ML19260C260

Discusses 791109 Meeting w/C-E & C-E Owners Group Re Draft of Proposed Rept on ATWS in C-E Plants.Agenda Encl

ML19260C260
Person / Time
Issue date:	12/07/1979
From:	Parczewski K Office of Nuclear Reactor Regulation
To:	Hanauer S Office of Nuclear Reactor Regulation
References
FOIA-80-587, REF-GTECI-A-09, REF-GTECI-SY, TASK-A-09, TASK-A-9, TASK-OR NUDOCS 7912260126
Download: ML19260C260 (31)

Text

{{#Wiki_filter:.- k [ )o UNITED STATES g NUCLEAR REGULATORY COMMISSION a h. WASHINGTON, D. C. 20555 %..v.../ DEC , y, MEMORANDUM FOR: Stephen H. Hanauer, Director Unresolved Safety Issues Program THRU: Ashok C. Thadani, Task Manager, A-9 Task FROM: K. I. Parczewski, Reactor Safety Branch, DOR

SUBJECT:

ATWS MEETING WITH CE AND CE Pl. ANT OWNERS GROUP A meeting was held in Bethesda on November 9,1979, with Combustion Engineering (CE) and with the representatives of the CE plant owner's group to discuss the draft of the proposed report on ATWS in the CE plants. The meeting was opened by the NRC staff outlining the purpose of the meeting and indicating the importance for establishing a firm schedule for issuing the report. This was followed by the presentation made by a representative of the CE plant owner's group who described the nature and the function of the orouo. He indicated that this aroup is separate from the group formed previously to handle the Bulletin and Order concerns and it consists of executives from participatine utilities. He also stated that the ATWS renart, althouch orecared by CE, will be reviewed and issued by the owner's grouo rather than by CE. The NRC staff agreed with this approach. Following these introductory remarks, CE outlined the content of the recort and dis-cussed in more detail the material presented in its different sections. (1) The renort will address the BIN #1 issues raised in Mattson's letter of February 15, 1979, and will describe the early verification efforts reouf red by Alternative 3 of MUREG-0460, Vol. 3. The material prosided in the report will fulfill the commitments made by CE during the August 17, 1979 neeting (Slide 2). The CE plants will be grouped in three classes and the recort will address generically ATWS concerns for each individual class. It will be demonstrated that during ATWS structural inteority of the crimary coolant oressure boundary (PCPB) and the functionability of individual components will be maintained. Also the leakage throuch the reactor vessel "0" ring seal will be evaluated. The report will also contain a discussion of the conformance of the CE plants to the ATWS criteria other than RCS pressure and will

Contact:

K. Parczewski, RS/00R X28166 1616 215

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~# DEC Stephen H. Hanauer describe the systems used for ATWS mitigation. The evaluation pre-sented in the report will show that the requiremects of NUREG-0460, Alternative 3 are met by the operating CE plants. C2 has indicated that some of the infomation presented at the meeting is of the proprietary nature and should be treated as such. Therefore the proprietary data ha.ye been deleted from the enclosure 2 copy of the slides presented at the meeting. These proprietary slides are in-cluded as enclosure 3, which is to be withheld from the POR. (2) The following criteria used in evaluating post-ATWS condition will be defined in the report (Slide 4). a PCPB integrity and functionability for cooldown b Radiological releases smaller than 10 CFR 100 limits c Maintenance of coolable geometry d) Peak fuel enthalpy less than 280 cal /gm e) Low probability of DNB or limited clad degradation f) Containment pressure smaller than design The staff questioned use of the 280 cal /gm enthalpy limit as a criterion for fuel failure, however, they agreed with CE that the DNB criterion is overconservative. (3) The report defined the initial conditions and assumptions used in the evaluation (Slide 5). They are: (a) Nominal initial condition (b) Most positive moderator temperature coefficient (MTC) for 95% of core life. This is going to be shown by parametric study performed for individual plants. (c) Automatic auxiliary feedwater actuation as a mitigating feature (d) Reactor vessel flange "0" ring seal leakage. This leakage will be evaluated for different classes of plants. (e) Manual actuation of SIAS required for injection of boric acid solution needed for reactivity control. (4) The responses to ATWS of plants in each class will be evaluated in the report (Slide 6). The evaluation will consist of the following steps: Analysis of the models used in the evaluation Analysis of complete loss of nomal feedwater event Discussion of the Three Mile Island concerns related to ATWS CE discussed different aspects of the comolete loss of feedwater event analyzed in the report and illustrated it with the viewgraphs repre-senting the change of pressurizer pressure and reactivity with time 1616 216

.e' DEC ~' Step n H. Hanauer (Slides 7 and 8) for 3800 MWt class of plants and the change of pressurizer peak pressure with total pressurizer relief area and with moderator tenverature coefficient for all three classes of plants (Slides 9,10,11,12,13 and 14). CE analyses show that some negative reactivity was introduced even before boric acid injection due to the increase in primary coolant temperature (decrease of moderator density) and the generation of voids in the core region due to boiling. The staff expressed its concern about the generation of voids which may raise problems similar to those in TMI. The staff suggested that this question should be addressed in a manner similar to the cesponse to the Bulletin and Order Task Force request. CE responded that a proper analysis of this problem will be included in the report. The staff again asked CE to provide analyses of a stuck open PORY which would result in isolation si.onal actuation and a. subsequent increase in the primary system pressure. The staff asked CE to pay-close attention to the applicability of the codes to this type of ATWS event. CE presented a list of TMI concerns which will be addressed in the report (Slide 15). (a) Failure of PORY to open or close (b) Loss of offsite power (LOOP) (c) RC pumps operation and natural circulation (d) Reflux boiling (e Bortn precipitation (f Radiological releases (g Operator action The staff expressed their interest in the pump performance at high pressure when a slight change in tolerances is expected and in pump cavitation, natural circulation and reflux boiling problems. They felt that these problems should be discussed in the report. In response, CE indicated that the last two effects are insignificant because of a relatively small loss of primary coolant inventory expected during ATWS event. This small inventory loss would also preclude boric acid precipitation. The staff also noted that the earlier LOOP ATWS analyses may be invalid because of change in MTC. The report will. demonstrate that the radiological releases are within the 10 CFR100 criteria and that the operator action is not required immediately after the beginning of ATWS. 1616 217

MC 1 79 Stephen H. anauer -4_ (5) CE told that the RCPB post-ATWS structural integrity will be demonstrated in the report (Slide 16). The report will describe the models used and then will provide the analyses of the response of RCPB components (Slides 17 and 20) to 4000 psia pressure transient. It will be demon-strated that in most cases level C and D criteria are satisfied. These analyses will include the reactor vessel shell (Slides 18 and 19). The staff questioned ifsRCPB components in all plants will be bounded by the analyses provided in the report. CE assured that the CE plant owners will verify the results for individual plants and will provide their findings to NRC. At the present moment, it is known that all piping, with the exception of the line leading to the letdown heat exchanger, meet the level O criteria. (6) CE described the method of analysis used to evaluate the leaka occurs at the reactor vessel "0" ring seal (Slides 21 and 22) ge wnich and presented the results of the analyses. The analyses were performed for 2650 MWt and System 80 classes of plants. Analytical results dis-cussed included the system p essure at which seal leakage initiates and the variation in leakage as a function of system pressure. CE noted that at all calculated ATWS pressures, even though there is leakage at the vessel flange seal, the closure bolts stress level is below the material yield strength. The staff asked several questions on the hydraulic mechanism postulated in calculating the leakage flow and discussed the assumptions made by CE. (7) CE sunmarized its presentation by stating that the analysis presented in the report will demonstrate that the criteria specified in NUREG-0460 for Alternative 3 plants are met by the CE plants. At the conclusion of the meeting, the staff stated that the outline of the proposed ATWS report has indicated several new areas which would have to be carefully eve'uated by the staff after the report is submitted for their review. Esmially the sections of the report dealing with the discussion of models am codes and with stress considerations would have to be care-fully reviewed. The staff also asked the owners to provide schedule for responses to the remain'ng questions in the February 15, 1979 Mattson letter. It was agreed by CE and by the representatives of the owner's group that the report will be submitted to NRC by December 1,1979. An attendance list is provided in Enclosure 1. K.J. u - K. I. Parczewski Reactor Safety Branch Division of Operating Reactors Eaclosure: As stated cc: ATWS Distribution 1616 218 Meeting Attendees

4 ENCLOSURE 1 MEETING ATTENDEES ATWS EARLY VERIFICATION STATUS s NAM _E REPRESENTING R. C. L. Olson CE Owners Group Chris H. Poindexter CE Owners Group C. A. Kreps CE C. L. Kling CE W. E. Burchill CE C. R. Musick CE D. J. Ayres CE Dennis L. Terrill TVA K. I. Parczewski NRC/ DOR /RSB F. Odar NRC/ DSS /AB Medhat El-Zeftawy NRC/SD Francis Akstulewicz NRC/DSE/AAB M. Srinivasan NRC/ DSS /ICSB F. C. Cherny NRC/ DSS /MEB M. D. Stolzenberg NRC/RES/RSB D. K. James FP&L (CE Owners Group) S. H. Hanauer NRC H. VanderMolen NRC/ DOR /RSB Kulin D. Desai NRC/ DSS /MEB bfb 2l9

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.:a c ~ N., '. AGENDA ."it ds C-E/NRC ATVS MEFTING .qq. - " ~4. n MOVEMRFR8,133 y;.J INTRODUCTION D. A. KREPS ATWS REPORT OVERVIEW C. L. KLING [, TRAllSIENT ANALYSES C. L. KLlHG 'S RCPB ANALYSES D. J. AYRES 'S 0-RiliG SEAL LEAKAGE

0. J. AYRES

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.*~ Olik ATWS ANALYSES: RESPONSE TO 2/15/79 NRC LETTER ^ CENPD - 263 (DRAFT) . u-Ns ~ ' ' -

1.0 INTRODUCTION

A.: ...n'. . q.;.* 2.0 EVALUATION-0F RESPONSE TO ATWS FOR EACH PLANT CLA I J.0 DEMONSTRATION OF RCPB POST-ATWS STRUCTURAL INTEG AND FUNCTIONABILITY FOR EACH PLANT CLASS .4-

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4.0 ASSURANCE OF CONFORPANCE TO ATWS CilITERIA OTHE RCS PRESSURE t.

5.0 DESCRIPTION

OF SYSTEMS USED FOR ATWS MITIGATION s . ny-l-6.0 SUWARY AND CONCLUSIONS _,.i.,'. y 7 4-APPENDIX A SUPPLEMENTARY PROTECTION SYSTEM "5 1616 221 t

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, se, INITIAL CONDITIONS AND ASSUMPTIONS 'si .x

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Wde.@+. ATWS CRITERIA I . :,c: ^ 'l ben FOR PEAK RCS PRESSURE DENxtSTRATE RCPB 96 INTE6P.lTY AND FUNCT10f! ABILITY FOR C00LDOWN ' .m. [%- Tf ^m.19;

• RA310LOGICA1. RELEASES LESS fHAN 10CFR100 LIMITS ~' 'ETs e.

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• LOW PROBABILITY OF DNS OR LIMITED CLAD DEGRADATION

~ . PEAK CONTAINENT PRESSURE LESS THAN DESIGN .gg

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• AUTOMATIC AUXILIARY FEEIMATER ACTilATION e REACTOR YESSEL FLANGE 0-RING SEAL LEAXAGE

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a EVALUATICN OF RESPONSE TO ~ f-{.. ATWS FOR EACH PLANT Cl. ASS 4'? ,3.

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~ .. m. '?!%r. -usy ANALYSIS MODELS i. 6:-) .q,. e., . _ r 'fd ANALYSIS OF COMPLETE LOSS OF NORMAL FEEDWATER -e ',- 1 THREE MILE ISt.AND CONCERNS RELATED TO ATWS i.- 3 e 1616 225 e e 4 t I .*~ ' ?g ' s k 4

slede. e 7 I i l i i 3900 g I m o-3400 N 5? O E 2900 5 ti 5 N 2400 u m 1900 1400 I I I I O 200 400 600 800 1000 TIME, SECONDS t c-e Are PRESSURIZER PRESSURE TRANSIENT DURING casesic aspaar 3800 MWt COMPLETE LOSS OF FEEDWATER 2-3 1616 226

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I DOPPLER N SHUTDOWN BORON 2 <3 EO4 L TOTAL \\ -6 5 6 MODERATOR cc -8 -10 12 I I I O 200 400 600 800 1C00 TIME, SECONDS A ua REACTIVITY COMPONENT TRANSIENTS DURING 9 c-E uws GENERIC REPORT 1616 227

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i i i I y4500 c. E'5' e 4000 E !5 C:f g 3500 Gu c. llm 6 3000 c. 2500 I I I I 0.10 0.15 0.20 0.25 0.30 0.35 2 PRESSURIZER TOTAL RELIEF VALVE AREA, FT c-E n ws EFFECT OF PRESSURIZER RELIEF AREA ON FEAX Figure PRESSURIZER PRESSURE DI)EDWATER RING 38CO MWt GENEA!C REPCRT COMPLETE LOSS OF Ft 2-35 1616 228

stMe.4to E i i i i i $4500 ~ w si 5; e 4000 f ei C:t 83500 C, u c. M $3000 . c- -t Q l i I l 1 -0.5 -1. 0 -1.5 -2.0 -2.5 -3.0 MODERATOR TEMPERATUE COEFFICIENT,10-46p/ F 0 16.16 229 c-E Arus EFFECT OF MODERATOR REACTIVITY FEED 8ACK ON F 5 FEAK PRESSURIZER PRESSURE DURING GENERIC REPORT 3800 MWt COMPLETE LOSS OF FEEDWATER 2-32 Jo

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PRESSURIZER TOTAL RELIEF VALVE AREA, FT EFFECT OF PRESSURIZER RELIEF VALVE AREA ON F'wm c-E Ar.vs PEAK PRESSURIZER PRESSURE DURING '-3' GENERIC REPCRT 2560 MWt COMPllTE LOSS OF FEEDWATER n

sicd e ** 1.t 5000 i i G 4500 :- c. ~ W' 5 m 0 4000 E 5 Ci c: m 3500 - O E x @3000 gg i i I I O -0.5 -1.0 -1.5 -2.0 -2.5 MODERATOR TEMPERATURE COEFFICID1T,10~4aplop EFFECT OF MODERATOR REACTIVIT/ FEEDEACK ON F "N" c~*Arws GENERIC REPCR7 PEAK PRESSUP.IZFR PRESSURE DURIT:G 2560 MWt COMPLETE LOSS OF FEEDWATER 7 34 ~ 0 @P ]D '* T))d ]U%. 1616 231 T Q" ' :3hN wN

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PEAK PRESSURIZER PRESSURE DURING GENERIC REPORT 34XX MWt COMPLETE LOSS OF FEEDWATER 2'20 a 3mm G~ 1616 232-~

S \\ide.'* \\ A 5000 l i $4500-y- s m $ 4000 - c:: C. 5 Cf s 3500 m tA I!i n. M 6 3000 n. Um 0 -1. 0. -2.0 -3.0 MODERATOR TEMPERATURE COEFFICIENT,10-4apl F 0 c-s A r.vs E:FECT OF MODERATOR REACTIVITY FEEDEACK O'!fA= PEAK PRESSURIZE? FRESSL'RE DURI.' 3dXX MWt COMPLETE LOSS OF FEEDWAN GE?:ERtc REPORT R 2-33

sua e w POST TMI CONCERNS ON ATWS i FAILURE f PORY TO OPEN FAILURE OF PORY TO CLOSE i. LOSS OF 0FFSITE POWER RCP OPERATION AND NATURAL CIRCULATION REFLUX BOILING BORON PRECIPITATION RADIOLOGICAL RELEASES OPERATOR ACTION lflg g34

stiAe. * \\ G DEM0ftSTRATI0ft 0F RCPB POST-ATWS STRUCTURAL INTESRITY AND FUNCTIONABILITY FOR EA01 PLANT CLASS I e'INTRODUCTI0li- . ETH00 0F ANALYSIS e ASE III AllALYSIS OF RCPB C0ff0NENTS e , VESSEL FUNGE 'O' RING SEAL LEAXAGE 1616 235 4 ,=.mmy . - = = - O

~ slid 417 COOEiLEVEL 0 ALLOWAst.E PSE55URE (FAULTEDI ATWS PRESSURE MME. ' CCOE LEVEL C ALLCWA8t.E PRES 3URE g (EMERGENCY) 7 5 = i 9 w 5- - ~ I T g x Y3 4 / / / /,' / / / i f/ / / h e s s 2 1 0 SJ S. G. 12" ELECW HEATER 42** EL3CW SJ RV & *R ESS. 35 35 43 TUBE ELEMENT 38 a36 a38 VESSEL DFSS PUMP (A PI (EXTERNAIJ OFSS WALL CASE FNMP CASE COMPCNENT 1616 236 m o M o m u l ti l* D)[l'R YQA 0 *

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c-s Arws PRIMdRY SYSTU.1 CCMPONENT LCA DS F 's'e GENERIC REPORT COMPARED TO COCE LEVE!.C AND D STRESSES 3-E ~~

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SIide.e.1o SIM ARY OF ATWS EVALUATION AT 4000 PSI 1. SURVEY OF PRIMARY COMPbHENTS - BAR CHART 2. SURGE LINE R n0W: STRESS - STRAIN CURVE (FIG. 1) 3. SHUTDOWN COOLING ISOLATION VALVE:16x12x16 MOTOR OPER. STPISSMAX = 17,100 PSI AT PIPE STRESSYIELD = 18,500 PSI 4. PRESSURI7FR SAFETY VALVE A. DISC AND N0Z7LE: FINITE ELEMENT ANALYSIS; DISC BEHAVES ELASTICALLY WITH SLIGHT LOCAL PLASTICITY. N0ZZLE REACHES YIELD AT 6350 PSI. a. INLET FLANGE AND BOLTING . FLANGE FIACHES LEVEL C STRESS LIMITS AT 4535 PSI (BASED ON DIRECT PRESSURE RATIO 0F DISCHARGE FORCE). BOLT STRESS REACHES LEVEL C STRESS LIMITS AT 7500 PSI. 5. TYPICAL 1E00 LB. VALVE EVflUATION STPISS LEVELS ARE WITHIN LEVEL C LIMITS AT 4000 PSI. 6. REGENERATIVE HX STPISS LEVELS ARE WITHIN LEVEL C LIMITS AT 4000 PSI. o o

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N FIGURE 3-4. VESSEL FLANGE OdING SEAL LEAXAGE t00EL ~ 1616 241

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Slide *J+ s M* w d 4 W 5 8 m i 1-I G } EC3 FUSSUU, PSIA FIGU E 3-6 O-RIE SEAL OPENING ARIA AS A FUNCTION OF FESSURE FOR THE 3800 Mist PIJerr CIJLSS f 1616 243 DPfD R D NfW l $n

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Sida..*.15 s c."5 w d a 5 8 u E 2 s ECS PRESSURE, PSIA FIGUEZ 3,3 0-EING SEAL OPENING AREA AS A FUNCTION OF PRESSURE FCE THE 2540 MWT AND 34XX MWT PRNT CMSSES I (( [i L

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