station, Units 1 and 2), CLI-76-22, 4 NRC 480, 487-488, 491, n. 11 (1976). 2282 127 Amendment 47 dated August 25,1978) to investigate the effects of ECCS diversion to containment in the event of a small LOCA had the effect of raising the peak cladding temperature for the small break analysis and promised this Board and parties more infonnation on the matter as the Staff investigation progressed. Shearon/Frahm testimony at 2-5 fol. ? Tr. 6484; see Tr. 6324-6329. The matter of LPCI diversion was addressed in supplemental testimony by Ronald Frahm; the BFS ECCS reviewer. Frahm Supp. Testimony on Board Question 2-3 fol. Tr. 8033. The following are the results of the investigation on the effects of the Allen's Creek LPCI diversion. 36. Automatic diversion of the low pressure ECCS coolant injection system from the reactor vessel where it normally is used, to the containment has been provided in the GESSAR-238 reactor to insure that the reactor containment will maintain its integrity should there be a substantial steam flow bypass of the suppression pool which normally accommodates steam and pressure blowdowns. LPCI diversion will occur only if a high pressure signal ( 9 pounds per square inch) is present in the contain-ment 10 minutes af ter the occurrence of a LOCA. Frahm Supp. Test. at 1 fol. Tr 8033. 37. Because such a high containment pressure would not be present in contain-ment the first ten minutes after a small break LOCA and because a large 2282 128

break LOCA would have reflooding accomplished prior to the ten minute point by operation of the low pressure coolant injection system, the core will be adequately cooled at the time of diversion. The limiting break / single failure for LPCI diversion PCT would be a 0.02 square foot break (that break size which allows LPCI flow into vessel starting at the time of diversion) in the high pressure cora spray (HPCS) line, com-bined with an assumed low pressure core spray-law pressure coolant injection diesel generator failure. The analysis conservatively assumes that no flow enters the vessel through the broken HPCS line and that two low pressure coolant injection pumps are diverted to the containment spray mode at 10 minutes. Frahm at 2 fol. Tr. 8033. 38. Bounding LPCI diversion calculations by General Electric for BFS assumed the worst single failure of the low pressure core spray diesel generator combined with the additional failure of one automatic depressurization valve to open (not required by Apper. dix K). The calculation with these assumptions yielded a peak cladding temperature of 2085 F, 0.17 percent hydrogen generation of the metal in the cladding and total oxidation of less than 2 percent of the total cladding thickness before oxidation. Frahm Supp Test at 2 fol. Tr. 8033. Based on the testimony above, the Licensing Board finds that even if a small break and LPCI diversion were to occur at BFS, the plant would 2282 129

still be in conformance with the provisions of Appendix K to 10 C.F.R. Part 50 and meet the requirements of 10 C.F.R. 550.46. r 39. d. Two Loop Test Apparatus Results (TLTA) In its monthly report to the NRC submitted in October of 1978, General Electric informed the NRC of the existence of new test results from its Two Loop Test Apparatus as part of the Blowdown / Emergency Core Cooling-Program, a cooperative experimental research program jointly funded by the Electric Power Research Institute, General Electric and the NRC, which was being conducted in San Jose, California. A comparison of test results of an average power bundle with low pressure emergency core cooling injection flow and test results with the same initial conditions but no ECC injection showed that the system depressurized more slowly with ECC injection than without it. Hodges at 1-3 fol. Tr. 8033. Since the slower depressurization with ECC injection was contrary to intuitive expectations and had the possibility of adverse implications, discussions with General Electric took place on October 10, 1978 and the Staff informed the Black Fox Licensing Board and parties of the situation at the ongoing safety hearings on October 20, 1978 by presenting an October 17, 1978 memorandum for D. B. Vassallo, Acting Assistant Director for Light Water Reactors, DPM, from Frank Schroeder, Acting Assistant Director for Reactor Safety, DSS. Tr. 6305. The Staff promised to report back to the Board when more information became 2282 130

available on how the new TLTA data would affect the GESSAR-238 ECCS evaluation model which was the subject of Board Questions 2-2 and 2-3. Tr. 6305-6306. 40. Staff testimony on the matter of TLTA on BQ 2-1, 2-2 and 2-3 was pre-sented by Mr. Wayne Hodges, a Principal Reactor Engineer in the Reactor Analysis Branch of the Division of Systems Safety and consultant to the Program Management Group which manages the blowdown, heat transfer test program facility known as the TLTA. Mr. Hodges is a reviewer for ECCS code changes. Tr. 8035-36. 41. Testimony by Mr. Hodges revealed that while the TLTA test results indi-cated a need to investigate further a portion of General Electric's ECCS evaluation model, sufficient margin exists in the present ECCS calculations to assure that the GE evaluation model (and thus the BFS ECCS analysis) is appropriate and is in accord with the general requirements of Appendix K, even though changes to certain portions of the GE model may be necessary. Hodges at 1 fol. Tr. 8033. To resolve Staff concerns steming from TLTA test results, the Staff is requiring specific analyses from GE. The test data are to be analyzed by the test group at GE to verify the data and identify the various sources of steam in the test. Also, GE is required to perform calcula-tions with the ECCS evaluation model to test its essential features 2282 131

against the available experimental data for tests with and without ECC injection. If the results of these calculations suggest the need for model improvements, then these improvements will be required by the NRC but will be considered in conjunction with other model improvements requested by GE. Hodges at 1, 6 and 8 fol. Tr. 8033. 42. Information from the Blowdown / Emergency Core Cooling program showed that favorable as well as unfavorable data were obtained fran the test results. On the one hand, dryout of the core was delayed, higher heat transfer rates were indicated, and maximum cladding temperature was lower with ECC injection than without it. On the other hand, the test results showed that there was slower depressurization rate for the test with ECC injection than normally would be obtained from tests without the injection. Preliminary calculations performed by the Staff and GE suggest that the slower depressurization is due to larger vapor genera-tion with ECC injection than without it. Staff calculations indicate that the vapor generation exceeds that calculated in counter current flow limiting (CCFL) conditions under GE's staff approved SAFE and REFLOOD codes. Hodges at 1-4 fol. Tr. 8033. 43. After reviewing the data available from the TLTA test, the Staff con-cluded that the results of the present BFS ECCS analysis are conserva-tive because the original vapor generation and CCFL models used in the GE analysis were simplified and ignored physical phenomena such as heat 2282 132 transfer from the fuel rods to the channel box by thermal radiation, the condensation of vapor on the walls of the channel box and vapor generation from sources outside the fuel bundles. Thus was the judg-ment of the Staff that the inclusion of the new TLTA data with its higher heat transfer and the phenomena listed above would result in a PCT no higher than presently calculated and thus that the BFS would still comply with Appendix K to 10 C.F.R. Part 50 and require no hard-ware changes even though the GE ECCS model may ultimately require some correction. Hodges at 5-6 fol. Tr. 8033. 44. In regard to TLTA's effect on BQ 2-3, TAP-16 and the core spray distri-bution tests, Mr. Hodges testified that because the core spray tests will be conducted using a varying steam flow rate designed so that would exceed the upper bound of steam generation as revealed by the TLTA tests, the TLTA test data is expected to have no effect on the ability of the core spray distribution tests to separate thennal effects from hydraulic effects. Hodges at 7 fol. Tr. 8033. 45. Based on the unrebutted evidence described above, the Licensing Board finds that the new TLTA test data will have no effect upon the core spray distribution tests under TAP-16. Further continued use of the GE ECCS evaluation model is appropriate and is in accordance with the general requirements of Appendix K and 10 C.F.R. 550.46. 2282 133

- 4 6. 3. Containment Suppression Pool Loads (Contention Number 3) Contention 3 was combined with Contention 16 by agreement of the parties. A discussion of that contention is contained in Section D-16 of this opinion infra. 47. 4. Containment Buckling In response to a Board notification by the Staff regarding certain questions raised by a consultant to the Commission, the Board directed the Staff and/or Applicants to present evidence about containment buckling analyses in response to the following questions;

a. Has any Staff evaluation (other than the one paragraph enclosure to Mr. Paton's letter) of the significance of this report been made?

b. The report (at pp. 2 and 3) is severely critical of 2 out of 3 predictive methods specified through Reg Guide 1.57 and ASME code limits NE-3224.

Are the criticized methods to be relied upon in the design of BFS?

c. Has the buckling stress for BFS been determined by the method set forth in Section 5 pp. 4 and 5 of the report?

If so, how does it compare with values detennined by other methods? 48. Only Staff and Applicants presented written testimony addressing these questions. Staff witness Hafiz explained that the information submitted in the Board notification was preliminary in nature and need not be evaluated; that the buckling analysis methods criticized in the report will not be used in the design of the Black Fox Station; and that the Applicants have not yet presented the buckling calculation for the BFS 2282 134 design to the Staff, because that document is not required until 'she final design stage. Hafiz at 2,3 fol. Tr. 7970. The Applicants' testimony agreed with the Staff's testimony. Miller, Hagstrom, Guyot at 3, 4 fol. Tr. 7964. Therefore, the Board finds that the informa-tional documents submitted to the Board by the Staff concerning con-tainment buckling research do not affect the design of Black Fox Station. Although the Applicants submitted extensive information in their testimony concerning their proposed method of buckling analysis, this has not been presented to the Staff for review. Therefore, the Board finds submission of the analysis to be premature and makes no decision as to the adequacy of the method. 49. 5. Reactor Pressure Vessel Supports and Pedestal Contention 5, as originally submitted by Intervenors, read as follows: Intervenors contend that the Applicant has not adequately demonstrated that the reactor pressure vessel supports and the pedestal for Black Fox,1 and 2, can withstand the loads resulting from the design basis requirement of 10 CFR Part 50, Appendix A, Criterion 2, relating to earthquakes. Both Staff and Applicants moved for summary disposition of this conten-tion. Summary disposition was granted, in part, based on Staff affida-vits which showed that the Applicants have referenced the General Electric GESSAR 238 Standard Steam Supply System (Docket STN 50-550) which is designed so that the reactor pressure vessel support skirt 2282 135 will resist the effects of an earthquake up to 0.30g ground accelera-tion. Further, the pedestal which supports the reactor pressure vessel skirt is designed to 0.12. This is adequate since this Licensing 9 Board has already ruled that the design basis acceleration for Black Fox Station is 0.12g.O Consequently, the Board granted partial summary disposition on the matter and submitted the following question to be addressed in written testimony by the parties: 5-1. Is the treatment of vertical' motion in an earthquake of importance to the design of pressure vessel supports and pedestals and, if so, has it been accommodated? 50. Only Staff and Applicants submitted testimony responding to the ques-tion. Based on this testimony, the Board finds that the treatment of vertical motion in an earthquake is important and will be accommodated in the seismic mathematical analysis for both vertical and horizontal movements. The seismic loads derived from this motion will be incor-porated into the design according to the American Institute of Steel Construction specification and the American Concrete Institute Standard 318-71. Staff witnesses Polk and Kovacs at 2 fol. Tr. 8023. The General Electric Company will furnish the pressure vessel support skirt for the Black Fox Station using the GE design basis enveloping all expected soil and seismic conditions to be experienced at any site location. Applicants' witness Gang at 2 fol. Tr. 8010. The Applicants -1/ Public Service Company of Oklahoma (Black Fox Station, Units 1 and 2) LBP-78-26, 8 NRC 102,111 (1978). 2282 136

will provide the reactor pressure vessel pedestal and will provide interface loading data to GE to support the design verification of the reactor pressure vessel skirt. Applicants' witness Guyot at 20 fol. Tr. 7546. 51. 6. Tornado Missile Protection and TAP-32 Intervenors' original Contention 6 on tornadoes read as follows: Intervenors contend that the Applicant has not adequately Criterion 2j/ compliance with 10 C.F.R. Part 50, Appendix A, demonstrate for Black Fox,1 and 2, with respect to tornadic phenomena related to: (a) Missile penetration of the containment; (b) Rapid exterior atmospheric pressure transients or excursions on the containment; (c) Protection of new fuel; and (d) Protection of the spent-fuel storage facilities. 52. Upon application by the Staff and Applicants for summary disposition on Contention 6 and a finding by the Licensing Board that the Plant would _1f Criterion 2 of Appendix A to 10 C.F.R. Part 50 provides: Criterion 2 - Design bases for protection against natural phenomena. Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perfonn their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate con-sideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be perfonned. 2282 137

withstand the impact of an auto traveling 100 feet per second; that Tornado Reg. Guide 1.117, Rev.1, will be applicable to BFS; that there is no evidence that the Plant does not comply with the most current standard review plan; that combined tornado and LOCA loads need not be considered; that the BFS fuel pool is designed against both earthquake and tornado loads (but need not be built to withstand both forces combined), and that no other triable issues of fact were raised by the Intervenors or known to the Board, we granted the motion in part and denied it in part, leaving the following Board questions extant: 6-1. What relevance do Task Action Plans TAP-32 and TAP-38 have to BFS and, if they have relevance, what is their status? 6-2. What connection, if any, is implied between the UHS cooling tower discharge nozzles and the off-gas systems' potential for radioactive release by the statement at pp. 1.9-22 and 1.9-23 of the PSAR? September 8,1978 Licensing Board Summary Disposition Order at 19-24.

53. The issue of tornado protection of the UHS cooling tower discharge nozzle is treated in Chapter 11 of this opinion, leaving the relevance of TAP-32 and TAP-38 under the River Bend decision / the only issue 1

remaining to be resolved in the area of tornado missiles. If Other TAP testimony of uncontroverted issues are described in Chapter 9 of this opinion. Those TAP issues which relate to controverted issues are discussed in both the TAP testimony connected to Chapter 9 and also in the writeups relating to those contention numbers. 2282 138

54. TAP-32 and TAP-38 Testimony on the details of TAP-32 and 38, which deal with the general areas of all " missile effects" and " tornado missiles" respectively, was presented by both the Intervenors and the Staff. Testimony as to Task Action Plan 32, which deals with the effects of all missiles, both tornado and turbine, on safety related structures and equipment was provided by Staff witnesses, Dr. Kazimeras Camp 3 and Mr. Harold Polk (experts on missiles and the effects therefrom) fuliowing Tr. 6246. 55. Currently, it is thought that the large conservatisms used to compensate for the uncertainties in available empirical data and state-of-the-art analytical methods make evaluations of the risk from missile effects ul tra-conserva tive. TAP-32 consists of a description of analytical methods which will be used to make a quantitative measure of those conservatisms used in the current analysis and evaluations. Pol k - Campe at 6/19-2 following Tr. 6246. Since the program is expected to be completed in 1980 and will be a confinnatory quantification of those conservatisms thought to be already present in risk calculations (and thus could lead to less stringent design criteria in the future) the results of this study were expected to have little effect on Black Fox. Tr. 6247-6260. Testimony by the Staff Task Action Plan witnesses was to the same effect. Aycock, Crocker, Thomas testimony at 32-34 follow-ing Tr. 8309. 2282 139

No witnesses on the subject were presented by the Applicants. While Intervenors' testimony generally discussed the applicability of certain Reg. Guides and stated that the results of TAP-32 could affect Black Fox, they gave no specific details based on knowledge of the BFS design. Hubbard at 66-64 following Tr. 6294. For that reason and in view of the uncontroverted Staff testimony above, the Board finds no rational basis for belief that the results of TAP-32 will affect Black Fox and finds that the current conservative design criteria give a reasonable assurance that Black Fox can be constructed and operated without any undue risk to the public health and safety pending completion of the Task Action Plan. 56. 7. Load Combination Methodology After Board notification by the Staff that load combination methodology was undergoing Staff review and that the results of the review would be reported to the Board, the Staff submitted NUREG-0484 entitled " Method-ology for Confi rming Dynamic Responses." The report indicated that the Staff had approved use of SRSS (Square Root of the Sum of the Squares) technique for combination of SSE and LOCA dynamic responses within the reactor coolant pressure boundary and its supports. The report also indicated that 4 :rther study of possible greater use of SRSS was being conducted and that a supplement to the NUREG document would be published 2282 140

.. later. The report indicated that the Staff did not approve the use of the SRSS method of combining loads other than that previously approved.1/ Staff testimony infomed the Board that the Applicants tcd issue with the position of the Staff described above, and instead had proposed to use "Newmark-Kennedy Criteria." Staff written testimony fol. Tr. 8221. The issue before the Board was whether the Applicants should be required to use the absolute sum method for load combinations except for the limited use of SRSS methodology approved by the Staff or whether the Applicants should be pennitted to use a newly developed system called "Newmark-Kennedy Criteria." 57. The Applicants submitted extensive testimony concerning use of SRSS and the Newmark-Kennedy Criteria. One of their witnesses was Dr. Robert P. Kennedy, co-author of the method in question. Dr. Kennedy explained thet his criteria is still undergoing study for Black Fox Station, that a modification of the criteria had been made specifically for Black Fox Station and that he had not yet determined whether the modification should be applied generically. Tr. 8118. The Applicants testified that the creation of criteria justifying the generic use of SRSS in combining loads for reactor design by Drs. Newmark and Kennedy was a recent development. Applicants' witness Conrad at 2, fol. Tr. 8113. 1/ The Staff has previously approved uses of SRSS for combining loads from modal responses to earthquakes and for LOCA + SSE loads for PWR fuel bundle analysis. NUREG-0484 at 2. 2282 141

_ 40 _ Staff witnesses testified that expanded use of SRSS, including the Newmark-Kennedy criteria, is presently under review by the Staff. Varga at 3 fol. Tr. 8221; Dr. Mattu, Tr. 8237; 8239; 8271. The Ir.tervenors' testimony stated that SRSS methodology could result in reduction of loads of 50% or nore. Bridenbaugh at 11 fol. Tr. 7709. 58. In view of the evidence presented, i.e., that the Newmark-Kennedy Criteria is newly fonned and still the subject of modification and study by the authors, and because the conservatisms of the method are uncertain and still under review, the Board finds that the construction pennit should be conditioned so as to require the Applicants to use only the load combination methodology approved by the Staff. 59. 8. Fire Protection Intervenors submitted three contentions (contentions 7, 8 and 9) con-cerning fire protection as follows: Intervenors contend that in order for the Applicant to meet 10 CFR Part 50, Appendix A, Criterion 3, Black Fox 1 and 2 must utilize cables with fire retardant insulation. Intervenors contend that in order to meet 10 CFR Part 50, Appendix A, Criterion 3 the Applicant must separate the cable trays including those in the cable spreading rooms so as to prevent a recurrence at Black Fox 1 and 2 of the type of fire which took place in the cable spreading room at Browns Ferry. Intervenors contend that the Applicant has not designed an in-depth fire protection system for 2282 142}}