ML19308D692
| ML19308D692 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/13/1971 |
| From: | Vaughn R GILBERT/COMMONWEALTH, INC. (FORMERLY GILBERT ASSOCIAT |
| To: | Obrien W Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19308D691 | List: |
| References | |
| 14374, NUDOCS 8003120823 | |
| Download: ML19308D692 (5) | |
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December 13, 1971 C ' L1 - 'I e
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[f f ww ? 953' Up .i. Mr. W. S. O'Brica 1:ncicar Projecc Coordinator e .- w: -o .- T ' . :? 4 M. e n'.
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' Re IIPSil Reactor Eu11 dint; Spray C.
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. . .9 . .* . c Dacay IIcat Recoval Pu=ps Crystal River t.' nit 3
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Dear lir. O'Erlan:
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The EC issued Safoty Guide IIo. I cn !!ase=ber 2,1970. This in a (cuido) sug3ceted cotbod of cniculating the Het Positiva Suetics I! cad (11?:7.1)' for
- the cr.nr;7.:ncy coro coolina cad ccntai== cat heat rcroval syste:2 pu=os (son-attah ) .-t k's e-ly call rhme As tha Ecactor Du11dina Upray (R:1S) -
. and Decay llont Reacnral (011) peops.
Uwe asst:sa the debatable nasu=ptions in Safoty Guida r.o. I are correct it is possihlo that the present n.3.S. and D.!!. ptz:7s ecy not hava sufficient EPSIl to sarisiy tha EC Salaty Guida requirments. . Theco nyatnes vera desi,v:ad and the current pt==ps purchased prior to tim is:n.u:nce of tua safety cuido IIo.1. Ua now havs . .
- 1. Pericued these systen denicas aconnin;g the requirrnmM act forth
.. in Saicty Guida Iso. I hold.
- 2. Vo also performed prolininsry calculatic is to deter:sino t.hother or not tha '.'. 3.S. c. .d cita i).11. re cial n rate : p: na h.:d e diicient li?.TJ .Trailable to catisfy clu S ciety Guido require?. ants. ha results of the:o prelicinary calculationa arn chova in the attachnent to this Iceter.
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Calculatio-.s Perfz. sed ' l Several Incs-of-Ccolant-Accident (14CA) rceircul:t$1on conditions havu been' n.a.217:cd . 02 these the taret caso is d:a >oct LLCA rceirculatica ccads.tien calculated in GPS 11 Calculacions Soc.:ica III (pa ,cu 23 to 32) of tho atta ch-
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' FPC 5043 Pass 2 l
Ecst. %is vorst possible caca assumes a bres!: in tha reactor ecolant piping ; at the top of the acca:n nencrator (clevation 177'3"). ha primar7 icop cco. l poets vill retain, and vill not loose, their vster with a break. at this cicvation. 1:sdcr this condition less water is available to ficod tha reactor building floor.resulting in a lower untar level within tha reactor, building su=p. - - -
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Calculation Results fo- Uorst case NPSH Calculatio .s Section III
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A deficiency of .62 = (23.34 ft - 22.72 ft) ft H,io in U?SH availabla is real-i=ed under tha above recirculation condition (scE Tabla V ite:a 2 of attach-ment). - <- - . - -
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Duo to the cancervativa assu=ptions and the fact that en estimated (probably conservative) UPSU value of 13'0" Ii20 is used as a requirc=cnt for the re-actor building sprn.' pumps (RBSP-33)(refer to I?orthington pe=p curva E-135235 ^ dated 5/9/67) the roti=sted LPSH therefore calculated and given in ?sbles I throus V, inclusive, is c=pected to be larger than the actunt UPS11 required (appro=inately 11'0" H 2 O). lia believe the I:FSH deficiency of .62 ft H2 O vill. then be core than cocpensated for, and va probably vill ha 01. Calculation nesults - N?sn Calculations Sectien I . De calculations perfor=ed in this ocction of' the attachment ara nada with the ascu=ption that no ascrqcacy cooling water is retained by the pri=ary cysten cocponents. This allous a larger vole =a of H,,0 ta flooa the reactor building su=p thus a higher cicvation of water relacIvo to tha reactor build-ing cpray and the decay heat' removal pu=ps cuction ccatcrlina. Under these. condition?, adcquata 1:PS3 is availablo. Results of calculacions ara chos.n
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on Tablo I of the attachment. Calculation Renults - r*Ft Calculattens Section II na calculations perfor=cd in this cection are cada with the aanu:ption that a breati occurred at the reactor vessel ec::les. tiith this nas_- pcien, ade-quato NPSU is available to tho reactor buildin3 apray and tha decay heat recoval pucps. Ecsults of calculations are shotnt on Tablo 17 of tha attach-ccat. Conclusion The only danner of the ccaracncy core cooling (D11) aol contair:2 cat heat ro'- coval cyaten G.M) pt:2p3 not havina suf ficient *i/S3 availabia to satisiy Safety Guida 2:o. I requirements, if this turns cut to bo' t;ta actual cass,11. _ _ , "
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ltr. ft. S. O' Brian . Florida Power COIAretica
. December 13, 1971.
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*0 3 3~}h' w h. fft1 ::2 sotaly dependent on the allcr.rsbie reactor building vatar Icvel, (i.e. static unter heads .dovo the Pu=p suction pipe centarline), for titich thm AEC l (Safety Guida 1 essu:::ptions) vill allev credit to ba ta'.:en in calculating l the availablo 1.7511. , ,
In essence, with AEC Safety Guida 1 assu=ptions, any arbitrarily imposed - water IcVet other than the calculated =inicu:2 post LCf.A vater level in Sec- l tion III of the attachnent could seriously inpair the ability of tha pumps ) to perform their safety fu=ction.
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Arbitrarily i= posed H2O level are discussed and evalunted in the following attach:cen% If there are .'ny questions, please advisa. ,
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Enc. - R. E. Vaughn
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( - ab,i . ' ,, ' . . l SIDDIARY OF INPUT ASSUMPIIO;iS USED, RESULTS OF NET POSITIVE' SUCTION !!EAD CALCULATION, BRIEF - EVALUATION OF AEC SAFETY GUIDE NO. 1 DATED NOVEMBER 2, 1970, AND POSSICLE RESOLUTIO!!S IF . AEC INFORCES SAFETY CUIDE.MO. 1
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The NPSil availabic to the decay heat and reactor building spray pump.s during post LOCA recirculation phase has been calculated based on " Issued for Con ~struction" piping drawings and the follouing assumptions: $
- n. Pipe and fitting losses calculated using the information in Crane Technical Paper No. 410.
. b. Total required flow in a single string (i.e. consisting of one decay heat and one reactor building spray pump which have a common suction -
line from the reactor building sump) which is 4500 gpm or 3000 gpm to the decay heat pump and 1500 gpm to the reactor buil. ding spray pump.
. ,c . Sump water temperatur6 equals 226 F (peak value from Figure 14-61 FSAR). '
- d. Reactor building pressure equals saturation pressure at a temperature O
of 226 F. ,
- c. Water Icvc1 in the reactor building is at 99.82 ft.
, This water Icyc1 in the reactor building is derived from preliminary calculations in which all sater capabic of flooding the reactor building and all dis- -
placements of water such as components, sumps, tunnels, ete have been
. analyzed. This wat'cr Icyc1 will be referred to as the post loss-of-coolant accident (LOCA) water Icyc1 throughouc t.his HPSII revicu.
The NPSH's available to cach pump, as calculated on the above basis, are com-pared belou uith the NPSil required by cach pump at the f, lows indicated above.
' Rcquired NPSl! is from curves furnished by the pump manufacturer. The requir'ed .
NPSil in feet of 1120 is an approximate figure, it is expected that the actual l I pump performance of NPS}l required vill be somewhat 1 css as compared. to thus -- w
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3 , ,. - , l figure shown below. Calculated NPSif Flow Availabic ft U O 2 NPSH Pump Ra.te ' El. Head.+ Vel. Required gym . Head _ f t M2 O_ R.B. Spray Pump A (BSP-3A) ' 1500 13.25' 13.00' R.B. Spray Pump B (BSP-3B) 1500 12.38' 13.00' _ 1 D.H. Pump 3B (DHP-3B) 3000 16.46' 13.50' 114 D.H. Pump 3A (DHP-3A) 3000 - Similar 13.50' ; The above method of calculating available NPSH is in agre'ement with AEC Safety
- Cuide No.1 in that no credit it taken for blowdown pressure in the reactor building. The Safety Guide states that "no increase fri containment pressure from that present prior to the postulated loss-of-coolant accident' may be assumed.
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- This position can readily be applied when c'alculating NPSH available with a sump water at a temperature less than 212 F. This po'sition [s* unrealistic when the '
sump water is above 212 F in that it would 'mean the entire large inventory of water in the reactor building would boil away without increasing the pressure in the reactor building. It is more realistic to ass'ume reactor building pressure will achieve equilibrium with the sump water at t a saturation pressure of the-sump water. Any abrupt loss of saturation pressure in the Contaitunent Vessel would result in flashing of the sump water at temperatures above 212 0F and Jm-mediate equilibrium of the reactor building pressure with the cump water at saturation pressure.vould be restored. There'are two potential po,t LOCA vater elevations in the reactor building which . AEC could impose other than the calculated minimum post LOCA water elevation of i , 99.82 ft, currently being used. These are the top of the reactor building ' sump l (elevation 95'-0") or the centerline of the outlet pipes from the reactor build- - l .
. ing sump (elevation.86'-3") .
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'u 4 Thn effrct o[ thesa louir cicvations of, the water icvel, uithin the reactor building on the available NkSII vs required NPS11 are i:ompared on Tables II arid ^ III. Assuming a reactor building water clevation / of 95'-0" (top of reactor- building sump) in reference to the reactor 'ouilding ,sp* ray pump suction centerline eleva-tion of'78.5", the total elevation head available plus the existing velocity head at design f' low would equal approximately 17.9 f t. As noted in Table II the total required head .to the pump is 22.47 ft. Thus, by calculating the available NPSH to the reactor building spray pump (R.B.S.P.- 3A) with a water elevation of 95.0 ft, in the reactor building sump a deficiency in net positive suction head available of 4.57 ft is realized. I Therefore, in conclusion, any arbitrarily imposed unter elevation other than the currently calculated post LOCA water elevation would ' severely effect the - net available NPSH. In order to remedy the inadequate available NPSH, due to water elevations lower than POST LOCA levels, it appears that major design changes and/or modifications to the present design would be necessary.
~ , In the event that the elevation in the reactor building sump is to be a maximun '
of 95.0 ft., the following alternates could possibly be used in resolving the l 1 inadequate NPSH. - l
- a. Exchange present pumps and purchase new pumps with an NPSH requirement of approximately 8.0 ft. ~
- b. Relocate pumps from the present design location to a lower elevation, in reference to the Rx building sump clevation. .
- c. Design an independent and mechanical means of increasing the net avail-abic NPSl!. . .
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Review Calculations and Results regarding the' height of the II2 0 in the Reactor Building. - The follouing preliminary calculations were perfor.med with the assump-
.q tion that all the H 2O available 'will flood the reactor building floor
- l following a 100A. Results by calculating the available NPS11 in. this manner indicates the highest possible H 2 O level within tI reactor building. -
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Tatc1 110 9 in Rector Systcm Under Nermal Operation (hot) Reactor Vessel 4,058 ft3 Steam Generator (2,030 ft ) each (2) 4,060 ft3 Pres'urizer s 800 ft3 Reactor I'nlet Piping ' 1,085 ft3
' Reactor Outlet Piping 979 ft3 Pressurizer Surge Pipe 20.ft 3
Pressurizer Spray Pipe , 2 ft3 - Reactor Pump Casing (98)'(4) 392 ft3
. 11,396 ft3 -
Total Water Available/ Sources Outside the Normal Operating' Primary System -
- Makeup Tank .' 600 ft3 ~
Core Plooding Tanks (2 tanks) 1,880 ft3 Borated Water Storage Tank - 50,800 ft3 Sodium Thiosulf, ate Tank ' 1,600 ft3 Sodium Hydroxide Tank 1,500 ft3 56,380 ft3 Total Cubic Feet of Water available and capabic of Flooding Reactor Building floor 56,380 ft3
, . _11,396 ft3 67,776 ft3 e
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- 9 The following components will help in lowering the height of H2O in the Reactor Building since they displace or are filled with water: ,
Instrumentation Tunnci . V 4' x 4' x 40' , 650 ft3 R.C. Drain Tank V (4')2 (22') , 1,101 Building Sump Volume - V 10' x 10' x 9' . 900 . 2,641. It3 _Hence the net volume available and capable of flooding the reactor building
- floor- ,
V 67,776 ft3 - 2,641 ft3 = 65,135 ft3 . e s e
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Tot ,1 Area displaced by Components within the Reactor Huildinn Steam Generator (2) , A = 5.14 (6)2 = 113 . 226 Pc2 Pressurize.r ,
. = 3.14 (4)2 = 50 .
50 . Primary Shielding ' A'= 3.14 (16.5)2 = 855 855 Main Coolant Pumps A = 3.14 (3.75)2 = 44.2 44.2 Main Coolant Pump Pipes (Vertical 4) A = 3.14 (1.5)2 = 7.07 , 28 Misc. Shielding ,
.A = 13.x 5 = 65 . ~
= 3.14 (142 -11.52 ) = 201
= 45 x 5 (2)
= 400 -
= 62 x 2.5 = 155 ,
= 10 x 9 = 90
, = 75 x 1.5 = 113 .
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g a c,1culnted Voke <te. of teedge Surreunding Reector Building Baso 9.0 - 5.6' = 3.4' - c c tancC .= x x . a a+b; tanot = a .'. a + b ^ a - *
"' " a +c b X " (3.4)(1.5) [ 5.1 3.4 + 5.6
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.567' '
9 - Area triangle A = 1/2 bh = 1/2(.567')(3.4') =
.96,4 ft 2 Area triangle B
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1/2 bh = 1/2 = 6.75 ft2 Cire. Rx Bldg. = Y d = 3.14 (131') = 412 ft , e Total Volume displaced by surrounding wedge
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Arca A A - Area A B = 5.804 '
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Volume displaced = 5.804 ft2 x 41,2 ft = 2389 fit 3
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(. ~ . (. Total Volume it O followino: LOCA vs Height in Reactor Building 2 Total Volu::te of H2 O uhich is vapor = 5'/. of 65,135 f t3 = 3260 ft3
~65,135 - 3260 = 61,875 fe Total H O in Rx Bldg. following LOCA' 61,875 ft3 .
2 _ Total Area of Rx Bldg. floor = 13,460 ft 2 To'tal volume displaced by wedge = 2389 ft 3
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Total area displaced by components = .2410 ft 2 V
- h =
I = _ (61,875 ft3 + 2389 f t3 ) h = 13;460'ft2 - 2410 ft2 , c . 64,264 ft3 '
.h = = 5.82 ft -
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Available NPSH vs Required NPSH with Total H O Available 2 Hence; maximum2 H O level in the Reactor Building is establi::hed ' (All Reactor Bn41Mng components empty except the reactor' vessel)
. Based on BSP-3A & 3B and m P-3A & 3B Operating at 1500 gpm and 3000 gpm, respectively Total Elevation Potential Elev. Head Available ' Total Head Required Head Available on this Basis to Plans - Ft . 1. From maximum post IDCA water M cicvation to centerline of BSP-3!L Pump Suction 100.82' l.'40' Vel. Ed. '
to 78 5' W 9.47' Ed'. Loss -
. 22 32'Elev. Ed 13.00'N MH Req'd.
23 72' 22.47' p u. e i m 2. From ma h. post LOCA water H elevation to centerline of 1.40' Vel. Ed 10 34' Ed. Loss - VSP-3B Punp suction 100.82' 22.32',Elev. Hd 13.00' NFSH Req'd. to 78 5'
. 23.72' 2'3 34'-
3 From maximum post LOCA water i elevation to centerline of - 8 MF-3D Fump Suction 100.62' 2 30' vel. Ed to 77 5' 9 16' Ed. loso - d [fLQj - . 23 32' Elev. Ed . _13 50' Ese)? . . g' Q: . 25.63.'
. 21.66' .
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. Available NPSH vs Required NPSH i
J This table shows the available NPSH vs the required NPSH in the event AEC does impose a mnviam water level of 95' (clev.) which is the top of the reactor building sump Total Elevation -
. Potential Elev. Head Available Head Available .Tctal Head Required on this Basis to Pumps Ft
, 1. From Top of. reactor building (R.B.)
Sump to Centerline of BSP-3A Pump 1.4' Vol. Ed. Suction 95' - 78 5' 9.47' Ha. Ioss 16.5' Elev. Hd. 13.00' NPSH Required. *
'17 9' 22.47'
- 2. From Top _ of reactor building (R.B.)
p Sump to Centerline of BSP-3B Pump 1.4' Vol. Ed. Suction 99' - 78 5' 10 34' Nd. Loss m 16.5' Elev. Ed. 13.00' HPSH Req'd. 17.9' o 23 34' *
- 3. From Top of reactor building (R.B.)
Sump to Centerline of IMP-33 Pump 2 3' Vol. Ed. Suction 95' - 77.s' 8.16' Ed. Loss 17 5' Elev. Ed. 13 50 . 19.8'
,' 21.66' . 6
- 4. DEP-3A Similar . . .
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.~ d Availabic NFSH vs Required NPSH . .
This table shows the available UT3H vs the required NFSK
. in the event AEC does i..: pose a maximum clicwable water level '
of 86.25' '(Centerli'ne of reactor building sumn outlet piping) ~ Total' Elevation
. Potential Eley. Head Available Total Head Required Head Available on this Basis to Pumns
- 1. From Centerline R.B. Sump. : ,
m' - outlet to centerline of BS'P-3A 1.k' Vol. Hd. 9 47' Ed. Loss Pump Suetions 86.25' - 78 5' 7.75' Elev. Ed. 13 00' NFSH Req'd. -
. 9 15' 22.47'
- 2. From Centerline R.B. Sump -
- outlet to centerline of BSP-3B 1.4' vel. Ed. 10 34' Ed. Loss
- Pur.p Suction 86.25' - 78.5' '7.75' Elev. Ed. 13.00' NPSH Req'd.
9 15' . 23 34'
- 3. From. Centerline R.B.' Sump '
@ ,g d outlet to encterline of DHP-3A Pug Suction 66.25' ~ 77 5' 1.4' Vel. Ed.
8.75' ' 8.16' Ed. Loss 13 50' UPSH Req'd. .
. 10.15' ,
21.66' ^ M y @ Kote: For AEC to deter.ine elevction level 66.25 feet cs the :r.c:d:mm cllc,.able vctor level to be u:ed in calculatig the evcilabic E2SH vould be unreclistic,.in that at this elevatio:t g level there is an incufficient volur.: of veter available to the pur.ps. ' a y . e j'
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( - ( .
, NPSil CAI.CULATIONS SECTION II -
Results of a reactor coolant break at elevation ,128'-0". ~ The following preliminary NPS:t and 2110 height within the reactor building calculation was perforced with the assumption that a primary piping break occurred at the reactor vessel nozzles. Calculating a breaI: at this cicva-tion vould mean that a minimum of 1I 2 0 uould be retained within the l>rimary System components hence, the post LOCA water icvel would be at its maximua in the reactor building . . y p . g .
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. 1 Postulating a reactof coolant break i.t the reactor coolant inlet and/or outlet l pip?ng to the reactor vessel at elevation 128'-0".
_ Volume of H O retained within the reactor building components , 4 Steam Generators would retain .- 998 f t3 (2) = 1996 ft3
~
Reactor inlet piping wou1d retain 979 ft3 Reactor vessel would retain . 3080 ft 3 Total retained H 2O 6055 ft3 - Total H90 available (storage) Makeup tank 60.0 ft3 Core flooding tanks (2 tanks)
^1880 ft3 Borated water storage tank 50800 ft3 Sodium thiosulfate tank ..
1600 ft3
- Sodium hydroxide tank 1500 ft3-
, 56380 ft3 Total cubic feet of water available and 2
, capable of flooding r'eactor building floor 61721 ft3
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The following ca.nponents will help in lowering the height of H2 O in the reactor building, since they displace or are filled with water. Instru :entation Turmel . 640 ft3 - P..C. Drain Tank , V2 (4 )2(22') 1101 ft3 . Building sump volume V 10' x 10' x 3' 900 ft3 264'1 ft3 C
- s. .
WIH PIIB
. s . .u La 9
16
TOIAI. AREA DISPLACED BY COMPONh2 TIS . l
~
Sec.~.:a Generator (2)
. A= (6)2 ,. 113 '
226 ft3
~
Pressurizer . A= (4)2 = 50 50 Primary Shielding A= (16.5)2 = 855 855 Main Coolant Pumps A= (3.75)2 = 44.2 44.2 Main Coolant Pump Pipes (Vertical 4) 2 - A = < (1.5) = 7.07 , 28 ~ Misc. Shiciding A = 13 x,5- = 65 '
=
(142 x 11.52 ) = 201 . -
=
45 n 5(2) = 400
=
52 x 2.5 = 155 .
- =
10 x 9 = 90
=
75'x 1.5 = 113 1074 2410 ft2 TOTAL AREA CF REACTOR BUILDING FLOOR
- A = r2 =
3'14 (65.5')2
. =
13,460 sq ft
\
G t
. m
- . t' >... . . . - -
The height of H2 O in reactor building = 8 28 H2 O available Total area of R.B. floor y 3 61.721 ft - 2641 ft3 59,080 h.A . . .h . - - 13.460 ft2 - 2410 fc2 11,050
- 57. of Vol. = Vapor = 2,954 ft3,
~ 59,080 - 2,954 56,126 -
" ~
11,050 11,050
/
Total Height in Reactor Building = 95 + 5.08 ft = 100.08 ft. *
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( .- TorAL VOLIME DISPLACED BY WEDGE 9.0' - 5.08' = 3.92'
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cance = c ; e,na = x ..
- c =
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..x =
ae
- a+b a ,a + b a- ,a+b (3.92)(1.5) 5.88- -
.555' x-= = = - -
3.92 + 5.08 9 s Area d A = 1/2 bh = 1/2 (.655')(3.42) = 1.28 ft2 Area-d B = 1/2 = 1/2 ( * }(' } = 6.75 ft2 , t .
. Circumference Rx buildin'g = if d =
3.14.(131') = I12 t ft Total volume displaced by surrounding wedge . Area triangle A - Area trinagle B = 5.47 ft2 - Volume displaced = 5.47 sq ft x 412 ft = 2260 ft3' - D e s WI5Y EU - m o,
- 01. _
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TOTAL H2 O AVAILABLE VS 'llEIGHT IN Rx BUILDING (incidding volume displace by wedge) - 56,126.+ 2260'= 58,386 '
~
58,386 . . l h = = 5.28 ft . 050
.~
TOTAL HEIGHT H2 O IN Rx BUILDING -
~
95 ft floor elevation -F 5.28 f t (H 2O) = 100.28 fe eley '#
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IW. 12/1'/.72 TABLE IV - Availabic NPSH vs Required HPSH '
, Postulated break at El.128'-0" s Based on Building Spray Pump BSP-3A & 3B ond Decay Heat Pump M P-3A & 3B Operating at 1500 gpm and 3000 gum Respectively
- Total Elevation Potential Elevation Head Available ~ Total Head Required Head Available -
on this Basis - ft _to Pudps - ft
~
- 1. From post LOCA wate r elevation
' to centerline of BSP-3A 1.40' Vcl. Hd. 9.47' Hd. Loss Pump Suction 100.28 to 78 5' 21.78' El. Ed. 13 00' NFSH Req'd'. .
23 18' 22.47'
- 2. From post LOCA water elevation g to centerline of BSP-3B 1.40' Vel. Ed. 10 34' Hd. Loss Ptm!p Suction 100.28 to 78.5' 21.78' E'_. Ed. 13 00' HPSH Req'd.
. '23.18' , 23 34.'
g
~
3 From post LOCA water elevation to centerline of MP-3B - 2 30' Vel. Ed. 8.16' Ed. Loss
- q' . trull yu.:p Suction 100.28 to 77 5' 22.78' Ele. Ed.. _13 50' NPSH Req'd.
3)CJ m jq
- as.08' -
a1.ss' c]g 4. mP-3A similar
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, NPSH CALCULATIONS SECTION III Results of Postulating a Reactor Coclant Piping Break at Elevat.lon 1N7'3"
. The following preliminary NPSH and2 H O level within the reactor building cal-culations were performed with the assumption that a piping break occurred at
~
"the top of one or both steam generators . (elevation 177'3"). A" break in the piping at this elevation would mean that all the primary system er aponents s would retain a maximum ambient of H2 0. Hence,' the post LOCA water level would be at its minimum within the reactor building.
e
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9
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e b G g e e e 9 9 9 9 1m p*. y1 3 g 9 23
, .; . . , t - r .. .. . . - \ . ( .
e Results of Postulating a Reactor Coolant Piping Break at the top of one of the Steem Generators - Elevacion 177'-3"
- Total 110 2 ncactor Systen under llornal Operations which tEuld bn retained.
t Steam Generator (2,030 f t?) each (2) 4,060 ft3 Reactor Vessel . 4,053 ft3 Pressurizer- ' 800 ft3 Recctor Inlet ' Piping 1,085 ft 3 Reactor Outlet' Piping' 979 ft3 Pressurizer Surge Pipe 20 ft3 Pressurizer Spray Pipe - 2 ft3 Reactor Pump Casing (98)(4) - 392 ft3
'11,396 ft3
~
. 1 s .
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~
- . MIRV "HHF '
]o 'l?ligfL,
.- 24
R. " ..:, v
\ , [
TOTAL WATER LVAILABLE
. e
~
Makeup Tank - 600 ft3'
. s .
Core Floo, ding Tanks (2)' . 1,880 ft'3 Borated Water. Storage Tank 50,800 ft3 Sodium Thiosulfate Tank 1,600 ft 3 - Sodium Hydroxide Tank 1,500 ft3 g..~'
, . 56,380
- Total Cubic Feet H 2O Available and capable. of f1poding Reactor Building Floor 56,380 t.
c t
. m - .
~
. g 25
... ... / , .. , .t , \ .
The follouink components will lower the height of 11,0 3 in the Rencto.- building since they displace or are filled with H 2 O_. - Instrumentation Tunnel . 640 ft3 R.C. Drain Tank - 1101 ft3
, Building Sump .
.900 ft3
. 2641. It3 t
4 e
. I VIH. IIB r -
I n%. ., Jy ngLu.
- a. . .
t 26
m, m "
-r -
. s .
_ TOTAL AK% DISPIACED BI COMIONE.*TS , ,. -
~
Steam Generator (2) , 226 ft;E . Pressurizer 50 fb2 ~ Primary Shiciding 883n2 Main Coolant Pumps hh.2 ft2 Main coolant Pump Pipes (vertical 4) 28 2 . Misc. Shielding
- 1,074 ft2 .
2,410 ft2
'JOTAL APEA OF PICTOR BUTrnING FICOR A= x2 = 3.14 (65 5')2 = 13,400 sq. ft '
s
. O O
b L% S e
=M
- 27
. . - .y .- ~
u - e ,. e, .>. , Total volus.2 er H O Availablo
) Dight cf H2O in reacto'r building = 100aj arc (L Ol' It.52. t100r Y
- h = 56,380 ft3 - 26'4 1 ft3 53,739 n 3 h=E **
13,460 ftd - 2410 ft2 " 11,o50 ets ~
~
57, or volwt.c (H2 O) = vapor = 2687 ft3 . h = 53,739 ft3 - 2687 ft3 ,51,052 ft3 , , 11,050 fte 11,o5o pg2 , 1.62 ft 5 , TOML H2 O IEIGHT IN RMCTOR BUILDDiG = 95' +6 1.62I = 99.62.' 9 9 e
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~
9.o' - 4.62' = 4.38' . Tan oc = * ~ a 6 o_; tan x = E . . M =@..x= aeo - (b.38)(1 5) ,6.57 x,
= .73 ft ,
4.38 + 4.62 9 - Area triangle A = 1/2 bh = 1/2 (.73)(4.38) = 1.6 ft2 - Area triangle B = 1/2 bh = 1/2 (15)(9 0) = 6.75 rta CircumferenceExBb.dg.= d = 3 14 (131) = 412 ft . Total volwne displaced. by wedge Area triangle A - Area triangle B = 515 ft2 . , , Volume displaced = 5 15 ft2 x 412 ft = 2120 ft3 e e S e
- e e
e e i
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. . 1 30
=v, v.. u . . Total HoO Availabic vs hei,a.ht iip 0 in Reactor B R11n,3 (including volut:e displaced by veda;c) 51,052 ft3 + 2120 ft3 = 53,172 ft3 , . h=- O = 4.82 ft . Total height }Ip,0 in Rx Bldg. = 95' + 4.82' = 99 82' . e 4 e O e 4 0 e 6 9 8 e e o e r O
+
g O e 0 4 e
- 0 0
WlH '"11B
. : 3: 3IL -
9 8 6 31 -l
TABLB V
?-C Available NPSH vs Required HPSH Postulated Break at Elevation 177'-3" ~ Total M evation Head Available Total Head Bcquired Potential Head Elev. Available on this Basis ft to Pumps ft
- 1. From top of Reactor Building (R.B.) Sump to centerline of Building Spray Pump 1.4 Vel. Ed. '
' 9.47' Rd. Loss (BSP-3A) Suction 99.82 ,78 5' . 21 32 El. Ed. v - .^
. .. 7: <13 00' IESH Req'd. -
22 72' 22.47'
- 2. From top of Reactor Building (R.B.) Sugip to centerline of Building Spray Pump 1.4 Vel. Ed.
BSP-33) Suction 99 82 78.5 ec 10 34' ud. Loss
- 21. 32 ' E. Hd. ,
~"
.'13 00 NPSH Req'd.
u . 22.72.' 23 34' 3 From top of Reactor Building (R.B.) Sump ' ' ' ' - to centerline of Decay Heat Pump (IEP-3B) 2 3' Vol. Hd. .... 22.32' El. Ed/ g 8.16'-Ed. Loss
. Suction 99 82 - 77 5'.
fp. 13.50' NFSH Req'd. .
~
@Q 24.62' 21.66' '
. DHP-3A Similar *
'. m
,j, .1 S Tote:
Thic table chcus the availcble NPSH vc Required KISS when the post toc g H2O' level is at its
.em1=m. -
g p @D A noted in the cbove Tcble V No. 2, the building spray pump (BSF-33) has insufficient availabic NPSH. This is a result of a required estincted NSSH of 13 00 ft H20 g 'Et is believed that cetual . required MISH of thcae pumps (performe.nco test data) will comewhat less than the estincted ficere chown,on this tc'olo. Thus, providing sufficient available not positive onetica h d, 9
, I .
e e
S.;.J (. ( . S.WETY GUIDS 1 - IEf POSITb!E SUCfIO?i HEAD FOR
)
EfER3Ei!CY CORE CCOLI!iG /SD COiT/CDDDI!T HEAT RaiOVAL SYSTEi_IibuI I A. Introdu_ction . WIBVFill
- l Proposed General Desi 6 n Criterion 41 requires that the emercency cooling l
- and containment heat removal cystems be capabie of accomplinhing their required safety functions assuming partial loss of inctalled capacity. !
In current designs the ability to accomplish these safety liinctions l reliably depends in part on the proper performance of systcIt pmaps which, in turn, depends on tne conditions under which the yumas nu'afdperate. One of these conditions is suction pressure. This gttido describen a. suitab3.e relationship between increases in containment prensure caused by postulated loss of coolant accidents and ~the not ponitive atectio'n head (NPSH) of emergency core cooling and containnant heat re:.= val systeat , pwnps which may be used to implement General Design. Critorion 41. B. Discussion A significant consideration related to emergency core cooling anct containment heat: removal systems is the potential for deg2adell pinap performance which could be caused by a nimiber of factors, including inadequate NPSH.
. If the HFSH available to a ptcip is not sufficient, cavitation of the pumped fluid can occur.
Y is cavitation na3 reduce si Enifice.ntly the capability of the synten to a::co.r.plish ite .Mety i\tnctions. . It is important that the proper perfornance of emergency core coalit g and containment heat removal systenn be independent of calculated increa.kes in containment pressure cauced by postulated loss of coolont accidents in order to accure reliabic operation under a variety of p..icnible accider.t-e -,, , ,. , - - - -- , n - - ------- - - -
. _ e , . ,. (-
(.
/
\ -
. iar... conditions. For example, if proper operation of the emergency core
~
cooling system depends upon maintaining the containment pressure above a specified minimum amount, then too lou an internal pressure (resulting from impaired containment integrity or oper'ation of the cont.ainment heat removal systems at too high a rate) could significan517 affect the' ability of.this system to accomplish its safety functionn by causing pump cavitation. In addition, the deliberate continuation of a. 7.i.gh con-tainment pressure to maintain an' adequate pump HPSR \tou1d r sult in greater leakage of fission products from the containment end. higher po-tential offsite doses under accident conditions than would otherwise result, Changes in NPSH for emetgency core cooling and containment heat reuoval system pumps caused by increases in temperature of the puraped fluid. under loss of coolant accident conditions can be accor.vnodated yrithoutylianeg c on the calculated increase in containment pressure. Adequate NPSE can be assured by locating pumps at suitable elevations uith respect to the storage volumes connected to their suction sides, by using cultistage or booster pumps, by .a combination of these nethods, or by other techniques. C. Regulatory Position Emergency core cooling and containment heat removal systemn should be de-
=. .#
~
signed so that adequate. net positive suction head (NPSH) in provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated ' . loss of coolant accidents. * ~ l MED
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- 0 9Y37f GILBERT ASSOCIATES, INC. " " * *
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