NUREG-0124, Forwards Addl NRC Testimony by KM Campe,A Hafiz,Sa Varga & Others to Be Presented at Hearings.Nrc Exhibit,Gessar SER, Suppl 1 (NUREG-0124) & Draft Hearing Schedule Encl
| ML19263C441 | |
| Person / Time | |
|---|---|
| Site: | Black Fox |
| Issue date: | 02/05/1979 |
| From: | Chon Davis NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Purdom P, Shon F, Wolfe S Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 7902220253 | |
| Download: ML19263C441 (46) | |
Text
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Fr!m r v r, 1979 sac ru alc Decu M Sheldon J. Wolfe, Esq.
Mr. Foderick J. Shon,flember Atomic Screty and Licensing Board Atomic Safety and Licensino Board U. S. ?!uclear Regulatory Comission U.S. fluclear Rr!!ulatory Comission Washincten, D. C.
20555 Washincte.n. D. C.
20555 Dr Pcul Purdom Directnr, Environmental Studies Group s
Drexti liniver,ity
+'Nh 32nd and Chestnut Street Philadelprnc. Pennsylvenia 19100 NT
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In the flatter of 9
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PUBLIC SERVICE C0'<PANY OF OKLAHO'iA, b.'
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ASSOCI ATED ELECTRIC COOPERATIVE, It'C. Af!D
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WESTERfl FARMERS ELECTRIC COOPERATIVE, If;C. Q:
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(Black Fox Station, Units 1 and 2)
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Docket fios. STN 50-556 and STil-50-557 Gentlemen:
Enclosed is additional Staff testimony, as follows, to be presented at the upcoming hearings:
a.
Affidavit of K.M. Campe on Ultimate Heat Sink Cooling Tower Tornado Missile Protection b.
Testimony of Abdel Hafiz on Buckling Supplemental Testimony of Steven A. Varga on Load Combination c.
Methodology d.
Additional Supplemental Testimony for Radilogical Contention fiumber 16 by Mel Fields, John Kudrick, and Cecil Thomas e.
Testimony of Ronald Frahm on LPCI f.
Testimony of Wayne Hodges on TLTA On Friday, October 13, 1978 the Staff presented the testimony of Bernard Turovlin with respect to Board question 15-1 (inter-granular stress corrosion cracking - Tr. 5156).
Mr. Turovlin testified about certain events in Germany, Japan and at the Duane Arnold facility.
He stated that the Staff had not completed its review of all these events and that, therefore, he could not say whether subsequent study would have any impact on the conclusions he had previously reached with respect to inter-granular stress corrosion cracking in the Black Fox case.
THis DOCUMEi!T CONTAWS 7 90 222 0 A53 POOR QUAUTY PAGES
V,~
4 For some time the Staff has been involved in a study of the referenced evcnts in German',, depan and at the Duane Arnold facility.
The Staff expects that stude to be available t ithin the next few weeks.
When it becca;es availablt, the Staff will forc:ard conies to this board and to all parties and will as' this I;oard to recoive it in o"ihnco at an appropriate tine-Included for co:.uiir nce of the Board i nd r rties are copies of Staf f's exhibi t, GESSAR lei;, Supple:..;n t !,0.
l (liUREn-0124 ).
Also enclosed is a copy of the draf t hearing schedule agrecd to by the parties.
Sincerely,
,O s
(-#1 [e df0A s.
w L. Dow Davis Counsel for NRC Staff Enclosures as stated cc (w/ enclosures):
Joseph Gallo, Esq.
Mrs. Ilene H. Younghein Michael I. Miller, Esq.
Mrs. Carrie Dickerson Mr. Clyde Wisner Andrew T. Dalton, Jr., Esq.
Paul M. Murphy Atomic Safety and Licensing Appeal Board Docketing and Service Section Lawrence Burrell Mr. Gerald F. Diddle Mr. Vaughn L. Conrad J oseph R. Farris, Esq.
Alan P. Bielaw;ki Atomic Safety and Licensing Board Panel Mr. Maynard Huma1 Mr. T. N. Ewing Dr. M. J. Robinson
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UNITED STATES OF AMERICA
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NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
PUBLIC SERVIC.E COMPANY OF
)
Docket Nos. STN 50-556 OKLAH0MA, ASSOCIATED
)
STN 50-557 ELECTRIC COOPERATIVE, INC.,
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AND WESTERN FARMERS ELECTRIC
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COOPERATIVE
)
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(Black Fox Station, Units 1 and 2)
)
SUPPLEMENTAL TESTIMONY OF RONALD K. FRAHM ON BOARD QUESTION 2-3 I am the sponsor of the a+;av>ed document entitled " Supplemental Testimony of Ronald Frahm on Board Question 2-3." and believe it io' be accurate to the best of my knowledge and belief. This attachment is an additional response to the question 2-3 noted in the " Order Ruling on Motions for Summary Disposition," by the Atomic Safety and Licensing Board, dated September 8, 1978.
Ronald K. Frahm 9
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Supplemental Testimony of Ronald Frahm on Board Question 2-3
'The Supplemental Testimony of Brian Sheron and Ronald Frahm, in response to Board Question 2-3, iridicated that calculations submitted on the Allens,
Creek docket show a peak cladding temperature.(PCT) higher than those submitted on the GESSAR docket referenced by Black Fox Station (BFS).
We further stated that we would update our review after requesting a.ddi-tional information from the applicant. The applicant responded to staff inquiries by letter dated November 7,1978 and December 18, 1978.
Automatic diversion of low pressure coolant injection flow to containment spray has been provided in response to a staff requirement.to assure containment integrity for potential steam flow bypass.of the suppression pool. The applicant has stated that this flow diversion will occur only if a high containment pressure (>9 pounds per square inch gauge) signal is present after 10 minutes. General Electric has indicated that suffi-cient steam bypass to cause such a containment pressure will not occur after any small break for which the core may not be reflooded prior to 10 minutes.
For larger breaks, the core will be reflooded prior to the 10-minute low pressure coolant injection diversion, thereby assuriog adequate core cooling at the time of diversion.
The applicant performed single failure, break size, and break location sensitivity studies to determine the worst break size, location, and
. single failure -ombination.
The worst break size (0.02' square foot) is that break L.ze which allows low pressure injection tiow into the vessel star, ting at the time of diversion. Small Ar breaks have a lower peak cladding, temperature (PCT) because:
(1) the core is uncovered for a shorter ' time period since less mass is lest through the break during the blowdown until low pressure coolant injection operates; and (2) the decay heat is lower at the time of core uncovery.
Larger breaks will obtain some benefit from the low pressure coolant injection flow to the vessel before diversion, thereby yielding lower peak cladding temperature than the worst 0.02 square foot break.
G G
f The sensitivity studies showed that the worst case break location and single failure to be the high pressure core spray (HPCS) line break with an assumed low pressure core spray-low pressure coolant injection s
diesel generator (LPCS D/G) failure. The conservative assumption is made that no flow enters the vessel through the broken high pressure
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core spray line, and that the two low pressure coolant injection pumps are div~erted to the containment spray mode at 10 minutes, even though no diversion would occur since the containment (wetwell) pressure is expected to be.below 9 pounds per square inch gauge.
The calculation submitted assumed the worst single failure (LPCS D/G) combined with additional failure of one automatic depressurization system (ADS) v?.lve to open (not r.equired by Appendix K), yielding a peak cladding temperature of 2085 degrees Fahrenheit. This value meets the limit of 2200 degrees Fahrenheit as specified by Section 50.46 of-10 CFR 50.
The maximum amount of hydrogen generated is calculated to be 0.17 percent of the total metal in the cladding, which meets the allowable limit of 1 I
percent.
The total oxidation calculateo is less'than 2 percent of the total cladding thickness, before oxidation, which meets the allowable oxidat. ion limit of 17 percent.
The staff concludes that the analysis is acceptable for use in the evaluation of emergency core cooling system performance to show that the requirements of Section 50.46 of 10 CFR 50 are met for the Black Fox Station construction permit.
The design details of the automatic low pressure coolant injection diversion system will be reviewed during the Final Safety Analysis Report (FSAR). ' The applicant has confirmed
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that no operator action is required prior to 20 minutes following a loss-of-coolant accident.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
PUBLIC SERVICE COMPANY OF OKLAHOMA
)
ASSOCIATED ELECTRIC COOPERATIVE, INC.,
)
Docket Nos.
STN 50-556 AND WESTERN FARMERS ELECTRIC
)
STN 50-557 COOPERATIVE, INC.
)
)
(Black Fox Station, Units 1 and 2)
)
Testinony of Wayne Hodges on Two Loop Test Apparatus (TLTA) Board Notification and Its Effect on Board Questions 2-2 and 2-3
Testimony of Wayne Hodges INTRODUCTION This supplemental testimony is a follow-up to the staff's Loard notification of the existence of new test results obtained from General Electric's Two Loop Test Apparatus tests (TLTA) as part of the Blowdown / Emergency Core Cooling Program being conducted in San Jose, California.
The test results were provided to the NRC by inclusion in the September 1978 monthly report issued by General Electric. Discussions with General Electric took place on October 10, 1978, and the staff's preliminary view of this information was presented to the Licensing Board and parties at the Black Fox Station Safety Hearings on October 20, 1978 in the form of the October 17, 1978 memorandum for D.B. Vassallo, Assistant Director for Light Water Reactors, DPM, from Frank Schroeder, Acting Assistant Director fc* Reactor Safety, DSS, TR 6305.
The first purpose of this testimony is to present the staff's view that the preliminary analysis of the TLTA test results indicate a need to investigate further a portion of General Electric's ECCS evaluation model.
If further analysis indicates that a part of the model is inadequate, GE will be requested to revise the model in a timely manner.
However, the staff believes that sufficient margin exists iri the present Black Fox ECCS calcu-lations to assure that the Black Fox design is sufficiently conservative as proposed so that no hardware changes are required.
It is the staff judgement that the continued use of the GE ECCS evaluation model is appropriate and is in accordance with the general requirements of Appendix K.
The staff believes, however, that following completion of the TLTA test results review, changes to certain portions of the GE model may be necessary.
The second purpose of this testimony is to explain the bearing of these data on the Core Spray Distribution tests, under TAP-16, mentioned in connection with Board Question 2-'a.
The Blowdown / Emergency Core Cooling (BD/ECC) program is a cooperative experimental research program jointly funded by the Electric Power Research Institute (EPRI), General Electric (GE) and the Nuclear Regulatory Commission (NRC). Tests are conducted by GE under this program in the Two Loop Test Apparatus (TLTA) in San Jose, California.
The purposes of the program are:
1.
Obtain and evaluate basic BD/ECC data from test system configurations which have calculated performance characteristics similar to i BWR with 8 x 8 fuel bundles during a hypothetical LOCA.
2.
Determine the degree to which models for BWR system and fuel bundles describe the observed phenomena, and as necessary, develop improved models which are generally useful in improved LOCA analysis methods.
The TLTA configuration used for the BD/ECC is scaled to a BWR/6 design (624 bundles) and includes the following major components:
(1) pressure vessel and internals, (2) an 8 x 8 electrically heated bundle, (3) two recirculation loops, (4)
ECC systems (HPCS, LPCS, LPCI), (5) Automatic Depressurization System, and (6) Auxiliary Systems.
During August of 1978, test number 6405 was conducted; the test had an aver:ge power bundle (5.05MW) with low ECC injection flow. Results of the test were compared with those from test 6007 which had the same initial conditions, but no ECC injection. The comparison was presented
. in the monthly report issued in September,1978 and in a program manage-ment group meeting on September 21, and 22, 1978.
The comparison showed that the system depressurized more sli ily with ECC injection than without ECC injection.
Since the slower depressurization with ECC injection was contrary to intuitive expectations, GE was requested to discuss the test results and implications with the NRC. This discussion took place on October 10, 1978.
Subsequently, tests 6406 and 6414 with average and high buridle power respectively and with average and low ECCS flow respectively, havc also been run. These tests are consistent with the earlier test.
The overall results of the tests, when compared with the results of tests conducted without ECC injection, clearly indicated the benefits of ECC injection.
Dryout was definitely delayed for most bundle elevations and maximum cladding temperature was lower with ECC injection than without injection. The tests also indicated higher heat transfer rates than those used in licensing calculations.
The test result which led to staff concerns about the conservatism of the GE ECCS evaluation model is the slower depressurization rate for the test with ECC injection. Since no pre-test calculations were performed by GE, as part of the program to predict test behavior, a direct comparison between the test results and those that would be predicted by the GE evaluation model is not yet available.
Post-test calculations performed for NRC with RELAP-4 at INEL (1) do predict a slower depressurization rate with ECCS injection but not to the extent observed in the test. The RELAP-4 calculation shows more ECCS flow penetration into the core than is shown by the data; consequently, the calculation gives significantly lower cladding temperatures than were
_4 measured in the tests. The increased ECCS flow penetration in the RELAP-4 calculation could be partially attributable to the slip flow model used for the calculation and ir not necessarily consistent with results that would be obtained from calculations with the GE evaluation model.
Preliminary calculations performed by the staff and similar calculations performed by GE suggest that the slower depressurization is due to larger vapor generation with ECC injection than without ECC injection. The staff calculation indicates that the vapor generation significantly exceeds that predicted from either the GE proprietary vaporization correlation used to calculate counter current flow limiting (CCFL) conditions by the REFLOOD code or by the average core heat transfer model in SAFE.
SAFE and REFLOOD are part of the approved ECCS evaluation model.
The energy required to increase the rate of vapor generation could come from the heated core, from the stored energy in the structural elements of the lower plenum or from other structural parts of the TLTA. The infor-mation presently available to distinguish between these sources of vapor generation is not conclusive. The observed water accumulation in the core, however, represents a potential source for higher core steam generation.
If the extra steam is generated in the core, then the vaporization correlation and the SAFE heat transfer model predict too little vapor generation which could mean that nonconservative assumptions have been made about actuation of the LPCI and CCFL breakdown in LOCA analyses. On the other hand, TLTA has a larger surface to volume ratio than a BWR and this could lead to a greater steam generation in TLTA than if the surface to volume ratio were the same.
If a large portion of the extra steam is generated in the structure,
, then the test results are atypical of a BWR and do not have unfavorable implications on the currently approved licensing models REFLOOD and SAFE.
The NRC will require that these factors be considered in the evaluation of the test data.
Although TLTA tests clearly show that the core spray has the beneficial effect of reducing cladding temperature, the test results have both favorable and unfavorable interpretations with respect to the conservatism of the evaluation model.
Favorable results include higher heat transfer than used in licensing calculations for high power bundles (even without ECC injection).
Unfavorable results include the possibility of greater steam generation within the bundle than would be predicted by correlations in either SAFE or REFLOOD due to increased heat transfer in the average power bundles. This could result in delayed initiation of LPCI, LPCS, and result in later reflood.
The vapor generation and counter current flow limiting (CCFL) models in the GE evaluation model are very simple.
Because of their simplicity, several physical phenomena such as heat transfer from the fuel rods to the channel box by thermal radiation, condensation of vapor on the walls of the channel box and vapor generation from sources outside the fuel bundles are ignored in the present model.
It is our judgement that the inclusion in the evaluation model of the cumulative effects of these phenomena, coupled with the higher heat transfer observed in TLTA, would result in a peak cladding temperature no higher than presently calculated.
The staff believes that sufficient margin exists in the approved GE evaluation model to assure a conservative evaluation of the ECCS design performance even if the model
, does fail to predict accurately the vapor generation in the core.
- Thus, the Black Fax ECCS design is sufficiently conservative as proposed to assure that no hardware changes are required in the proposed Black Fox ECCS system.
However, we will require that staff cncerns about the vapor generation and CCFL models be resolved in a timely manner (well ahead of the Black Fox operation application).
Toward this end, the following steps are being taken:
1.
The test data are to be analyzed by the test group at GE to verify the data and to identify the various sources of steam in the test.
This effort is now in progress and partial results have already been submitted to the NRC.
2.
GE is required to perform calculations with the ECCS evaluation model to test its essential features against the available experimental data for tests with and without ECC injection.
This will check the capability of the evaluation model to predict the observed phenomenon.
Initial results from these calculations are scheduled for June,1979.
If the results of these calculations suggest the need for model improvements, then these impravements will be required by the NRC but will be considered in conjunction with other model improvements requested by GE.
Documentation providing GE's arguments + hat the ECCS evaluation model remains conservative is given in the January 30, 1979 letter and attachments to Robert L. Tedesco from E.P. Stroupe (letter no. MFN-022-79). Some of the information in this letter, and previously discussed in meetings with GE,
, has been used in the justification of continued use of the GE evaluation model; however, this does not constitute acceptance of all of the GE arguments in the letter.
In regard to the TLTA relevance to core spray distribution (Board Question 2-3), the impact on the spray tests is ninimal because the present strategy is to vary steam flow rate up to a rate sufficient to cause flooding at the top of the bundle.
The Lynn facility is designed so that more steam than is expected for a scaled reactor, even considering the results from TLTA, will be injected for some tests.
The question of separability of thermal and hydraulic effects on spray distribution is not affected by concerns on the rate of vapor generation within the bundl,
The separability question deals with the heat transfer to spray water as it emerges from the nozzles above the periphery of the core.
Regardless of whether or not the bundle vapor generation model is correct, sufficient steam is available to heat up the water droplets near the spray nozzles as claimed by GE.
In regard to the TLTA relevance to model errors (Board Question 2-2), TLTA indicates four potential model deficiencies; namely, the water level calcu-lation in REFLOOD, the vaporization correlation in REFLOOD, the water level calculation in SAFE and the heat transfer calculation in SAFE. These potential deficiencies will have to be corrected in a timely manner if they prove to be non-conservative.
Counter-current-flow-limiting (CCFL) effects to delay core spray flow penetration into the lower plenum were not discussed during the 1972 ECCS hearings and
Appendix K to 10 CFR 50 has no specific requirements with respect to CCFL.
The GE model was reviewed in 1974 under the general requirement of para-Jraph I.D.6. that "Following the blowdown period, convective heat transfer shall be calculated using coefficients based on appropriate exoerimental data...".
The recert TLTA tests indicate phenomena; improved heat transfer rate, water retention in the bundle, and increased steaming rate which have the potential to benefit the heat transfer parts of the model as well as adversely effecting the reflood/ refill portion of the model.
Both of these effects derive from the same aspects of the test; that is, a heat transfer rate for spray cooling greater than that used in the calculations.
It is the staff judgement that our present evaluation of the TLTA test results discussed above which involves the complexity of the CCFL phenomenon and the observed test results, supports the conclusion that the continued use of the GE ECCS evaluation model is appropriate and is in accordance with the general requirements of Appendix K.
The staff believes however, that following completion of the TLTA test results review changes to certain portions of the GE model may be necessary.
_g-REFERENCE 1.
Letter to R.E. Tiller from J.A.
Dearien,
" Test 6405 Preliminary Calculation and Data Corporation (A6039) - JAD-243-78,"
November 1,1978.
PF,0 POSED hEA!!IiiG SCliEDULE OF WITNESSES FOR BLACK FOX RADIOLOGICAL tlEALTH
,dlD SAFETY liEARINGS (February 19 - March 3,1979 M r.d /
Tuasday Wednesday Thursday Friday SaturJaz
- ./ l -
2/20 2/21 2/22 2/23 2/24 0:
Cont:im.ent Loads Containment IGSCC (Loadcombination Load Con. Lit.a t io Cu na in.
- 4. : L u.:s (CTD)
Buckling & BQ.5-1 methodology Method (CTD Staff:
Turovlin (exot n..
S ta f '.
Kadrick Aap:
Gang Shao (Cont. 3, 16, App. rebutul
- c ;.
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F1 eld:
Guyot BQ 5-1) on fire protect T uot;.a s Staff Testimony (Cont. 7, 8, 9)
APP K8""*dY Sta f f:
Polk on LPCI, TLTA
- c. ;
.[
hcs Staff:
Polk App.:
Conrud Kovacs Connely in r an Int:
Bridenbaugh Engaan P1 :s I-l
-1 Int.:
Bridenbaugh Int.: Fire S u r ret,u t t Staff Testimony
. _ fi-l on LPCI, TLTA P
A;.p. c: r...
.an; Containnent lead:
Turbine missiles Staff:
Frahm Load con.bination App.testin.unj c L,:.n(Ciu, (CTD)
(BQ 19-1 to 19-3)
Hodges Method (CTD) special QA processes Int.. Bridenbaugii App:
Stippich (BQ 10-4)
App.: Gang Staff:
Pol k Walkci Campe Int.:
Bridenbaugh i
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(Ccr.t.) PRJi0iED liEARIliG SCllEDULE OF WITflESSES FOR BLACK FOX RADIOLOGICAL HEALTH AfiD SAFETY HEARItiGS (February 26 - March 3, 1979 Tuesdn Wednesday Thursday Friday Saturdaz
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2fz7 2/28 3/1 3/2 3/3 tf A.:'..
iN,c.,
(T/ua CTD)
(TAPS CTD)
Reed heport Reed Report Rebuttal or e
.a Surrebuttal (cart. T ;-;'
St ':
._ ir r
d g
w.
,s-p i
i r es t (Tiwa CTD)
Loose parts Reed Report Reed Report Remaining monitoring Rebuttal or (BQ1-1)
Surrebuttal Staff:
Phillips Int.:
f4inor i
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Additional Supplemental Testimony for Radiological Contention Number 16 Black Fox Station, Units 1 & 2 by Mel Fields, John Kudrick, and Cecil Thomas The purpose of this supplemental testimony is to revise our original testimony on pool dynamic loads to reflect changes made by the Applicant since the original testimony was prepared.
Originally, the Applicant had committed to use the pool dynamic load criteria presented in NEDO 11314-08 (Preliminary) with certain changes to meet current NRC requirements. This information was presented as Appendix 3C to the PSAR and is discussed in our original testimony.
The Applicant in a meeting with the NRC staff on January 23, 1979, stated that Appendix 3C was to be revised to include the latest refined pool dynamic load criteria developed by General Electric Company and presented in GE Report 22A4365, " Interim Containment Loads Report (ICLR) - Mark III Containment," Rev. 2 dated November, 1978. The ICLR Rev. 2 consists primarily of load definition refinements brought about by interpretation
.of additional test data, confirmatory in nature, that have been generated
.since the original pool dynamic load criteria were issued. Also included in this report is clarification of the treatment of multiple SRV loads in determining the structural response to be used for the equipment evaluation (i.e., time phasing and statistical considerations).
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Tre Applicant has examined the loads provided in ICLR Rev. 2 with respect to the design of the containment systems and found, with two exceptions, that no changes were required in the design requirements of the containment structure. These design changes, strengthening the weir wall and possibly relocating or adding more stiffeners on the outside of the containment steel vessel, are not of the type that would normally be reflected in licensing requirements; nevertheless, the Applicant wanted to make the NRC staff aware of them. We acknowledged the infonnation provided by the Applicant and agreed that the design changes are minimal and can be readily accommodated in the Black Fox Station containment design.
Although the staff has not completed its review of ICLR Rev. 2, it is our judgement that the information available at the time the Safety Evaluation Report for Black Fox Station was prepared is sufficient to demonstrate the adequacy of the present design and that the additional confirmatory tests completed since the SER was published will have no significant effect on the present day design basis.
The Applicant has also indicated that the methodology contained in ICLR Rev. 2 to establish the loads on equipment from multiple SRV actuation would be used for the equipment to be installed at the Black Fox Station.
s.----.
Previously, our position was that, for the cases of multiple valve actuation (including the case of all 19 valves actuating), the quencher loads from each SRV line shall be assumed to reach the peak pressure simultaneously and oscillate in phase.
Our position also indicated that these in phase loads should be applied not only to the containment structures ~and internals, but also to any equipment and components that could experience these loads.
In a meeting with the Applicant on January 23, 1979, the Applicant stated that the containment structures would be designed to accommodate the loads associated with the simultaneous actuations of all 19 SRVs with all the bubbles assumed to oscillate in phase in the suppression pool. The Applicant also stated that this assumption results in the maximum vertical loads which control the design of the containment structure.
Powever, the Applicant was not prepared to commit to design the equipment and components on the same basis. Rather, the Applicant stated that the in phase case was not necessarily bounding or conservative for design of all the equipment and components.
Conservative assessment of overturning moments, for example, would require consideration of out-of-phase bubble oscillation. Accordingly, for design of equipment and components, the Applicant proposed to utilize those techniques described in the above cited ICLR, Rev. 2, as it may be revised upon completion of the staff's review.
. -. ~.
4_
e With respect to construction schedules, the Applicant stated that installation of equipment and components affected by the subject design techniques would not occur until well af ter the first quarter of 1980.
~
Accordingly, if design changes were needed to accommodate the loads associated with use of a " staff-approved" version of the techniques described in ICLR, Rev. 2, the Applicant stated that these changes could and would be made prior to installation of the affected equipment and components.
If, however, the Staff review extended beyond the first quarter of 1980, the Applicant would install equipment on his own risk.
Upon final resolution the Applicant would replace or modify all affected equipment and components, as required.
The staff's review to date of the techniques described in ICLR, Rev. 2 has found that there is merit to its use as proposed.
However, the staff has also found that a number of important concerns need to be resolved before any approval of the techniques can be assured.
On the basis of the above discussion, taking into account the absence of any foreclosure of utilizing a " staff-approved" version of the design techniques described in ICLR, Rev. 2 in design of the equipment and components for the Black Fox Station, the staff finds acceptable:
- 1) the Applicant's proposed use of the assumption that all 19 S/R valves will actuate simultaneously and the resulting bubbles in the suppression pool will oscillate in-phase in design of the containment structures; m.+,
a9 4en ev=-emo ewes.o=e-4=*-e-w=+'
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5-r 2) the Applicant's proposed use of the generic resolution; f.e., the
" staff-approved" version of the design techniques described in ICLR, Rev. 2, in design of the affected equipment and components; and
- 3) the Applicant's commitment not to install any affected equipment or components prior to second quarter of 1980 or until a " staff-approved" version of the design techniques described in ICLR, Rev. 2 is available and is used in analysis of the affected equipment and components.
If, however, the Staff review extended beyond the first quarter of 1980, the Applicant would install equipment on his own risk. Upon final resolution the Applicant would replace or modify all affected equipment and components, as required.
It should be noted that our current schedule for review and evaluation of the phasing methodology indicates completion by the first quarter of 1980.
It is the staff's view that the above detailed basis is acceptable to both the staff and applicant and permits the staff to conclude its construction permit stage of review of this matter for the Black Fox docket.
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r UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE TIIE ATOMIC SAFETY AND LICENSING BOARD In the Matter
)
)
PUBLIC SERVICE COMPANY OF OKLAHOMA
) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC.
)
STN 50-557 AND WESTERN FARMERS COOPERATIVE, INC.
)
)
(Black Fox Station, Units Nos. 1 and 2) )
SUPPLEMENTAL TESTIMONY OF STEVEN A. VARGA LOAD COMBINATION METHODOLOGY My name is STEVEN A. VARGA.
I am employed by the Nuclear Regulatory Commission as a branch chief in the Division of Project Management, Office of Nuclear Reactor Regulation.
The purpose of this testimony is to provide an update of my previous testimony on Contention 16 in light of recent meetings with the applicant on load combinations methodology.
First, a chronology of events leading to the most recent discussions with the applicant would be useful.
1.
On August 7, 1978, we sent a Ictter (Enclosure 1) to the applicant requesting his commitment to the forthcoming generic resolution on load combination methodology.
2.
On August 18, 1978 the applicant responded (Enclosure 2) and reaffirmed his position that SRSS is an acceptable 4
method for load combinations and declined to commit to the generic resolution.
3.
On October 31, 1978 we sent a letter (Enclosure 3) to the applicant apprising him of the results to date of our generic evaluation (NUREG-0484) which allowed a limited application of the SRSS method and that until such time that additional studies may allow extension of the SRSS method to other systems or components, we will require conformance to present criteria for those systems or components or load combinations not included in the conclusions of the staff report, NUREG-0484.
4.
On December 20, 1978 the applicant responded (Enclesure 4) by acknowledging the results of NUREG-0484 for combining LOCA and SSE loads within the reactor coolant pressure boundary and its supports and, in recognition of continued staff efforts in considering the application of SRSS methodology to other systems and components, proposed the use of certain criteria (the proposed "Newmark-Kennedy Criteria") for load combination methodology.
A meeting was held with the applicant on January 23, 1979 to discuss his proposal.
The staff concluded that, altheugh
the proposed criteria had merit and was actively under review as part of our continuing generic review of this matter, it was not yet in a position to make a determination on this matter and pending that deter-mination we will require a commitment to present criteria for load combination methodology.
Thus, the conclusion of my previous testimony on this matter remains unchanged.
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WASmNG TON D. C. 70555 j
,C August 7,1978 Docket lio. STil 50-556/557
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Public Service Company of Oklahoma ATTfi: fir. T. ft. Ewing, flanager Black Fox Station fiuclear Project
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P. O. Box 201 3
Tulsa, Oklahoma 74102 P.
Gentlemen:
SUBJECT:
t1UCLEAR STEAM SUPPLY SYSTEft AtlD BALANCE-0F-PLANT EQUIP!1EtiT i
ADEQUACY EVALUATION FOR BLACK FOX STATION, UNITS 1 AND 2 C
qi The evaluation of the nuclear steam supply system (NSSS) equipment for lf the BUR /6 design was discussed generically at a meeting held with the 3;
General Electric Company (GE) on March 1, 1978.
7 5
As you are aware, the staff's Black Fox Safety Evaluation Report describes the overall dynamic analyses which will be required at the 4}
final design stage to confirm the structural design adequacy of the I
flSSS equipment. Our generic review of this matter is proceeding to characterize the dynamic loads, load combination acceptance criteria, 4
and load combination methods which meet regulatory requirements and I
can be used by GE in developing its final design.
We anticipate that
}
we will provide interim guidance to GE during the sumer of 1978 1
concerning our preliminary judgment on the relative conservatisms to i
be required of the individual evaluation assumptions in its overall
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dynamic analysis program.
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i To further assure the acceptability of the final design of the NSSS 3
components and in order for the staff to recommend proceeding with the j
forthcoming radiological safety hearings and to support the issuance g
of construction permits for the Black Fox facility, we are requesting q
a commitment from General Electric (your standardized NSSS supplier) g to analyze the dynamic effects for the final design of the standardized a
fiSSS components in a manner consistent with the forthcoming generic A
resolution and to submit the results for review and approval by the 3
staff.
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Public Service Company of Oklahoma, -2,-
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.Simi1arly, we request a coranitment on your part (1) to adopt the General
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Electric solution for tiSSS components as finally reviewed and approved t
by the staff and (2) to analyze the dynamic effects on the final design
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of all balance-of-plant components in a manner consistent with the
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generic resolution and to submit the results for review and approval by the staff.
We believe the scheduling of our generic solution, the probable h
nature of the required modifications, if any, to equipment or supports l'
resulting from this reanalysis, and the status of the Black' Fox project j
provide reasonable assurance that no foreclosure of alternatives will
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occur as a result of proceeding in this manner.
1' If you require any clarification of the matters discussed herein, plea',e
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contact the staff's assigned licensing project manager.
Sincerely, J
W I
Roger S. B yd, Director
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Division of Project flanagement
- c Office of tiuclear Reactor Regulation ccs
Listed on following page 1
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1 Public Service Company of Oklahoma
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Michael I. Miller, Esq.
Ms. Roberta Ann Paris Funnell h Isham, Lincoln & Beale 3115 Harvey Parkwy
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One 1st National Plaza
- 0klahoma City, OAlanoma 73118 E Suite 2400 Chicago, Illinois 60606 Mr. Lawrence Burrell
,j Route 1, Box 197 t
Robert L. Lawrence, Esq.
Fairview, Oklahoma 73737 f'
General Counsel h,
Put'lic Service Company of Oklahoma Ms. Sherri Ellis P. O. Box 201 3228 Northwest 44th Street r_
Tulsa, Oklahc.na 74102 Oklahoma City, Oklahoma 73132 I
t;p, Mr. M. E. Fa te, J r.
Fr. John B. Axton T
Vice Presi. dent - Power P. O. Box 388 Public Sersice Company of Oklahoma Coalgate, Oklahoma 74538 b
P.O. Box 201 F
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Tulsa, Oklahoma 74102
.3 Mr. Maynard Human t-General Manager y
Western Farmers Electric Cooper)tive g
P. O. Box 429 a
Anadarko, Oklahoma 73005 7
t Mr. Gerald F. Diddle General Manager
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Associated Electric Cooperative, Inc.
P.O. Box 754 Springfield, Missouri 65801 6
3 Ms. Carrie Dickerson E.
Citizens Action for Safe Energy, Inc.
I P.O. Box 924 1/
Claremore, Oklahoma 74107 f
Ms. Ilene H. Younghein
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3900 Cashion Place C
Oklahoma City, Oklahoma 73112 A
Andrew T. Dal ton, J r., Esq.
2536 E. Sist Street i
Tulsa, Oklahoma 74105 k?
Wallace Byrd, M.D.
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316 N. Byrd P.'
'Coalgate, Oklahoma 74538 h
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ENCLOSURE 2 6212 DIN 8-012-334
,PUBLIC SERVICE COMPANY OF OKLAHOMA y$
b A CENTRAL AND SOUTH WEST COMPANY P O box 2ol i TULSA. OKLAHOMA 74102 ' t918) 58313611'
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Public Service Company of Oklahoma August 18, 1978 Black Fox Station File 6212.125.3500.2 NSSS &' BOP Equipment Acequacy Evaluation Docket STN 50-556 and STN 50-557 Office of Nuclear Reactor Regulation Division of Project Management U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Attention:
Roger S. Boyd, Director Gentlemen:
We have reviewed your letter of August 4,1978 regarding the Staff's generic resolution of the analysis methodology to be used by General Electric in their equipment adequacy evaluation program.
You have requested a commitment from both Public Service Company and General Electric to analyze the dynamic effects for the final design of the standardized NSSS components in a manner consistent with the forthccming generic resolution and to submit the results for Staff review and approval.
In addition you have requested a similar commitment from us regarding BOP analysis.
It is our understanding that the primary point of discussion pertains to the acceptability of the SRSS load combination method.
We have previously submitted our position on the use of SRSS at Black Fox Station by letter dated July 7, 1973 in response to Staff question 130.19, The operative portion of that submittal is attached for your convenience.
Therein we acknowledged the expectation of a generic resolution in this matter which we would review for applicability to our project.
Absent the existence of your generic resolution for our review and censideration, Public Service Company cannot commit to such an open-ended proposition either in whole or in part. '
We respectfully reassert the attached position; if still unacceptable to the Staff we renew our request that our position in this matter be carried forward as an open item in your upcoming SER supplement and placed as an issue in controversy before the Atomic Safety and Licensing Board.
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782370259
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Public Service Company of Oklahcma August 18, 1978 Black Fox Station Page 2
'NSSS & B0P Equipment Adequacy Evaluation Docket STN 50-556 and STN 50-557 If you. require any clarification of the matters discussed herein, please contact Mr. Vaughn L. Conrad, Manager, Licensing & Compliance.
Very truly yours, f71 La T. N. Ewing Manager, B1 x Fox Station Nuclear Project TNE:VLC:fd Enclosures e
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Question 130.19 Question:
The procedures outlined in Section A8.4 of Appendix 3C and described in Section 16.0 for combining simultaneous transient loads are not acceptable to the staff.
It is the staff's position that the SRSS method may be used only for the combination of seismic loads in three directions,
but not for the total combination of all loads.
For the latter, the combination should be done by the absolute sum method unless additional Laforsation is provided to justify your method on a case-by-case and item-by-icen basis.
Response
Appendix 3C, Section A8.4, which is entitled "Reco= mended Design Load Sn-mntion," was added to the BFS PSAR and filed with the staff for review on April 1, 1977 and utilized the SRSS method for combining certain simultaneous loads.
Subsection 3.8.1 of the Safety Evaluation Report for BFS, NUREG-0190, dated June 1977, entitled " Steel Containment," paragraph 8, states that:
"We have concluded that the criteria that will be used in the analysis, design, construction, and testing of the steel containment structure to account for the loadings and condi-tions that are anticipated to be experienced by the structur.
during its service lifeti=e are in confor=ance with estab-11shed criteria as contained in codes, standards, and specifications acceptable to the staff."
Additionally, Subsection 3.8.2, entitled " Containment Interior Structures," paragraph 5, states:
'?We have concluded that the criteria that will be used in the design, analysis, and construction of the contain=ent intern structure, to account for anticipated loadings and postulate; conditions that =ay be imposed upon the structures during their service lifetime, are in conformance with established criteria, codes, standards, and specifications acceptable tc the staff."
Hence, the Applicant believed that the SRSS method was acceptable as described in Section AS.4 of Appendix 3C.
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Response 130.19 (Continued):
It is recognized that Roger J. Mattson, Director, Division of Systems Safety, has established a Task Force on SRSS for the express purpose of providing General Electric Co=pany guidance regarding the acceptability of SRSS in connection with their NSSS Equipment Adequacy Evaluation.
In a memorandum dated June 27, 1978, Colleen P. Woodhead, Counsel for NRC Staff, advised the ASLB panel for Black Fox Station of the GE program and noted that interim guidance can be expected about August 1, 1978.
Ms. Woodhead noted that the load combination unhodology has been under review by the staff for some time as one item in generic technical activity 3-6.
The Applicant hereby reaffirms its position that SRSS is an acceptable =ethod for load ce=binations. We assert that the Staff has previously accepted the SRSS =ethod for use at 3FS.
Af ter the generic resolutica is made by the staff on the above matters, the Applicant will consider that determination in relation to the material presented in SFS PSAR Appendix 3C, Section A8.4.
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STN 50-556 STN 50-557 Mr. T. N. Ewing, Manager
~ ~ Black-Fox Station Nuclear Project Public Service Con; 'ny of Oklahoma P. O. Box 201-Tulsa, Oklahoma'.74102 Gentlemen:
MEE0DOLOGY IDR COSBIkING Ih'NAMIC RESFONSES OF STRUCTURES,
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SUBJECT:
SYSTEFS AND COMPONENTS IMPORTANT TO SAFETY (BLACK IDX STATION, UNIT NOS.,1 ANT 2)
In Supplement No. I to our Safety Evaluation Report related to the construction of the proposed Black Fox Station, Unit No.1 and 2, NUREG-0190, we indicated that we were evaluating generically the methodology for corbining the dynamic responses of structures, systems and co.Tponents important to safety. We further indicated that we expected to have avniinhle the results of our evaluation of this matter by mid-Septerber 1978.
A working group was established to develop reco=endations for combining dynamic responses in the analyses of stnictures, systems and components important to safety. The working group has completed its evaluation of this catter and its findings are documented in a report entitled " Report on Methodology for Corbining Dyna.mic Responses,"
NUREG-0484,.a copy of which is enclosed. We have adopted the working group's recommenda.tions which. states:
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"For -the codbination of the dynamic responses within the Reactor Coolant Pressure Boundary (RCPB) aid its supports, which result from the coincidence of an SSE and LOCA, the Square Root of the Sums of the' Squares (SRSS) technique is acceptable contingent upon perfomance of a linear clastic dynamic analysis to meet the appropriate ASME code,Section III, Service Limits."
This conclusien endorses a limited" application of the square-root-of-the sum-of-the-squares (SRSS) mathodology; namely, to.the reactor coolant pressure boundary. The report ~does not preclude future extension of the SRSS method to other systems or components. or to
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Mr. T. N. Ewing.
other load combinations when appropriate bases have been demonstrated, g
and provides guidance as to the additional work that should be pursued to attain that goal, but suggests that such bases have not yet been sufficiently quantified. 'Ihe staff has these suggestions under considerati.on for incorporation in our long-term studies.
For all such other systens or components or load co:rbinations not included in the conclusions of the staff report, NUREG-0484, we require conformance to present criteria for the load combination methodology.
h'e request that you amend your application to reficct a commitment to perform the required dynamic analyses for the proposed Black Fox Station structures, systems and components important to safety in-accordance with the provisions of working group's report. Should you desire further information or wish to meet with us concerning this matter, please contact Cecil Thomas, Project Manager for the proposed Black Fox Station.
Sin 6prely, s
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Igen L [ WA teven A. Varga, ne j Light Water Reactors % anch No. 4 Division of Project Management
Enclosure:
Report on Methodology for Combining Dynamic Responses cc: See next page e
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ENCLOSURE 4
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PUBLIC SERVICE COMPANY OF OKLAHOMA e
A CENTRAL AND SOUTH WEST COMPANY f
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t P O DOX 201/ TULSA. OKLAHOMA 74102 / (918) 583-3611 December 20, 1978
- u. erin t. n re, se.
Eaecutiw %ce Presafsnt File: 6212.125.3500.::
Public Service Company of Oklahoma Black Fox Station Methodology for Combining Dynamic Responses of Structures, Systems and Components Important to Safety Docket STN 50-556 and STN 50-557 Office of Nuclear Reactor Regulation Division of Project Fbnagement Light Water Reactors Branch No. 4 U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Attn: Steven A. Varga, Chief Gentlemen:
A working group of the NRC Staff issued a report in September 1978 (NUREG-0484) that reviewed and discussed the present methods used by nuclear power station designers for combining various load combinations.
As a result of the NUREG-0484 analysis, the NRC Staff approved the application of the SRSS methodology for combining SSE and LOCA dynamic loads within the reactor coolant pressure boundary and its supports.
This application of the SRSS methodology was in addition to uses previously approved by the NRC Staff, i.e.,
the combining of certain modal combinations and direction combination responses caused by earth-quakes, and the combining of LOCA and SSE loads in the analysis of PWR fuel bundles.
Significantly, NUREG-0484 recognized that the NRC Staff is actively considering the application of SRSS methodology to the evaluation of Mark I and Mark II contain-ment dynamic loads, and suggested the development of the justification to expand the application of SRSS methodology on a generic basis.
Since the publication of NUREG-0484, a significant new development has occurred with the creation of the criteria justifying the generic use of SRSS in combining loads for reactor design by Dr. Nathan Newnark and Dr. R. P. Kennedy (hereinafter referred to as the "Newmark-Kennedy Criteria"). A copy of the criteria is attached They were most recently discussed during a meeting of the ACRS Fluid Hydraulic Dynamic Effects Subcommittee on November 28-30, 1978.
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Page 2 December 20, 1978 We believe the criteria justify the use of SRSS methodology for generic nuclear power station design, and that the criteria can and should serve as a basis for complying with your October 31, 1978 request for a PSAR commitment on load combina-tion methodology for the design of the Black Fox Station.
In view of the foregoing, the Applicants are prepared to submit an amendment to the PSAR for the Black Fox Station which would provide:
A.
Although we believe that the above Criteria allow the use of the SRSS methodology for structures, the absolute summation method will be used to combine dynamic loads as defined in Appendix 3C of the PSAR, as amended through Amendment 13, in connection with the design of structures.
If these loads are subsequently redefined in a significant manner, or if loads for structures other than those defined in Appendix 3C are defined, the loads may be combined as provided in B. below or by absolute su=mation.
B.
The SRSS load combination method will be used to combine dynamic loads defined in connection with the design of piping and components for the entire Black Fox Station provided that each load can be shown to meet criterion 1. or each load combination can be shown to meet criterion 2.
of the Newmark-Kennedy Criteria. Applicants agree to cooperate with the NRC Staff during the implementation of the Newmark-Kennedy Criteria, including the opportunity to review the actual application of the criteria by Applicants' NSSS supplier and Architect / Engineer.
C.
In the unlikely event the application of the 3RSS load combination method-ology for any particular load or load combination, specified in A or B above, results in a failure to meet both criteria 1. and 2. of the Newmark-Kennedy Criteria, Applicants agree to cooperate with the NRC Staff to detetuine an acceptable load combination method.
We would be pleased to meet with you to explain and discuss further the details of our proposal.
s'D Sincerely,
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2 October 1978 CRITERIA FOR COMBINATIONS OF EARTHOUAXE AND/OR OTHER TRANSIENT RESPONSES r
Preamble.
The intent of the methods proposed f,or combinations of transient, dynamic responses is to achieve a non exceedance prcbability of approx-imately 84 percent for the peak combined response of the system, com-ponent, or element considered.
This goal is achieved by compliance with any one of the following criteria, or any alternative method that meets the intent stated above, provided that the intensity of loads or acceler-ations for each input are conserva'tively represented (approximately at the level of the 84th percentile, or the mean plus one standard deviation, of the expected input intensity).
1.
Criterion.
Dynamic or transient responses of structures, components and equipment arising from combinations of dynamic loading or motions may be combined by SRSS provided that each of the dynamic inputs or responses has characteristics similar to those of earthquake ground motions, and that the individual component inputs can be considered to be relatively uncorrelated; i.e., the individual dynamic inputs or responses considered are either from independent events or have random peak phasing.
This similarity involves a limited number of peaks of force or acceleration (not more than 5 exceeding 75 percent of the maximum, or not more than 10 exceeding 60 percent of the maximum), with approximately zero mean and a total duration of strong motion (i.e., exceeding 50 percent of the maximum) of 10 seconds or less.
Explanation.
v' Since earthquake motions in various directions produce responses which are combined conservatively by the use of SRSS, the descriptions of dynamic ce te---taa' i-y
The coef ficient of correlation for these is less than 0.4, and motions.
the pattern of peaks is based on Table 2 of Circular 672 of the USGS describing earthquake ground motions for use in the design of the Alaska Oil Pipeline.
The probability distribution for the, responses to earth-quake motions is based on the concepts underlying U. 5. tiRC Regu'latory Guide 1.60, where the standard deviation is 30 to 40 percent of the median value.
It has been proved some decades ago that modal response,s to earthquake motions may be conservatively combined by SRSS methods with the same degree of conservatism as th5t of the motions.
If each of such response.
is considered to be at the level of mean plus one standard deviation, the SRSS value is also at this level.
For the same reasons, responses from the three component directions of earthquake motions may'also be conservatively combined by SRSS methods.
2.
Criterion.
When response time-histories are available for all multiple dynamic loadings being combined, SRSS methods may be used for peak combined response when CDF calculations, using appropriate assumptions on the range of possible time lags between the response time-histeries, show the following criteria are met:
a.
There is estimated to be less than approximately a 50%
conditional probability that the actual peak combined response from these conservatively defined loadings exceeds approxi-mately the SRSS calculated peak response, and b.
There is estimated to be less than approximately a 15% con-ditional probability that the actual peak combined response exceeds appror.imately 1.2 times the SRSS calculated peak response.
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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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PUBLIC SERVICE COMPANY OF OKLAHOMA
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Docket Nos. STN 50-566 AS':0CI ATED ELECTRIC COOPERATIVE, INC.
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STN 50-557 AND WESTERN FARMERS COOPERATIVE, INC.
)
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(Black Fox. Station, Unit Nos. I and 2)
TESTIMONY OF Abdel Hafiz My name is Abdel Hafiz.
I am employed by the Nuclear Regulatory Commission as a Structural Engineer in the Division of Systems Safety, Office of Nuclear Reactor Regulation.
I have been employed in this position since January 1975. My professional qualifications are attached.
The purpose of this testimony is to reply to the Board's question 2, which reads as follows:
2.
In clarification of certain matters raised in the January 1978 report of Messrs. Masri, Seide and Weingarten, which was trans-mitted to the Board by Mr. Paton's letter of December 22, 1978, Staff and/or Applicants are requested to present evidence in response to the following:
a.
"Has any Staff evaluation (other than the one paragraph enclosure to Mr. Paton's letter) of significance of this report been made?"
The subject report is the first quarterly progress report on containment buckling prepared by International Structural Engineers, Inc., which is under contract to the Staff to study this matter. The Staff has not prepared any written technical evaluation of this report. The one ay
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s paragraph enclosure to Mr. Paton's letter is not considered to be a technical evaluation of the report, but a brief status report on factors of safety against containment buckling.
However, the staff looked at the subject report and concluded that the information contained therein was preliminary in nature. The final report of this contract is scheduled to be completed by the end of February 1979.
It is expected that any final design recommendations or guidelines ultimately resulting from this program will be evaluated for possible use in our licensing review work.
b.
"The report (at pp. 2 and 3) is severely critical of two out of three 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?"
Neither of the two criticized methods will be relied upon in the design of Black Fox.
In response to staff questions, the applicant indicated that a minimum factor of safety to be used in the design will be 2.0.
The method of analysis used to compute the factor of safety will be based upon classical (linear) analysis reduced by appropriate knock-down factors; knockdown factors reflect the difference between theoretical and actual load capacities.
This method was considered acceotable in the subject report.
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 determined by other methods?"
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' The applicant at this stage has not presented the buckling calculation in the PSAR of BFS.
The Staff does not require such an analysis at
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the preliminary design stage. The result of such analysis will be pro-vided at the final design stage.
In response to staff questions, the applicant indicated that a classical (linear) analysis will be relied upon in the design of BFS as indicated in our response to question b above. The method set forth in section 5, page 4 and 5 of the subject report is considered a detailed methodology of the classical (linear) analysis. The Staff accepts the classical (linear) analysis method. However, the use of the detailed methodology addressed in Section 5, pages 4 and 5 is not endorsed by the Staff.
We shall consider the information already available as well as any information to be developed as a result of our contract on buckling, and we shall advise the applicant of any recommendations that should be considered in his design. We shall, also, review the information provided by the applicant at the final design stage.
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e Abdel Hafiz PROFESS 10tiAL QUALIFICATIONS STRUCTURAL ENGINEERING BRANCH I am a Structural Engineer in the Structural Engineering Branch of Nuclear Regulatory Commission.
I am responsible for the review and evaluation of adequacy of criteria used in the structural design and analysis of Seismic Category I structures, system and components of nuclear power plants assigned to the branch.
I received a B.S. degree in Aeronautical Engineering from Cairo University in 1966.
I received the degree of Master of Science in Mechanical Engineering from California State University of Long Beach in 1970.
In 1974, I received a Master degree and a Ph.D. (Civil Engineering) from University of Southern California, Los Angeles (USC).
From 1967 to 1970, I worked at Union Oil Company of California where I was responsible for operation of petro chemical refineries including start up, shut down and emergency conditions.
From 1970 to 1974, I joined USC, Civil Engineering Department as a research assistant.
During that period, I conducted research in the area of shell buckling, n*te. il frequencies of shells, non-linear vibrations and stress con-centr." ion around small holes.
From January 1973 to June 1974, I was employed by Bechtel Power Corporation where I was involved in design of nuclear piping class I and II.
As a member of Structural Engineering Branch, I have participated in developing criteria for structural design and analysis of Seismic Category I structures in nuclear power plants, performed evaluation of technical reports concerning structural behavior under accident loading conditions and reviewed the safety analysis reports of nuclear power plants of Phipps Bend, Marble Hill, Davis Bessee, River Bend, etc., in the areas relating to the design and analysis of Seismic Category I structure.
I am a member of Section XI of ASME Code on Inservice Inspection of Nuclear Power Plants and.an alternate member on Section III of ASME on Subsection NE.
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_Irl THE MATTER OF THE APPLICATI0fl 0F
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Docket flos.
PUBLIC SERVICE CUMPAi;Y UF UKLAHUMA,
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STfl 50-556 M SUCIATED ELECTRIC COUEERATIVE, IriC. )
STri 50-557 AND WESTERf1 FARMERS ELECTRIC COOPERATIVE (BLACK FOX UrlITS 1 Afl0 2)
Affidavit of K. M. Campe on Ultimate Heat Sink Cooling Tower Tornado Missile Protection
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5 Affidavit of Kazimieras M. Campe My name is Kazimieras M. Campe.
I am employed by the Nuclear Regulatory Comn;ission as a Nuclear Engineer in the Accident Analysis Branch.
I have been in this position since December 1972.
My qual-ifications are listed on an attached qualification sheet.
The purpose of my affidavit is to address the matter of tornado missile protection with respect to the applicant's proposed ultimate heat sink (UHS) cooling towers.
In the course of responding to Board Question 6-2 in the matter of Black Fox Units 1 and 2, the staff found that the cooling tower tornado missile protection was inadequate with respect to the mechanical draft fans.
My affidavit will show that in the interim, the applicant has made a commitment to provide tornado missile protection for the UHS cooling tower fans that will be satis-factory to the staff and that there is reason to believe that an engineering solution is feasible.
ULTIMATE HEAT SINK COOLING TOWER
~TTRNADO MISSILE PROTECTION As indicated in the staff's written testimony on Board Question 6-2 in the matter of Black Fox Station Units 1 and 2, the applicant's proposed design of the ultimate heat sink (VHS) cooling towers was found to be incomplete with respect to tornado missile protection.
Specifically, the staff's review indicated a lack of protection of the cooling tower fans against potential tornado missiles entering the fan discharge nozzle areas from above.
It was noted in the testi-many that this would be resolved and the findings reported in a later SER supplement.
Since the time that the above testimony was submitted, the applicant N
has made a commitment to protect the UHS cooling towers in their entirety, i.e., inciuding the mechanical draft fans, against tornado missiles.
We have reviewed a cooling tower fan tornado shield design on another case.
On the basis of discussions with the applicant in that case, we believe that feasible designs exist which are acceptable to the staff.
In view of this, we consider the proposed UHS cooling towers to be acceptable with respect to tornado missile protection and will reflect this in a supplement to the SER.
We also note that the staff will evaluate the applicant's detailed UHS design during the OL review and will determine at that time if the tornado missile protection provided for the UHS cooling towers is adequate.
3/ Letter from Public Service Company of Oklahoma to the NRC, November 7, 1978.
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I have read the foregoing affidavit and swear that it is true and accurate to the best of my knowledge.
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dZimieras M. Campe Sworn before me this 1 day of Idrweg,1979.
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KAZIMIERAS M. CAMPE PROFESSIONAL QUALIFICATIONS ACCIDENT ANALYSIS BRANCH DIVISION OF SITE SAFETY AND ENVIRONMENTAL ANALYSIS I am a member of the Accident Analysis Branch of the Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
My duties include the identification and evaluation of hazards to the safe operation of nuclear power plants due to accidents external and internal to those plants.
Part of these duties involve the evaluation of turbine missile risks.
I have reviewed turbine missile generation, strike, and damage probabilities for nuclear plant license applications since 1973.
I have personally examined and performed photographic documentation of the Shippingport and Gallatin plant turbine failures.
I have preparad most of the technical input for the Regulatory Guide 1.115 which addresses protection requirements against low trajectory turbine missiles.
In 1975 I co-authored a paper with J. Read of the Accident Analysis Branch on the subject of high trajectory turbine missile strike probabilities. Currently I, am one of the contributing authors to the Staff's forthcoming turbine missile evaluation report.
I graduated from the University of Connecticut where I received B.S. and M.S. degrees in Mechanical Engineering 1958 and 1960, respectively. Between 1960 and 1962 I completed some advanced mathematics courses at the Rensselaer Polytechnical Institute branch in East Hartford, Connecticut. During.this period I was employed by Pratt and Whitney at the CANEL Analytical Physics Group as an analytical engineer.
From 1962 to 1966 I attended Purdue University, w'here I received a Ph.D. in Nuclear Engineering.
From 1966 to 1972 I was employed by Hittman Associates, Inc. where I worked in the Radioisotope Department.
During this period my responsibilities included radiation shielding analyses, radioisotopic generator design, and computer code development for reactor core physics calculations. Since 1972 I have been employed by the Nuclear Regulatory Commission in the Accident Analysis Branch.
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