ML20010G629

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Forwards Potential Reactor Sys Voiding During Anticipated Transients & B&W Document 86-1107006-00, Sequential Auxiliary Feedwater Flow Analysis, in Response to NUREG- 0737,Items II.K.2.17 & II.K.2.19,respectively
ML20010G629
Person / Time
Site: Midland
Issue date: 09/17/1981
From: Jackie Cook
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To: Harold Denton
Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.17, TASK-2.K.2.19, TASK-TM 13800, NUDOCS 8109220328
Download: ML20010G629 (46)
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{{#Wiki_filter:' 4.., m Cubsumers Power James W Co.'k O GOf Vice Presidant - Projects, Engineering and Construction Gana al officas: 1945 West Parnell Road. Jackson, MI 49201 e (517) 78&o453 September 17, 1981 .!8 \\ SEpJ g 708/Ch "C 7 Harold R Denton, Director Office of Nuclear Reactor Regulation Division of Licensing '%g[sh4O US Nuclear Regulatory Commission Washington, DC 20555 ', e,. MIDLAND PROJECT S1 l i MIDLAND DOCKET NOS 50-329, 50-330 ~ RESPONSE TO NUREG-0737 ITEMS II.K.2.17 AND II.K.2.19 FILE: 0505.18 SERIAL: 13800 ENCLOSURE: (1) Potential Reactor System Voiding During Anticipated Transients Dated January 1980 (2) B&W Document 86-1107006-00, Sequential Auxiliary Feedwater Flow Analysis The enclosures to this letter are submitted as our response to NUREG-0737 items II.K.2.17 and II.K.2.19, respectively. The table entitled "TMI Action Plan Requirements for Applicants for an Operating Reactor" and contained in NUREG-0737 was used by us to determine the actions necessary to respond to NUREG-0737. This led us to believe that items II.K.2.17, Potential for Voiding in the Reactor Coolant System During Transients, and II.K.2.19, Sequential Auxiliary Feedwater Flow Analysis, need not be addressed by Babcock and Wilcox license applicants since a response had-been previously submitted by licensees. We were recently informed by the Staff that this previously submitted information should be provided on the Midland docket. This response to NUREG-0737 items II.K.2.17 and II.K.2.19 will be reflected in the appropriate volume of the FSAR. /' JWC/JRW/fms CC RJCook, Midland Resident Inspector 5 DSHood, US NRC TPSpeis, US NRC /( DBMiller, Midland Construction (3) RWHuston, Washington oc0981-0272a131 8109220328 810917 PDR ADOCK 05000329 PDR p

s _e, A m V b, POTENTIAL REACTOR SYSTEM VOIDING DURING ANTICIPATED TRANSIENTS - -~< ~..; ~.y. :... r..

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o o CONTENTS 1.0 Problem Introduction and Sumary 1.1 Introduction 1.2 Summary 2.0 Analysis, Assumptions and Event Descriptions 2.1 Analysis of Steam in the Upper Vessel Plenum 2.1.1 Stagnant " Hot" Water 2.1.2 Residual Heat Stored in Metal 2.1.3 Conclusion 2.2 Analysis of Steam in the Candy Cane 2.2.1 Pressurizer Outsurge Water Mixing in Hot Leg Flow 2.2.2 Residual Heat. Stored in Metal

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2.3 Event Description for Mass / Volume Balance 2.3.1 Description of Analysis 2.3.2 Derivation of Equations Used 2.3.3 Mass / Volume Analysis 2.3.3.1 Davis Besse 9/18/79 Trip 2.3.3.2 Davis Besse 10/15/79 Trip 2.3.3.3 Davis Besse 9/26/79 Trip '2.3.3.4 Oconee I 10/8/79 Trip 2.3.3.5 Oconee II 1/4/74 Trip 3.0 Conclusion 4.0 References e gin p

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o O List of Figures 2.2.1 Column Weldment Flow Paths 2.2.2 Upper Plenum Flow Paths 2.3.1 RC Volume Breakdown 2.3.3.1 TECO 9/18/79 Trip 2.3.3.2 TECO 10/15/79 Trip 2.3.3.3 TECO 9/26/79 Trip 2.3.3.4 Oconee 1 10/8/79 Trip 2.3.3.5 Oconee II 1/4/74 Trip A-1 Pressurizer level vs. corrected pressure tap measurements (320" pressurizer) A-2 Pressurizer level vs. corrected pressure tap measurements (400" pressurizer) A-3 Error in level indication vs. indicated level (400" pressurizer) A-4 Error in level indication vs. incicated level (320" pressurizer) ..... m.....:e........ s.,. -..., s... :, ; ..,,.. ~ 4,:.......,....,.._..... .: w.,,..., _.., _,,._. p. ..,s.... Attachments Pressurizer Level Error l l l t ,.-y y--pe7--- ,.,..,,-eq g.- e,-,-9.- .w-,g,,.-p_m,,-- ,-.y-w--g.y- ,,...---..-.p- ~em e--,*- -w-+ c.- -. +. -- -e --6 g----"

9 9 1.0 PROBLEM INTRODUCTION AND SLPHARY 1.1 Introduction The NRC has expressed concern that steam pockets can and do form in B&W primary systems during some reactor trips and furthermore, this steam could collect in the candy cane and potentially hinder natural circulation after loss of offsite power. The source of this steam is postulated to be from stored heat in metal, stagnant pockets of " hot" water, or pressurizer outsurge water (650 F); any or all causing flashing during low pressure periods after a trip. The purpose of this report is to show that the formation of a steam pocket in the primary system during most reactor trips is highly improbable and that the production of a volume of steam required to hin r natural :frculation is virtually . p. m z., 9........ ....,..... w -. u.:,....... ;;. a..,. - +4mposs%3e;.

z..
-.. -w 1.2 Sunmary Based on review of several past B&W reactor trips along with several assumptions and calculations discussed in Section 2,.

l the following statements can be made: (1) Minimum RC pressure after trips about the same time as the minimum coolant temperature (as expected) and not because of a primary system " Steam Pocket" restrict.ng a further pressure decrease. These minimum temperatures carrelatewith secondary side steam pressures. (2) With the most conservative assumptions steam pockets are virtually impossible during trar,sients where minimtra measured pressure is greater than 1740 psig.

  • Specific concern per reference 1 l

e m.e = -.-._,-,,,,,,,-,,.n-. ,.,.,-.-,,,-n,

O 3 (3) Using more realistic assumptions regarding potential steam production, it is unlikely that any net steam production can occur at measured pressures greater than 1500 psig. 2.0 Analysis, Assumotions and Event Descriotioni 2.1 Analysis of Steam in the Unoer Vessel plenum 2.1.1 Stagnant " Hot" Water Typically, the hottest bundle in a 2772 MWT core is a 1.5 relative peaked bundle. This localized power could conservatively result in localized stagnant water temperatures of s 620*F* in the upper control rod guide (column weldment). For this situation to occur, at least 53 in of flow area in the column weldment (see flow path 2 on figure 2.1.1) would have to be blocked. If this blockage could occur, then the following statement can be made: Any B&W reactor trio resulting

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..,g g.g g:y g - ,3yg yg9 g:9.g-k ~... s,.,.: ;. +, pressure i.s s 40 psid lower than core plenum pressure) voiding e may occur. 'Sincs this blockage is highly unlikely, a somewhat more realistic hot spot assumpt..on would be localized maximum temperatures of 610*F (instead of 620') in the upper vessel plenum. Th!s assumption is based on (1) water stagnating (or having an extremely lov flow rate) in the upper plenum (per Fig. 2.1.2) and acquiring the steady state enviromental temperature of the 59 column weldment ficw temperatures and (2) plugged return holes (per figure 2.1.2, Flow Path 3). This condition would not allcw flashing until the pressure decreases to 1660 psia, i.e.,1605 psig at the pressure tap. Typical outlet temperatures were verified frem TMI-2 pre Maren, 1979 100% power outlets themocouple data.

9 9 The most realistic assumption is the desian condition of % 5% core flow through the guide tubes into the upper plenum. Following a reactor trip, the coolant temperature in this region will decrease rapidly to below 600F as core outlet temoerature-eliminating the potentip for " hot" water pockets at any pressures above HPI injection levels. 2.1.2 Residual Heat Stored in Metal After a trip, the metal in the upper vessel-will be conservatively assumed to remain at 610 F until a minimum pressure is reached. The upper plenum coolant temperature at this minimum pressure should be

  • 550 F (in this region) but 575 F will be conservatively assumed due to hypothetical " slow" mixing. The available stored energy in the metal will be:

Stored Energy = Cp x Mass x AT

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n....,;, A....,,,... e.

. x,,,,..:.p.;,.y.... :. .w .., g. ,..:.u .:> y. <.... ' 4 ' .Cp,=.11' BTU /lby F-s. AT = 610 - 575 = 35*F Mass,s 5x10' lbm = total mass of metal above the outlet hot leg piping including reactor vessel Stored Energy = 1.925x10 BTU's The amount of water in this region is 1366 ft. 2 4 1366 ft.3 x 44.5 lbm/*t

.08x10 lb of water Conservatively assuming this energy ; instantly released to the water, the water temperature will increase from 575 F to s 601 The pressure required for flashing at this temperature is s 1505 psig (at the pressure tap). Therefore, in order to fill this region with
  • This is water density at 575 F,1500 psia.

6

    • 1.925x10 STU's/6.08x101b = 32 BTU /lb or s 26 F temperature increase.

o 4 steam (to cause spillover into the hot legs) from just residual heat 20 times more mass of hot metal would be required. (s 10 BTU /lb are needed to reach Tsat and another 533 BTU /lb to vaporize the water. This equals 543 BTU required to convert a pound of water to TV a pound of steam). 2.1.3 Conclusion The previous calculations and discussion basically show that it is impossible to produce steam in the upper head above a measured pressure of 1740 psig, highly improbable above 1620 psig, and highly unlikely above 1500 psig. ('ary f.ew. transients have even' pone below 1600 psig.) This. statement assumes no primary syi. tem breaks. 2.2 Analysis of Steam in the Candy Cane 2.2.1 Pressurizer cutsurce water mixico with hot lee Cow .n.:n. m.., .c.... m...y.. 5 m.... : :,... - n

....,.......*;..y.-.a...u.

.v.At th h period.of-minimum' pressure after a trip the hot. leg coolant temperature is s 550 F during 4-pump operation.and s 5700F during natural circulation. The respective hot leg flow 5 6 rates are s 70x10 and 2.5x10 lb/hr during this period. The pressurizer is outsurging s 650 F water at s 40000 lb/hr into the hot leg. Conservatively assuming only10% of the hot leg flow mixes with the pressurizer outsurge, the resulting maximum fluio temperature in the candy cane will be s 5510F for four pump operation and 5810F for natural circulation. ~,,. _

n p... V U The saturation pressures for these temperatures is well below 1500 psig. 2.2.2 Residual Heat Stored in Metal The residual heat stored in the hot leg metal is conservatively calculated to be s 1.89x10 BTU's. This calculation is based on (1) hot leg piping temperature remains at S 605'F for 60 seconds after the trip and (2) the residual heat is dumped instantaneously.into the hot leg flow volume at the RTD indicated temperature of s 550*F and 1600 psia. 3 Hot leg metal volume = 70.45 ft 3 4 Hot leg metal mass = 70.45 ft x 490,1b = 3.45 x 10 lbm ft" The aT driving function will be 605 555' or 50 AT. '.".6 , %_'{;. * *. _ 4 g ys* ;., e.';= r..,,.,*L,.*..,Q,*.;*,.,t,.. ; s,e. pe r..,*.,., ,.s,s o.. 5 .11 BTU x 50 F x 3.45x10, = 1.89x10 BTU's 1b~F The volume of water in the hot leg is s 197x10 lb Therefore: 1.89x10 BTU S10 STU/lb 4 197x10 lb This is equivalent to s 8 F in temperature addition. Therefore, the saturation pressures for 5510F* + 8cF = 559 F and 6 581oF + 8cF = 589 F are below 1500 psig. Per previous calculation 4 &M --,m. .-,,,-.-,-.--,,-.,.3--,-. < - ~ ,---.-,---.--,--,,r---m.,----

Column Wdcggp2 Flow Patp h *=i9* 2 I I s 1 h N ',4 i / ) FLENue? s s cover +++ + + +++ - \\ p Flow Path f 2 ,n c ,awg,e u l J, s s / e . j - t ? l t ' I

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// [ ?p//r // l /' 1 d. q-]J h;! - f s( /n' s s y ( -( d e=%, ~ m m = W A W A p ^ U U U U PPE R EN D FITTING ___ s _ _s l Euh . ~..

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1 M Upper Plenum Fion Patn k.2.1.2

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I 4 ,,1 t 3,-- y. j,s _.l ( j d_ column -75 weldment l l x J .? Cb L.'.a L..a N LIJ O; ( Ji j 1l ll u l' b_ h O o.Q' ( Top view of ut>er plenum r i cover for 17 FA Plants N Flow Path Flow Path ,s sx 9 a n u < m ; , n ,1 Ns /./,i\\ h h*ah ' / D b \\\\ /g j a 8 ~ m -p 7j p o _,ly;j..f,/'.....-r...\\. .; _....l.. . g..

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G A V U o 2.2.3 Conclusion Using the conservative assumptions on residur'. stored heat and hot leg - pressurizer outsurge mixing, the RC pressure which ] ~ would allow flashing is below.1500 psig. This low pressure 1 snoulu not ce reacned during normal transients. 2.3 Event Description For Mass / Volume Balance 2.3.1 Descriotion of Analysis A mass volume balance on the RC system before and after trips was analysed for five cases. These cases were chosen because of their abnormally low pressvi:or level, low RC pressure, or were referred to in reference 1. The method of analysis was basically: (1) To break the RC system down into 4 different temperature / pressure volumes (per figure 2.3.1)- ,- m.f, :-2 3,.g s g.- .,r,g. gg. .s7 g.,

e.. -l... ;.

r, temperature / pressure condition after a trip and (3) Determine how much water from the pressurizer is required ~ l to'maintaina' solid'sistem' conditions. (4) Subtract this volume of water from the initial pressurizer volume to get " level 1". (5) Calculate the amcunt of water that would have flashed in the pressu'izer due to the lower pressure condition. This is " level 2". (6) Finally, calculate the pressurizer level error per attachment 1. This is " level 3". (7) Subtract level 2 and 3 frem level 1 to get level 4. If " level 4" is less than the plant measured level (" level 5") then a viable explanation would be potential " steam pockets" l ._,-..w., ,-.-.-.n,.,_. ,..-m _-_,,-~.,-4.-.p

9 o. in the RCS. The results of this study show that the indicated pressurizer level (level 5) is less than or equal to the calculation of level 4 and therefore a steam pocket occurring in the RCS is highly unlikely. 2.3.2 Derivation of Ecuations Used The equation used for this analysis is: 1 1 1 I l V '4 + VE E + P VP11+YP22+VP33+VP44+V955" 1l 22 ' P + 4 5S 33 Where p = Coolant density at Time 0 1 p Coolant density at Time 1 = y5 = Volume of water in pressurizer at Time O V 5 = Volume of water in pressurizer at Time 1 1 Solving for V 5 V k. 53V)h.r.pj)..+.V(8.r. 2}..?;V(P3 8 3.'+.. V (8. ' P..).t Y P 3 7..,.2 w.. .: -..; 3 ......2.2 3 e :, .4,4 4 $. 5-a..- ..y 1 p 5 For raised loop plants. (TECO) '3} + (P 1 + 3.207 (level)] p V = 3 39(ap)) + 887( 42 4 5 5 p'S Eq.1 For lowered loop plants = 3800(ao)) + 887(49 )

  • 3I97 (AP ) + 277? (ap4) + [A) + 3.207 (level)] p5 V

2 3 1 P5 Eq.2 ,e---

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-gy e y-_,--,--,,,--,,,p-.--- --yg,,fw.,,w .,,,,,,,,,,mm7,wy%,.,-g-- ,,-e-,,,,,w .,,,,g ,.4 g,,,g.c.-, -..,,,%g-,--,-g.. -,--,y--,,q,

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c,; /,/ ~ / 1J *..,".* .. ) 1**,. 8.. \\,.,.. v,i y Raised looD Lowered Lcco 3 e 177 NSS (ft ) 177 NSS (ft3J l,,*. J ' = Cold Water Volume = 3739 3800 V = 1 s j=CoreWaterVolume= 887 887 .i Y 2 -. -.a .n. v: ~., p _ - ~~...~.;.... a....: Q}. h# j' = } foe Watir Volume '= .y.y :..,...,;... t,.: "~ ' ~ l' ' 3' 62 '- ' *~3197 0 L'.. ? w m,

Steam Generator Water Volume

2774 2774 V = 4 r,., Total External to Pressuri:er 10462 10.658 l

Pressurizer Water Volume

A + 3.207 X A + 3.207 X V l = 5 Level (inches) Level (inches) A = 263 (Davis Besse)~ A = 134 (Oconee) 1 = f(P) Teold) Pj= Pup 8 P2= Pup + 50 2 " # 8', T,y,) 8 3 = f(P, Thot) P3=Ptap 3 (P, T,y,) 4= P4=Ptap 4 PS=P 5 = f(M, Tsat) tap -. -. +.,.. - - e v-

w w

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--.wr -,g-n--- ,,-yg. ,.,.g v. ,c%,.--., ,.9- - - - -- -. _,, _._,g-w- 9

o e 2.3.3.1 Dav,s Besse 9/18/79 Trio d At 12:43 'on September 18, 1979, while operating at s 100% power a test was in progress on the main steam turbine Electro Hydraulic Control System (EHC) at Davis Besse 1. While l l transferring EHC pumps a low pressure signal in the EHC initiated a turbine trip. The Anticipatory Reactor Trip on Turbine Trip tc ed the reactor s 0.4 seconds after the turbine trip. The reactor trip caused the RC pressure RC T and average pressurizer level to decrease rapidly. The pressurizer level indication dropped off scala some 50 seconds after the reactor trip, remained down scale for some 50 seconds and then slowly increased. The RC pressure reached a minimum of about 1710 psig at 1 minute a-lter reactor trip and recovered. RC T reached average a minimum of 546"/ following the trip and stabilized at 550*F. . v.::. w e.;, ~.:;.s.; - '.y. .g. . g:g.. g>..,y.g.3-:. _..g..,7..- below. Reactor Power s 100% Rated Power RC" Tape'rature(7,y,h 58' R.C.S. Pressure: 2200 psig R.C.S. Flow: 4 pumps operating Pressurizer Level: 202 inches Tests in Progress: EHC Test The pertinent data during the transient is shown on figure 2.3.3.1. The data analysis is as follows. m-n .w. ,,s-y ,-3

Davis Besse 9/18/79 ~ ^" Reactor Trip Time (0) Initial Conditions Thot [ T = P cold tap " E * = Therefore p) = 46.24 p. 44.64 p 42.57 p = 4.60 4 p5" Time (1) = 42 sec T = 557 T = 550 p = 1740 psia hot cold tap Therefore p/ 46.40 = i '. 46.27 p 2 3 pC 45.97 = 3 1 46.23 pi 4. _;- 3, v::; c; y __ :;_. -l pgi-:.-l>g g..... <;:. j,..,..... -: : 5. g. -; y,;,;-::,, :. :

. j y,.:

Solving Equation 1 V ' = 415 ft 5 This corresponds to a calculated pressurizer level of 47.5 inches (level 1) The indic'ated pressuri:er level w'as '21.6"inctieT(15 vel'5'), 'The 'est.fdated 'pris'sud:e'r level temperature ecmpensatica error (per Attachment 1) is 1.;a inches (level 3). 3 3 I The initial mass of steam in the pressurizer is 650 'ft x 6.21 lb/ft or 1 3 3 4037 lb. The final sss of steam is 1214 ft x 4.37 lb/ft or 5304 lb. This additional 1267 lb (5304 - 4037) of steam will ecme frca s 10" cf water. (level 2). Therefore, level 1 - level 2 - level 3 = level 4 = 47.5-14-10 = 23.5". Level 4 is slightly greater than level 5; therefore, the probability of steam pockets is icw.

  • A 4 second hot and cold Icg RTD response time delay is incorporated.

l Note: Where possible, data used in these analyses is based on ec cuter print cut. The graphs arc representatic.s of this dats. J MM M"

O wie.2331 I m g a= t. t u. a o e o en a co.- c.c u os ao ve m -/m% v coe c e-c i e - o ee 6 o .D CG o eN k a GI 6w a% b by % are e o e 8.as-6 e cm>- a a o o g: \\ EeI6m-a oeo aUW 3oe 5 w W W o '. ** 3g6 g w - o- :.- 6 eaN usz \\ a rw w a \\ cn = ** \\ w cae 53C \\a w \\ --o ama Eo \\ o cs e.- 3-

c 5

m ~d eu a ^ \\ m-o .a o ema \\ .as o a > u.a a ~ n. 1 y u- => n = n >A an-c h "g a \\ ' *3 u v n. 1 m w I w g, g e 3 ; .m, N g op g e4 U \\ e a gr* 7

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3. '3HA1YH3dM31 I

f f I I I f I 8 8 8 = = = e. = = n ~ o e n n n SISd '3HOSS3Hd T H t e f f I I f I t e o. ..e .o ,o .o .o o e o e a n ~ N a= (S3H3NI) 13A31 H321HOSS3Hd ,n.-- -,,,-_,-,....w..,..._.,,.,,,.n._,.,n,,,,e.,, ,,.--,,-.m..-,-,--.,_m-___.,__..,-,-,,.n_.,-,-,,.,---w.,,,,,p_ ,,--,,,m-..,w..m..,,-,-a,-w,, -,,,wwm-,-,,,,,,,,-

4 4 2.3.3.2 Davis Besse 9/15/79 Trip Figure 2.'3.3.2 shows the RC pressure during this trip. Since RC minimum pressure was 1848 psig no steam production was possible per discussion in Section 2.2. . 9.. .t:.,.,.,.,..,....:.. ~. .w I O WW V '*WWres Y

s. 1 i a 1 ECO 9/15/79 TRIP i t i 21.840 . il i r-

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21.420 ~ y '.j = n 21.000 'i l, l ~ a 1 I M v m 20.580 w l A y l l N

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z ) 20.160 w m 3 I j wa w CC 19.740 i O m O J, 19.320 .n .r. 3 j 18.900 .h. '.;. u .n i 1 18.480 i ~ i i i i n 8 I P l 49.410 49.815 50.720 50.625 51.030 51.435 51.840 52.245 52.650 f 2 Tise(seconds)(X10) t

r h

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  • N.

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Q Q 2.3.3.3 Daivs Besse 9/26/79 Trip At 20:56:33 hours on September 26, 1979, a high turbine throttle pressure limit alarm was received. The throttle pressure limiter is used to prot'ect against an excessive decrease of steam pressure when steam generation of the NSS falls below the steam demand of the turbine. It acts directly to close the high pressure turbine control valves in an effort to naintain steam header pressure. Rapid closure of the control valves caused a mismatch between heat generation and heat removal with a resultant increase in Reactor Coolant System (RCS) temperature and pressure. Seven seconds later (at 20:56:40) the reactor tripped on RCS high pressure and was followed by a turbine trip. RCS temperature and pressure then dropped and the Integrated " Control System (ICS) reduced the feedwater flow abruptly. w..... ,..s.+.:.t.- - teacto d oolant System;Tave decreased to-548 5 F.approximately. - r,. _, 57 seconds after the trip. The pressurizer level indication dropped off scale at 21:57:21 and remained below the indication range for,pproximatety 21 seconds. The initial conditions prior to the trip are listed belcw. l l Reactor Power: $ 100% Rated Power R. C. Temperature (T,y,): 5829 R. C. S. Flow: 4 pumps operating Pressurizer Level: 200 inches The pertinent data during the transient for this anal' sis is shown on figure 2.3.3.3. The data analysis is as follows. v m6 wwy--- my w--= -v9w y-- p.--w.- - - - -- - - -. g-% g- +- w-g.-, y,

%,-w.7--w gw..%,--w,c.9

,-y, wg +ww--9,- ,..9-p-9g.,--q ,ogyvg y-, w,,gm9_.9 9,---, wy goc pp-

Davis Sesse 9/26/79 Reactor Trip Time (0) Initial Conditions T T = P 2215 psia hat = = eold tap Therefore p) = 46.18 44.59 P = 2 P3= 42.61 44.56 p = 37.35 p = Time (1) = 50 sec 552 548 T 1700 psia = p T = = hr' cold tap Therefore p/ 46.62 46.53 p = 2 3 P 3 1 ~ p' 46.35 = s ' f.;. f..,'.. =. b *. ~ 1 .' 41.00 --,h *.',,..L ,6 ...'s.. .' ta,4 wt.' h ' E;-ll '* *.) *tt?*.'== '.nkk *s ?. " r . W* ; * = S, i,1- . c. 'r pj. = 3 243 ft Solving Equation 1 y = 5 ~- . This. corresponds..t a cal.cula.ted. pressurizer level. of 25.4 inches. (ievel.1).. The indicated pressuri:er level was 0 inches (level 5), The estimated pressurizer level temperature compensation error (per Attachment 1) is 15" inches (level 3). 3 3 The initial mass of steam in the pressurizer is 648 'ft x 6.21 lb/ft or 3 4030 lb. The final mass o' steam is W34 ft x 4.23 lb/ft or 5219 lb. This additional 1190 lb ( 5219 - 4030) of steam will come from a, 9" of water. (level 2) l Therefore, level 1 - level 2 - level 3 = level 4 = 26.4 9 = 2 4". Level 4 is slightly greater than level 5; therefore, the probability of steam pockets is low.

  • A 4 second hot and cold leg RTD response time delay is incorporated.

Nota: L'here parsible, data 'ised in these analyses is based on ccmputer print out. The graphs are representations of t.11s data. - - ~ - g g v -v y -m w-a g w ym-mp---*'a ,ia. -.y -3%,-- g==mg y w --

    • e

+,w--Wmew-- w' w-- r-

,.' TECO 9/26/79 TRIP I h e i l 3 2300 { 1. 240 2200 200 2l00 j 3; l i.s n E 590 u, w I D M .i ^ 5 160 - m 2000 580 i 13 ~ J t a gg

  • 4 W

570 -) ^ e' na. u, t m N 120 1900 W 560 1 I.TAVE ~ ac 550 o i RC PRESS. 80 540 1800 m l. g,. j n. 5 .,y 1 r. 40 1700 520 t PPZR LEVEL s l . e 5 0 1600 'I O.0 l'.) 2.0 line altg.r trip (minutes) S j{  ? 1 .l.

-~ - - Q O 2.3.3.4 Oconee 1 10/8/79 Trio On October 8,1979, Oconee 1 was operating at 100% FP with on line Reactor Protection Tests in progress. At 13:27:35, Oconee 1 experienced a reactor trip attributed to RPS activation of pressure / temperature channels A. C, and D. Control Rod Drive Breakers CB 3 and 4 were tripped for the test. An additional breaker for Group 5 was opened which caused Group 5 to drop. This resulted in a P/T Trip. The initial conditions prior to the trip are listed below. Reactor Power 100% RCS TAVE 579'F Pressurizer level 221 inches The pertinent data during the transient for this analysis . ;..... a., 2,- s.e,.. n.- -are shown on.: figure 2.3. 3.4... The data analysis is. as. foilows......... v. ~ r. s. s. , - -, - + -. y--s*-e. 3 e3

p-%-,-

,-+ - - - -. ,,mg-gp.i- ,e-=g.g%pmg w .g.y#--,-p-e.yw- .e--e-g -y= ,en--w----se+--p-eg g-w-gy-p,, q.- y93++ye+-+jp.-- r

1' O Coidreiko[0NpTrip O o l Time (0) Initial Conditions T = 602 F T 558 p 2150 psia = Mt cold tap d 46.26 lb/ft l Therefore o = j 3 44.75 lb/ft p = 2 3 p = 42.87 lb/ft 3 44.70 lb/ft p = 4 U.69 N ft pg= Time (1)=60 seconds 553 550 1750 psia T = T p = hot cold tap 3 Therefore p' 46.55 lb/ft = 3 p/ 46.47 lb/ft = 2 3 3 P 46.40 lb/ft ~ = 3 1 3 p' 46.51 lb/ft = 4. .e .;.. p l.. :. .g.. 3,:, ..u,.;. ...g. f 3 Solving Equation i V 315 ft = 5 This corresponds to a calculated pressurizer level of!624 inch'es. (levei !) Thd# ndi' cati'd'presstri:er 1&ie1~ Uas 31.0'inchef(leva! ;); T.% estfIdated pressurhe? i level temperature : mpensation error (per Attachment 1) is 15.0 inches (lavel 3). 3 The initial mass of steam in the pressurizer is 709 ft x 5.97 lb/ft or 3 4236 lb. The final mass of steam is 1283 ft x 4.4 lb/ft or 5652 lb. This additional 1416 lb (5652 - 4236) of stecm will come from s 10' of water (level 2). Therefore, level 1 - level 2 - level 3 = level 4 = 56-15-10 31". t.evel 4 is equal to level 5; therefore, the probability of steam pockets is low. l l A 4 second hot anc cold leg p.TD response time delay is incorporated. Note: !!here possible, data used in these analyses is based on computer print out. The graphs are representaticas of this data. l I l w*w e -,,.+>--.9..,e,_-_.,, p-aas--+-+-*e- -w +-**-w-e-g-wei. ywg e.- g. p-+--eme<. ---**--M--w-

  • --*e---

-w9="e &"T e---- -=='t-- M --N*

...g. o o t-g o. .O >-e a Z w e / = c u w w w w a a = = s. an t-- =

== e og w en m w sn w a ~ = w = o .as w = m M = = m w a m = s 5 m y a w O. = = c= [,,,...,3 j#p# g, 7.* - ty e c, g s % + y .., g st s,. s c ,e ,J,. .,.. % 4, s 3 s g* J ' / Co o ~ N ed =" h e v i i f ii i .= d = .o o o a n 4 e ~ m = a .o .o o e. n SISd '320SS3Hd 38 e i i i e i e i f c3 C3 o o C3 o

== m ~

  • =

= = = o

3. '3HAIVB3d31 3H I

f I f I f I f i f I o cm o o o cm M o o P. f*) .=se N .= = (53H3NI) 13A31 H321HaSS3Hd e G ~ h e------- ,---------,.-..,,---w_,-.,.--,_,_ -_,_x,---,-,-w,m, -.,-,,wy ..,,.,,.me,..-,vw,m,e-%.,--,-,--.-,r-- -,..m, v--,

o c .] o 2.3.3.5 Oconee II 1/4/74 Tric Figure 2.3.3.6 shows the vertinent transient results of this trip. During this transient the minimum Tap pressure attained was s 1900 psig. Therefore, a 1940 psia upper vessel pressure would require a 631 F t::moerature to produce net steam. This temperature was never attained 4t any position or time during this transient. While it is given that operator action (or plant responses) could have been better, the ' steam production possibility did not exist. Furthermore, a review of the sequence of events of th1s t-ip indicate that the severe level and pressure changes at the begirning and later in the transient were primarily due to a temporary loss of ICS power. . p. ..+...,. c.y. ;w.......a

.v..,...m
1... :...,,...e y,.

..s : I 1 i l _.... 1, I..

i Oconee II 600 1-4-79' Trip t 580 HOT LEG TEMPERATURE j i ,^ s5 560 T~~~# k' ~ i \\ ) \\ 6 l l ~ g 2500 540 \\ . f. SIGNAL WENT GUT OF RANGE i \\ \\ / n 2400 o" 520 - REACTORC00LAAT 1 '/1 / 200 1 SYSTEN PRESSURE I I [ COLD LEG TEMPERATURE g [ I G 160 - g 2300

  • ' 500 W

G = l.l /~p.A fl 140 D .I /*./ h*/ O .v s g \\' j CI 120 - 5 2200 480 g = = .f i ' 100 E PRESSURIZER LEVEL [N e iri o I g 80 - sd 2l00 460 \\ t g 60 [ (; ,/ 'N' s / E P s s % -. / j 40 2000 440 t ~ 20 i ~ 420 lj 0 l900 i i I I I I I I -5 Of 5. 10 15 20 25 30 35 40 45 l ,,$ Time after trip (minutes) 1 i

~--

l

3.0 Conclusion During any B&W reactor trip where a primary system break does not occur or the pressurizer does not empty and HPI is operable - net steam production in the primary system is h,g 11y improbable and generation of a volume of steam required to block natural circulation is virtually impossible. 4.0 References 1. 11/15/79 phone conversation between B&W and NRC,

Subject:

Potential Voiding in-the RCS During-Anticippted Transients - :.< o ,<.a. c % .s....r......- ..n.v ;...;. s. v. ;:.... s s...,, <.. ..... y ; o - y..,,,.s.. .s ;.;,...,. ~ t~* ,... -~;:,. . s- , e e e e e se, gg g

  • e e

4 ,e g 1 m_ .g, g..m m = ,U Yp I --w _. M N - ~.. -

o o ATTACHMENT 1 Pressurizer Level Error The pressurizer level indication is based on the pounds of mass the pressure taps measure and then distributed to varying amounts of water and steam at the saturation conditions at the pressurizer pressure. Since pressure is not measured in the pressurizer (only temperature) pressure is assumed from the temperature measurement being at Tsat

  • During fast depressurizations the RTD in the pressurizer does not respond efficiently, especially in cases where the RTD is exposed to steam enviroment. The heat transfer time constant can vary from 30 to 180 seconds. Since most depressurization after trips occur in 30 to 50 seconds the RTD will not respond effectively at the low level point.

Figures Al and A2 show potential level errors for 400" and 320" pressurizers. Figures A3 and A4 show what typical errors can be expected as a function -#f- },Sf'Micabad leN1 an'disystem.'pressdre" ThesE curyn Ea'nh used. to 'coWecK [.~#l'*' the indicated level if system pressure (i.e. Tsat) and pressurizer RTO temperature are known. The difference between RTD temperature and 1 'caTchTakedT - iBased'on P t) w'111-givs the level

  • error.

e. sat sa In a typical trip 9 Oconee 1/1/77 the system pressure went frca 2150 to s 1800 psig in 27 seconds. The alarns printer monitored a 648 pressurizer temperature at 1000 psig instead of the expected 621 F. At the indicated level of 70" the real level was s 78" (per figure A-3). e. v-r,,v-,.-,w--w v

e-e.-,-

,-,-e,--m--w,- w,-- +---+---*,--w--y ,--wye,---.w- ,--,----r-wv.+,.,-g*,--w +-*, ,--www we,-. -v ,en.

O tyure e-i A A V V m

A

~ 4 T k M _.I ^g g s _?- n r,, --.. M e 4=_4 J

== O =7 u Y 5

== u % ^ w JC m =x-4~"""* \\ 5 1 K _.s__ M M t uw =,m Y 4 5 = g, T o M' ~ ,h ,~, C

  • E T f_ ~

7 T W l.

m

.Y ^ -2 at - -) .- 1, ~ g a g "#'~,# e

  • %w if

- u :s e e.J W n -M as Y _s Et. . i s AK ~. 1 f' __1. _ i __.. h_; _.; e W -_s

7. _

h h ^ "_ _ -O a t, ,gs ~ ~ m i 'O 5* W - (-_ gQM J t --t c) A ~ & --4 e ~~ W4 g 1.,.h W l^ m'. "3 m 7 _. w-.. .. w')s _. c '._ -. m % M f 6 %. C.".,A W y _" - + "_- a. & l., e-F ee -. .__.___-e..-._. E =.'El - -3 1 6 9 1--- v --- t-M 2 7: ~. -g.. s = a s._ z _ m = .$ __..-.m_':2.T kO r_~A. W - - ~' _. ? 2 = T S'~~~~ ~31 ~ ? I~~~ .em, o m, e .o ee m ~ 4 8 P 7-

Figure A-2 s ~ m- !.- D s-P -i--__ -.x. _=. i n r W J W ._L. m_ Ml 's ,~ c.5p, ^~,> w ^ ~ n y -i- - v M 4, te z T .J. 4_% W-L Iw E' Tw 1 ~ A K O-._- m., - G, r g n ~~ _ 2 g-Ik ~h = s 1 3 m mm Mm" W,i g-. ; - x - =,m 5 y t. L' 2 = =8 A. s* .-I. r . - t= . u_;,.. 7 3 a .o _e

  • 2 m

~,: C-W C.s r' x,_- m p. +-u m1 m _ _- zu* w 4 ^ 4%_ .._._g - ~ [ 44, .=L .-g v m -r_3 ^1 =- mam m___'- .O 1 i 'g

.=--

'"'" M M ( 9_ = -2 r. 1 M _e 1 ^ 9 ~~ s_s y A .+_-4 p_ ~ a .V_.) p.- -4_ g.__ _e eg, -9 '__4 m me. J. - JI. _. [ CiL _ ~ li ' ----- - e 3-6~ b '-5 A ._-._ ? a-___.* b_--. 4 __... %..r-- : _7 - ----_.g.. -- : r** C_.--7 _.......c- - -- r.-E n-9es __U a^ . 5.. t [ I $ 2 s y *.3) d Q d Q A f7 WW M tr.9g) SS4u).:i:t 1 ./ W e= 4.s -g.> I 1 1

3 Figure A-3 ,V t ___ = _ - - _ - = - - - - - _. = = - -.__- _-.= _. =_.__ _ - -..-.. w - - .a_-+_____ =_ ~ = . _. _m =. _ _ -. - -. _ = - = _ = --__-._- _ .y= - - =. a --a w_. Q -__ 3 ^ ._ e-a ^- r -._.- __. ~ _ .=.= -_=.- e ~ - - - _ - u _~ --=E_-.- - - , = = _ = r --. =_ _g ~ e- _ _ =

u..

.s al - -.:s 3__-. =Y-5 -- .-~ .1..__- O - -_m, a .m, w =' w _.=. L + _1] -m v- = % __ _ u._ .s._,__. __y

c e

= -r__- __ _- s =_- r- .. n v + ,s. _.y -1.; =,a m 5 i _f ~-El -=, _=- ^ e m._. er: m < __n ._ _ m_ -. 3 n-,'. s, s . t.=_s. %__ -. 3,,__w..__.. --.- g y _ mg _.7-- s-a . _.L _ _ = z w - = - - _. _. ..-- 9 E - - - - - - =---2%. - 4 g .._-W - U = _-.__h, ~ .t. .= +s _t-- ? _A x ND r--.- _ww = -. - - - - - g t E_ V .] h mg + - s i v M. ( e_. e M { w g er==: - ~ Q .4 O Z y r WM "._ _.- '~ [ .T.D N f 6. 9 .n!, J,&y-_ gry 22_37_Q{=ane. rW? > _ ~ apr 'b e 8 9-W**

  • M***u=*-'*T N--
  • N 6*a P-e

-'e W=,,

.3 9 M.'e u.4."

  • 7--_

4W %wespa* esn% 9,+ F ~# #", -r e"

t ~ Figurt A-4 _ _ _ _. {-- - * - 5. -i I%."'*,~._'~_~- -6 ?N 4 .g,,,gg . c t -- 4 ., g ,e~'I --J ;.. -. -. ...y__ N f _-m_- _y; 7C. .~~'%,.* =y-3- 9___ - _. _ - _..eN-.em-__ ~, ..m s .-- = }. -r4 -,- 2 -7E.~. M p -_s,e. w. C b - =- s~ m W c_: ' l' %.~---* 2 M _t ]- -m -e.,__ A h} -mm- .r \\ O e L._..- _ Q q_. a.7 _--4.J.b. o ~".? o a ~ N "Z._ . 7 ..b E ..__.___,__u___

  • ~~~"C p-

.m -

  • 6 2~

[. P-w _w ,r. M 2 1 -v wa. ee -._s w*

  • 4 ~~ g~ *,

d 2 { _._ _ ~ 3_% ~ ~f* - " ",~ '4f*T.*._TP- -. C g*rn m ' ^ '~~~~~L . __,2 -_ _2 3 c., _-j __..__- v.- z e .n __. f i..y _h

== q y_- o. 5- ~ -- S . s - e..-- nm

== som m. ,.= x s y. s ___q .3 ,. m. m w _.. d - _ g-f_- - e _h g 7 ( g ee-. ---__.4._._.-. 4_ -_4__ _ -. _ _ - _ -. _, - - _ _.. _ _ _. _. ~ A q W N + y W-4 J ~ ^ ? = g r-a w l_~f____.__ _ _ m _q_ r-L_ g e- _~ y ,-_e eee _%..---.-e e w-- r .r ww w -w._- + ,mir-yv.- wwy yr ww'a ~-

,[, cnr T,, ' 1.,f .0*n

  • rr a

? t ~ g:., Response to Oue.stion 1A_ ~ ~. ' L 1 -v Provide a benchmark analysis of sequential auxiliary feedwater flow to the steam pncrators following a loss of main feedwater. This analysis was provided in a letter from J. Taylor (S&W) to R. Mattson (flRC) dated June 15,1979. However, in this analysis the TRAP-2 code with a 6 node i steam generator model was utilized. All small break analyses presented to the ftRC have been perfomed using the CRAFT-2 code with a 3 node steam generator model. We require a benchmark analysis for sequential auxiliary feedwater ficw also be perfomed using CRAFT-2 with a 3 node steam generator representation. by The' Babcock.& Wilcox company Nuclear Power Generation Division ~ '.. ;,. ; :...;....,, ' : ;.1...; ;... : ;,,. .;:~ .y.;a... ;_ ;.., ~,. - ....,,,y.,l,,y.,y._,.y.

' M-[... m...y, ty n ~.,: b.;.Q r,..,.,s.',

,,.4 ;g,.;. 4 !.b n s.: ',.;. r,

s. g.g L,. : y...;. :

.y k.~.;.,s., 3; ;. .,..'P " ;&. y.. :,8.,. *p '. '* 7;. 3.-y-- t,...,j. j ;,.s.;s.:..g y..(:j ;.pe.;.,. 4,i.c. y..i 4,.. ip',..,.f.,-: p.,..r.,,.,,.h,. 3,,;,.,.., g,, 1 l t l 1

c. _ :-..

__ n

V*~' ~ /'.L.'.l m p' / / .2 ~ i *- ) .. I -i T O ;^.' '~ 1 E.7k00UCT!0lI t

t. i e

n- .D. y. V ..y ; s,

y. y II LITE EVEllT DESCRIPTIO(

~. III M2i005 ~ s. gy'. KESULTS ~' Y CONCLUSID:15 5 FIGURE 1 CRAFT-2 f;0DItiG DIAGE*1 FOP. St!ALL BREXES n FIGURE 2 STARTUP FEED!?ATER FLON ~ STEM 1 GEtiERdi0R LIQUID LEVEL (TE!!PERATURE ADJUSTE ~ FIGURE 3 FIGURE 4 STEAM GEtiERATOR SEC07:0ARY SIDE PRESSURE u.. . [ ;. TIGURE 5 PRD!ARY A LOOP TEMPERATURE FIGURE 6 . FRIl!ARY B LOOP TEfPERATURE o 3:..x FIGUR.E. 7.q,., .' ' PRESSURI2ER' LEVEL :'.,- L...:::.:.;..... i.

a. m:..'...<.

i' ,.......,, 4.:.s...i+A. n.h.a n >;W :.m:. u.a.>-3nll.x q.4:.:,d ys.. % + .-; o .,v.,.;.r.:,,- .. FIGURE 8 REACTOR VESSEL PRESSURE' ~ +.....g,,;.y;.;:,.... ;..h.::e.. p.g.2 ;. +. r.c:,... ;s:a: :., za. : - g,oe.w: v. & m:c.l.o 9: ::,s ~:...:p..?, - O O 9 e S l O t e l 1-4-80 ""-"y x-, ___r-p, .T =----sa ..7

  • I #F.l.'e*

M

  • U

'9 I e a,- ,g,-,, _, ' -,~_, ,,,m. ~,-eom-.

I. It!TRODt*CTIO!! m This report p, resents an analysis of sequential auxiliary feedwater (AW) flow / / to the once through seca= generators for a loss of eain feedvater transient. The CRAFT 2 codel and the s=all break model described in reference 2 have been used in the study. The calculated results have been compared to a loss of offsite power startup test data obtained from the Florida Power Corporation's Crystal River 3 Unit in which an i= balance in the auxiliary feedwater flows between the two operating loops resulted in an i= balance in the pri=ary loop response. This transient tests several features of the co=puter si=ulation, including conditions cf asy==etric loop te=peratures, an almost dry generator to feed auxiliary feed-water into, lors of RC pumps, and establishment of natural circulation. In =any cases the absolute validity of the boundary conditi:ns and test data were ques-tionable, and esti=ates had to be used. However, this analysis does show that the data trenda can be predicted by a 3 node CRAFT 2 SG representation. II. SITE EVE!!T DESCRIPTIO!! The Crystal River 3 Unit is a 2452 W e, 177-FA BtM reactor with a lowered-loop configuration. On April 23, 1977, a loss of offsite power test was perfor=ed. T5 is test was initiated frc= approxi=ately 15% full power operation. The secon-i ...dary. liquid icvels were ap2roxi=ately 2, feet and was sufficient to re=ove the s.. .de..,e s se n,t,iay..,, <.. ~...ste,ady-state. operation.:pr_ig,....to, test. ini,tiationl .s v; u;.,,-Pof' pr; and prdk$.. , y; . q,, The test was initiated by tripping the reactor, the reactor coolant pu=p, and feedvater pu=p pcuer scurces. The core power then dropped to the decay heat level and rn s...). :n. ;, as. the pri=.ary: coolant pu=ps, coast.ed down, the pri=ary flow decayed to _.y y.; ..,.. a..--a. A :. 9. y...:j,s..;,,,... n,. n,.;. n.,; ..,,s.,p n: naturs1 circul' tion level. One di~esel generator was started to provide" pot.*er for s the pressuri::er heaters, one =akeup pu=p, and other necessary services of secon-dary i=portance to this analysis. The =ain feedwater flew coasted down resulting in both stea= generators eventually drying out until the auxiliary feedwater flow beca=e sufficient to start filling the A loop stea= generator secondary at about two =inutes into the transient. The B loop stea= generator re=sined dryed out until twelve to fourteen =inutes into the transient when the A loop reached nor=al operating level and the feed-water flow was diverted to the B loop. The i= balance in the feedwater flows, cad hence levels, resulted in a corresponding i= balance in the pri=ary syste= re-sponse including the decay heat re= oval, the hot and cold icg te=peratures and Y - = = _. - ____-___.o_-,,,.,-=,.x_ J ~., r

,G : ) .r ,m < ~ c . I I ~,\\ '.n 2 l> wars (utdtocvalustoths h.)te] C e= ( .) g

)

f1p s batvern tha two loops. The trcnaiint resu y ability of ti.: 1 node'CMFT2 steam generator codel used in czall breck cvalua-tiens to calchlate the effect of the feedwater transient. III. METHODS A. CRAFT Input Model The input model developed for this calculation was based on the ses11 break model used for licensing.2 The schematic of the flow path nodalization is shown in Figure 1. The initial system conditions were defined based on the available aca-cur'ed data which were required to represent this test. The model was set up to provide a steady-state calculation until two seconds into the transient when the reactor, reactor coolant pumps, and main feedwater pumps were tripped initiating the transient calculation. 5. Initial Conditions The initial = ass flow was assumsd to be identical to the full power operation value. The measured het and cold leg temperatures were then used to determine a consistent core power to provide the initial steady-state operating conditions. This resulted in an initial power of 19% of full power operation versus the 15 power defined in the su= mary test report. Hand calculations, using the 15% core

.;3.; power.and.the measured. hot. And.colii;,te=p.cratures,. ;resulted in~ a; mass -flow 7 con-

.,.G.,; Thh actsal inass cflou is /Nd@ ?'* :'6 -didefably'belowtthaTnq'ud' red ts.balancs th pomp p6wer. I i believed to have been only 1 or 2% less than full power flow. The pressure dis-tribution around the system was revised, because of the new hot and cold leg s/,ig. Aemperatures, to paintain, the. loss'. coeffhience -defined by the< referenced;mofele J., J.,.y.c c The liquid levels in the pressurizer and steam generator secondary were changed to reflect the measured data. C. Boundarv Conditions The makeup pump flow was codeled by defining the pressure flow characteristic curve for normal operation with the recirculation line open. The makeup pump was actuated when the pressurizer level dropped to 30" below the initial liquid level value. The makeup pump flow was equally distributed between the two cold leg pu=p discharge nodes as shown in Figure 1. The feedwater flows were defined by the test data and are given in Figure 2. An auxiliary feedwater enthalpy of 58 btu /lbm, which is the nominal enthalpy of the system,was used. x--: r-3

86-1101006-OL ~ The safety relief valves w ue set to 1030 psia to model the effect of tha turbine bypass valves, which are fully open at 1030 psia. The safety relief flow is the only allowance made in the model for secam flow. The heat transfer to the secondary was assumed to be to the mixture in the lower portion of the steam generator and the fraction which may havo been depocited in 'the steam region was assumed to be negligible. A preliminary short-term transient evaluation demonstrated the need to define the heat transfer multiplier based on the steam senerator secondary levels. Consequently, the final model contained c heat transfe c:ultiplier as a function of time based on the measured secondary levels. IV. RESULTS This section presents a comparison of the CRAFT 2 analysis to the data taken for the first 20 minutes of the CR-3 loss of offsite power test. As will be shown, come of the data utilized in the evaluation is questionable and greatly influence the transient response. However, even with the uncertainties in the measured data, the CRAFT 2 code is shown to adequately calculate the RCS behavior. A. Secondary Resoonse 3 o.,, Figure. 3..shows.the~ secondary;s,ide SG levels.dur.ing.the test..The cest data shows wWi theti;..following. the-lossi of > main rfeedwatere the initial: level.tirr heth c. steam je,n., ' ;,e,../: Grators decreases. At approximate 1y'1 minute into the transient, the auxiliary feedwater system initiates, as shown in F.igure 2, and preferentially feeds the .. E. ;. A lcoP.atsaa,ge,neratoc 9 _Ihusg h.11guiql.Acvgl,in f4..A. increases g,A3,12 3miquqes

3. ;,., 7 t

the liquid level in SG A stabilizes because it has reached its control point. At that time, the feedwater flow 1.s diverted to SG B and its level increases. The CRAFT 2 code calculated results shows reasonable agreement with the SG A level during the first 12 minutes. After this time, however, the CRAFT 2 calculation continues to increase the SG 1evel while the data shows a level stabilization cfter this time. This difference is probably due to an overesti=ation of the cuxiliary feedvater flow to SG A after this time. The auxiliary feedwater flow, cs indicated in Figure 2, is very stable and at a relatively high flowrate after 12 minutes. Examining other data, such as the A loop hot and cold leg tempera-tures, does not support a high auxiliary feedwater flowrate. In light of the cbility of the CRAFT 2 code to reasonably predict the SG response up to 12 min-utes and the inferences obtained from other data, the flowrate given in Figure 2 citer 12 =inutes is believed to be in error. b =

~ .,3 .c,.,, r. 4 9 ca g i y vI'.\\ n?- 9.g Th3 SC B liquid icval rstponsa is stn: rally cvarpredicted by th2 CRAFT celcula-tion. This tgain is believed to be caused by an overestimation of the auxiliary feedvater flowrate to SG B, especially between 3 and 9 minutes. Figure 2 shows the auxiliary feedwater flow to be very low over this time period and very stable. This may be due to an initial instrumentation offset and no feeduater may have been delivered to the steam generator.in this period. Once a sustained auxiliary feedwater flow is established to the SG, the CRAFT calculated level increases are in reasonable agreement with the data. ~ Figure 4 shows the SC secondary side pressure respon'se during the transient. CRAFT 2 predicts the pressure response for the A loop SG reasonably. Between 4 and 6 minutes, the calculated SG pressure increases above the data. Over this time period, it is believed that the measured auxiliary feedwater flows are low. This' conclusion is consistent with the level comparison shown in Figure 3. For the remainder of the transient, the prediction is higher than the measured SG pressure. The secondary side pressure for SG B was generally underestimated throughout the transient. This is caused by condensation of the steam within the SG due to the cxcess auxiliary feedwater flow utilized in the calculation. .B. Priearv.Sestem Rescense .,y...,.. m. I [^W tigursI5 'sh6ws th'e< k I'oo'). c'empe'rsturs ' resp 6n'ss"dbring th' rtist.r ;Tlie'. hob ' eg :'tept "-N:~? e-e l perature co= pares well with the transient data until 13 =inutes. After this ti=c, the CRAFT 2 calculation continues to show a decrease in the hot leg te=perature 9 e '.du'e4o.: the contipued: feesling"of.3 the=A Loop SG.Pc.Jhe'dat.a'shows.,a. flastening;of.-~-. w.;. w.- the hot leg temperature due to the control of the SG 1evel. This supports the l b211ef that the auxiliary feedwater flows after 12 minutes is lower than the l values indicated by Figure 2. The calculated a loop cold leg te=perature response is censistent with the data i trend, but gencrclly overpredicts the data after 4 minutes. This is caused by the overprediction of the SG A secondary pressure discussed previously. The B loop temperature response is shown in Figure 6. Due to the overprediction in the B loop SG 1evel and underprediction in the SG pressure, the hot leg te=- paratures are underpredicted.

/ i r= 4 ,v ( d kj./.(/ .q.g < FigurGs 7 cnd 8 show the pressuri ar icvel and syctem pr:ssura corpsrisen. Hand calculations which were performed indicate that these parameters are not consis-tent. Examining these figures, it is seen that the calculated pressurizer icvel response is in good agreement out to approxinately 12 minutes. After 12 minutes, the continued overcooling of the A loop, due to the overecti=: tion of feedwater flow, results in an underestimation of the pressurizer level. The pressure response shown in Figure 8 shows that the CMFT2 calculation under-predicts the data. However, as mentioned previously, this is not unexpected as the system pressure and pressurizer level are not consistent. V. CONCI.USION A sequential auxiliary feedwater flow transient has been benchmarked in thic analysis using the CMFT2 code with the 3 node SG model used in small break cvaluations. The site data trends were reasonably reproduced by the code. In many cases the validity of test bcundary conditions were questionable and er.ti-mates of the test data were used. However, the results provide assurance that the CMFT2 code is capable of reasonably predicting the primary system behavior indicated by the test if the bo.:ndary conditions were well defined. Thus, this ctudy has de=enstrated that, in spite of the si=plicity of the CMFT2 stea= j generator codel, the CMFT2 code can esti=ste, with reasonable accursey, a tran-t;ieni hishly. dependent!- en thi ' stead generat'or! 'Thus; the ability 'os the.small- . r.,... -c. . *.. r,: - w u. w. ;

a..w,.

cv-w_sn ; break....d~ei,. n s..:,y. to calculate'th....w'ffec;t of... :.:. steam. generator h+ eat re= oval daring a :.~ %, p.. m co ee small break transient is reasonably assured. %.e. h,4...;;s -&.. W k.t :r..ta.~.:. &.v..v ?,*-.44.wp..:, ;%.+.+v.:.W-:m,.'.rjer 4lr,93.; e,;M..p - 8

m. A g ,7 t ^ 'gj h. (fm W 93 _REFE_RENCES 1 R.A. Hedrick, J'.J. Cudlin, and R.C. Foltz, " CRAFT 2 Fortran Progrars for Digital Simulation of a Multinode Reactor Plant During Loss-of-Coolant," BAW-10092, Rev. 2, Babcock & Wilcox, April 1975. 2 Latter J.H. Taylor (317) to S.A. Varga (NRC), July 18, 1978. e O 4 d. g, g*.. 8,- -., *, ..g .,,w.,y...), ,eei, e 6, g. #, 7f. ,,,s.4 +e .b.". , it' W.b ',N-i.'; < h i*~ i'.i".U. * '* W.Wr C's.wd';.,U-Q."' r N. 3 pr "q.'i4.N W s.A of. [cq9 '.* W..b i f.e. d.h. h..' ;.**.' *.r s.";, i.; hy;[ e,., e .p,

y. 9.pu h;Aws& v.~....
.s.q n y.. a i...;.w &.vqw.p;g.m..,;.y..q.;..._.*:.y-r e -4..:.g r;

,2.....r p..,.s.g i i l . _ = _ _ _ =. ~- e,-,,-- .-,.~e,_n.--.,

$0fm i e.o r.......... a..t.., 6 6 he:, en II 9 O E L. G a a < < < eleL e O a g G a rn.... e.L o,. 2..., .,u....on O S O 1,, H g e a n.,...,,,.:: t 01 Ol 10 !, g g. u.. ..,n. 3.t.. . u.1...< u t,, G g g u.n .u..,..... C ,...,,,r ..t..,m ,.r. g .t,,ut.,, n.r....t., i,.., i,,,t. . w... -. ,,,, c,3,,,,.., .,,.,,,,,,.,u .e... w m. n,. w,.. w. v...: + x m,,, +.. e.c.w:...;. .x.w..&.:..,s:,pi w +.,.. w.,s.. m.m.. + m. n., LCtte pit.u. 3.4.13.13 off itt Piri C 3 81

3. 2 3. 8..

=tf Lit 57788 4 is .cf Lil PtPi%G 3.11 $t fuett 1 Il 18 8 UP,4 4 9 f.!! !$ Latt.t.8 8 il 171.. Ottis.,;s, wit! O Cast g,s.it ',.N *t.f. i,', '-

  • f Mew /.* 3,83 ed.%..*.C.ltg;3ECC94481: *0i;..yh. ".*4,Ipp4 f.

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g. 7, J, y. :.,.,,p. g g p,,.,..., g-

, j 9.18 18 tt'tti ild .tg.14 33 Pi,wg

9. f 8.19 COLS LIE PIP'4G 11.11.15.84 28.2, C3tt Ltt P'Pt.G it.82,23 C:;S its 8!'it*

1 7.31 0:v*C:ste 13 grett 3:sc.te 23 t,I f Il P'E11dili8 28.29 OPPtt $3..CfMB l n n.,....n, n nsa ul. 23 UPetG P(tg. j! g(3{ g.Lg( ts.33 8 utets *t.s 33.34 Lt.: g stfu'. P.. n,38 .P t af C3.,.i t.t IPs.,s 48.43 58 ePete t.g i lo ._.e.,- ~ - _ ---z.=..

.s ~ FIGURE 2. STARTUPFEbcA 0 's. i l t A LOOP TEST DATA

... C !.00P IEST 0ATA

, y: -. n p ,s * / / / I / / 4 r- . l.......... -. . +.. ..,... +. .b.i.g - +..g. 8:... pJ

...:v;;. ;..u

.:,i.: 6W a. + - n.:[. .-+.a -a & 4.1.-ea

~ - %:.?-%:. 4.-n.:'.. M&E; v.

o I ~ I I u 3 l .c @.. Q.c+ cer. r.a g :... :.a. .ci. y -.;.~;.8;-- Q.,..wr.,.. i sf, aciv :v..c.9. : -:.~, :.. - v>.+-5.A,lg, :M - _i z u I -w 2 p I I 1 I ~ g t Ii / 1I A / I\\t\\ / I /# l I/ \\ I 1/ \\ J V L--- a/ 0 0 4 8 12 16 20 Time, Min \\\\

4 86-17 070 0 6 -Of FIGURE 3 STEAM GENERATOR LIQUID LEVEL (TEMPERATURE ADJUSTED) e A LOOP TEST DATA A LOOP CALC. DATA 8 LOOP TEST DATA 8 LOOF CALC. DATA 25 / / 20

f., 'a. ' -

);. .?.. (* ...,...:,. s. },, ~s se R:s.... .a. o .s.. ..s. ~.., 6.s

.b: r.cg.4 :..:
..,1... w...www.. ; a.,,w:..y.w

-v.v:<.: -e..y p i..z;+,...s~nw Mr. w a v = 15 x .,: ; s.,. sr.....,m.;, ; ;:.y..., y. h.J. :,.o. E6:, +...., .} u.:,.. .s.s..,..>:s$g..r.. f .., 4...y i%..: s;.e g -_ / <= / a w C-O* / O c 10 -- / / / / / o ./ / / // w E / / 5 / /. -/ /. = / n.5 / / / 5 / / a s -~---------~ ~~/ \\ 0 0 4 8 12 16 20 Tiaie af ter loss of power Min \\L .. = -. ~

k 6-li 00'0 0 6 - 30 ~ FIGURE 4 S.G. SECONDARY SIDE PRESSURE A LOOP TEST DATA A LOOP CALC. DATA

B LOOP TEST DATA 1100

- B LOOP CALC. DATA p\\ s

  • s 1000 -

~ I r N g g \\ \\ u p\\ \\ 900 - N ',. 'j.. ..\\.. u ~. 'c. p

~

3 ;.. .n_ p.. &.... ....N..<;. g...i e.z...:.- ~:..~ x.-v...s .;,:.. aum u..:,+.v . g s 9.5 i.. u.;n,. ...c,.. . :y :,.+y., ;. , 'f 5 g A %'5 \\ / \\ 5 .800 -- \\.. /:r.>.a;:. ..n , s.... :.:.4p 9 9 .w.- . e. n,~.<..gi. g w. <,.; r...@.x-a y.n. g :,.g;..,.g.:.;.; . u, z.:n.. y/... \\ \\\\ \\ s 700 - .g \\ "(*%. %.m \\ g \\ 600 - \\ \\ \\ \\ \\ s 500 0 4 8 12 16 20 l Time, min \\3

  • I Yfs

..'t .N ~. , WP,

2_. J2= .Y FIGURE 5 PRIMARYALOOPTE.NERATURE \\ l LOOP A HOT TEST DATA LOOP A HOT CALC, DATA

LOOP A COLO TEST DATA

  • ~ -, LOOP A COLD CALC. DATA 3

e 6 9 \\m -s 580 N 'N N s. x\\ 570 's, N h 1 N .t

a... =

3:... 5s0.;:...... . ( y.... e a...... u, a n.. e +y~ q. >,. ..~.-

..w

. :..,a. + - .. s... E t\\ 'N. .e 1

\\

\\ .:+.v, ~+.-c \\ ~ \\:+n :--:; ewe::hc.e-a.e.. \\ P~ - - > -:' 4

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..:.,m

.y. r 550 - \\ \\ \\ \\ \\ \\ \\. \\ \\. 540 -- \\' s \\ \\ \\ \\ '\\. \\ \\ \\ \\ \\ \\ 530 -- \\ \\, 3 \\ \\ \\ \\. \\N N \\ 520 x 0 4 8 12 16 20 Time, min O. - , see eng N, w, -- -~

..,... ca

)

,.i s, .:.- v' y I- / 't vv 14 v-e FIGURE 6. P'RIMARY B LOOP TEHPERATURE ~ s LOOP B il01 TEST DATA LOOP B,liOT CALC DATA LOOP B COLD TEST DATA LOOP B COLD CALC. DA1A l i N~ SEO \\ g\\ \\ .\\W \\ n(N v- \\ ~ \\*N,N N 570 - - \\\\ NN N N \\- .Ns s% 44. O...560 N

  • . * =..,...,

.s-, +: e..'..,..? s, g.: - 3: . - : R..A ", 1:.s.. ."-s-v: * '

*,.4
  • = '.

....,-;.?, w y r. s :.. y. e V, .= s 3 a... :. 9,+ =-c. ., :.. n. . s :.m ;w.- ~ 1\\ m O t CL Ee* 550 - - ..:n + Q s :;sp n:t:',~.,v.v..'. n -y.i. e1 M.I. 3- .u+.Y m: -:.s.:,;>.+.. w ',k G r.~ ..,y-:-.

  • tyc-.:: u-a

~ ~ \\ v i l 540 \\. l>t s 530.. 1 n \\ \\ k 520 0 4 8 12 16 20 Tirr.e, Hin \\S _,'T_. - _.. _ _.. '.. - _. ~ _,. _ __,,..._._,,--____,_._%-. " #_ ' ' _... ~.... '. _ _ - " _.. _ _, _. _, _. _.. _ *, - 37 """WF '"X.J'N m,.77%w.f s "

  • wM J #* M.'

-F*WP' Ge 8

  • 458 a

....~

86-110 TO O 6 -G4 FIGURE' 7. PRESSURIZER LEVEL r i TEST DATA CALCULATED DATA -m s 250 - - as O5 200 - .e,. .J ...,a* .g k.- .,*t...;,. ..w. ,. t..JS,....*- ,; a. - V e ?.- *.e. ,t. . t,- s. ,,..s - 6~.. -:;o..s..::,,:.s w. :t ,9 v.e !,-..g <.e;*.m L.:1. 6.cn :'e q.r...:....~ - ::s., w n ;.. s.s:;.:;.v L

  • . c

-e>e %.ss J N g% S..y,.*I50 " . y. pi p;.: b 5. ;.u ;.;.:.;.e:S-i,.:.,;.9;l. N qr.i. . s ^:9. ;-.(?;4 ie.i+.'j,1, -.t:. n.' q. s, w y p %. 9,. q,,.s. u 3 .b \\ l N u l A \\ l \\ N 100 - I N ' \\ g \\ \\ \\ \\ 50 0 4 8 12 16 . 20 Time, Min 6 = -,,-,-.----,-m --,-.-e-v- --r, ,---,-~se--v-~-~~ ~w~--ww~ ~ = = - - * ' * ~ ~ - * ~ ~ ~ * * * ' ' " ' " ' ~ ~ * " ' ' ' ' ' ~ ' ' ~ - " * " ' ' ' ' ' ' ' " ^ * * " ~ " ~ '

hh d I (/' '" ( },. FIGURE 6. REACTOR VESSEL PRESSURE - TEST DATA CALCt! LATED DATA 2200 t \\ \\ \\ 2l00 -\\ \\ NNN \\ s 2000 N c' N, N 'N E N N s E N N 1900 o N = NN\\. \\ \\\\ N 1800 \\ N N s, 0 4 8 12 16 20 Time, Hin l l !l l -- rw - ~ " * ~ ' _.... _. -, _ - - - - - - - - - - - - - - - - - ~ ' ' ' ' ' ' ~ - ~ ~ ~ ' ~ ~ ~ ~ ~ ' ', _ _ _ _ _ _. _ _ ... -,.}}

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