14
n etcg''o .. .e UNITED STATE # NUCLEAR REGULATORY g ( /255m - o /s'3 ' f',Vf 'g .' c ADVISORY COMMITTEE ON REA b [ W ASHINGTON. D. C.f M S/#7d - * o, A s ..... s t I June 8, 1979 . ACRS Members TRANSMITTAL OF ADDITIONAL INFORMATION ON NATURAL CIRCULATION COOLING AND RESOLUTION OF REVIEW COMMENTS Reference: J. H. Bickel, " Background Information for ACRS Members on the Natural Circulation Cooling Mode," May 15, 1979 Following issuance of the initial report on natural circulation cooling for emergency decay heat removal (Reference 1) a number of coments and inquiries were received. The purpose of this memo is to address these coments and to transmit additional information. Dr. Kerr has requested an opinion regarding the usefulness of existing startup test data as a mechanism for verifying assumptions made in the Safety Analysis regarding natural circulation capability. A generally accepted technical position is that startup test verification should be performed to verify the conservatism of certain critical assumptions made in the Safety Analysis which may not be verified to a high degree of precision using analytical techniques. (Examples of these types of assumptions are: control rod SCRAM times during reactor trip and pump coastoown flow decay curves during loss of pump motor A.C.) The effectiveness and availability of natural circulation as a mechanism for decay heat removal (for the complete loss of A.C.) is just such an assumption usually made in the Safety Analysis. Verifying the effectiveness and availability of natural circulation during startup testing is not technically possible, however. This is because startup testing is performed prior to the accumulation of a sufficient fission product (and activation product) inventory needed to produce appreciable decay heat. Reactor operators and designers have attempted to circumvent this difficulty by performing a natural circulation test by operating the reactor at low power and tripping the coolant pumps. Natural circulation is thus observed by using instantaneous fission power as a heat source. This technique clearly demonstrates the effectiveness of natural circulation for heat removal in the range of 1%-8% thermal power for steady states. However this does not demonstrate transient heat removal capabi nty for core heat fluxes decaying from near full power to the decay heat level. This is not to imply an area of non-conservatism, because pump flywheels are generally designed to assure transient heat removal by maintaining a sufficiently long free wheeling coast- down times. Another type of test performed to demonstrate the effectiveness of natural circulation is to simultaneously trip the reactor and coolant pumps from an intermediate-power level (25-40". power). This type of test more closely demonstrates transient heat removal capability but it suffers from the inability to observe sustained heat removal in the 5-7% power range typical of decay heat produced in cores with E.0.C. fission and activation product inventories. 80 030 co6 i / 6

n etcg''o UNITED STATE g ( /255m - o /s'3 f',Vf 'g

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n etcg''o.. .eUNITED STATE#

NUCLEAR REGULATORY g ( /255m - o /s'3'

f',Vf 'g.' c ADVISORY COMMITTEE ON REAb [ W ASHINGTON. D. C.f

M S/#7d -

*o, A

s ..... st

I

June 8, 1979.

ACRS Members

TRANSMITTAL OF ADDITIONAL INFORMATION ON NATURAL CIRCULATION COOLING ANDRESOLUTION OF REVIEW COMMENTS

Reference:J. H. Bickel, " Background Information for ACRS Members on the NaturalCirculation Cooling Mode," May 15, 1979

Following issuance of the initial report on natural circulation cooling foremergency decay heat removal (Reference 1) a number of coments and inquirieswere received. The purpose of this memo is to address these coments and totransmit additional information.

Dr. Kerr has requested an opinion regarding the usefulness of existing startuptest data as a mechanism for verifying assumptions made in the Safety Analysisregarding natural circulation capability. A generally accepted technicalposition is that startup test verification should be performed to verify theconservatism of certain critical assumptions made in the Safety Analysiswhich may not be verified to a high degree of precision using analyticaltechniques. (Examples of these types of assumptions are: control rodSCRAM times during reactor trip and pump coastoown flow decay curves duringloss of pump motor A.C.) The effectiveness and availability of naturalcirculation as a mechanism for decay heat removal (for the complete loss of A.C.)is just such an assumption usually made in the Safety Analysis. Verifying theeffectiveness and availability of natural circulation during startup testingis not technically possible, however. This is because startup testing isperformed prior to the accumulation of a sufficient fission product (andactivation product) inventory needed to produce appreciable decay heat.Reactor operators and designers have attempted to circumvent this difficultyby performing a natural circulation test by operating the reactor at low powerand tripping the coolant pumps. Natural circulation is thus observed by usinginstantaneous fission power as a heat source. This technique clearlydemonstrates the effectiveness of natural circulation for heat removal in therange of 1%-8% thermal power for steady states. However this does notdemonstrate transient heat removal capabi nty for core heat fluxes decayingfrom near full power to the decay heat level. This is not to imply an area ofnon-conservatism, because pump flywheels are generally designed to assuretransient heat removal by maintaining a sufficiently long free wheeling coast-down times. Another type of test performed to demonstrate the effectivenessof natural circulation is to simultaneously trip the reactor and coolant pumpsfrom an intermediate-power level (25-40". power). This type of test moreclosely demonstrates transient heat removal capability but it suffers fromthe inability to observe sustained heat removal in the 5-7% power rangetypical of decay heat produced in cores with E.0.C. fission and activationproduct inventories.

80 030 co6 i /

6

.

-2-. .

lAs a personal opinion the most "optimium" way to verify the actual naturah following

circulation assumed in the Safety Analysis would be to perform t e -

test:

ascertain experimentally what portion of the control rods iimust be SCRAMMED from full power to achieve an instantaneous fiss onFirst:(A)

power in the 5-7% power range.simultaneously trip this portion of the control rods and g

(Shutdown control rods would be maintained(B) Second: )available for complete reactor trip should this become necessary.the coolant pumps.

and availability of naturalThis type of test demonstrates the effectiveness thercirculation as assumed in the Safety Analysis in an integrated f ashion raIt would also appear prudent to examine

than by a piecemeal fashion.sensitSities of natural circulation to external parameters such as steamThis could be done by analysisgener;.cor pressure and steam generator level.as supplemented by test data. d resMr. Ebersole noted that the simplified analysis did not discuss the proce uThis is

for establishing natural circulation when subcooling is lost.If subcooling is lost, it would be best if subcooling be reestablishedPressurizer

as quickly as possible either via High Pressure Safety Injection orDuring a loss of A.C. the HPSI pumps can be powercd by the Emergencycorrect.

heaters.Diesel Generators. t ralMr. Etherington requested an assessment of whether coolant flows 0 6 g na u- acirculation were in the turbulent regime, thereby, justifying the useThe attached analysisdifferential pressure loss term proportional to d':The flow will be turbulent in a straight tube steamif it is

generator if it is greater than 2.0% and in a U-tube steam generatorIn a reactor core, the situaticn becomes a bit mo-As a firstt ould clarify this point.h

greater than 1.25%. difficult due to the distribution of design flows through the core.These requiredcut it appears flows greater than 2.3% will be turbulent.flow values (to assure turbulence) have a net decrease with increasingtemperature due to the decrease in the ratio of viscosity to density with

increasing temperature.

John H. BickelACRS Fellow

Attachmant:Evaluation of Limiting Flow Regions

for Ttrblent vs. Transitional Flow

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._ 1. t 1 1 Y: .._t= _ - * = - = u t- :- %----* =

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I- iN i ?".}:'_E!5= -M- 5j7- L I'~2.5! "EEEE . 'W . j '.' - P' r1 .^

~ | F :- - .

.|-- 4. u h.-3

.- iitiJ g:t - ._.p: .- : .;; t- . c .!- _a; ._ -- j . . jr.3 ..ili;i _iij-;_ njuri=~' J ;i_..

gi-iq s =E

: : : -.a a : . . _- -

gu it - + :q - -F i .i; i- - ci: - "F !c_ _ . _ _. 2e ____;.,_ .n _ .

.21.8 .q. qqi:wi.p;;

: 4. - ! ..b .. _ ul:. _. .. =.

-

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n. l ._ -.nt....

q:, | . . . I.. . . . . . . . - . - : ::u=*:: .- : .u :: _ .n. :

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r

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W MOWINIW~~

. ?|__ . _ . 1. .__.# ._. _ . g. ,L. . :_J =. . ..~.,p

_- .

riit WI i~~ ~ -

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I TABLE A-3 (Contimawd)

L1 QUIDS

r , ,, . x ie .x* * ed. x im . , um(F) Ob/'' ID'"# Ub# I'4 I'# I8""# F' #

ft) Ib F) fa see) see) br it F) (sq rt/ht) (1/D(t/F es tt)

Water

32 62.4 1.01 1.20 1.03 0.310 13.7 5.07 - 0.3740 G2.4 1.00 1.04 1.G7 0.325 11.6 5.21 0.20 2.3 X 10'50 62.4 1.00 0.S8 1.40 0.332 0.55 5.33 0.40 8.060 G2.3 0.9'.M OJG 1.22 0.340 8.03 5.47 0.85 18.4 |70 G2.3 0.098 0.653 1.06 0.347 6.82 5.57 1.2 34.680 G2.2 0.998 0.578 0.03 0.353 5.89 5.68 1.5 56.090 62.1 0.997 0.514 0.825 0.350 5.13 5J9 1.8 85.0 (100 62.0 0.09S 0.45S 0.740 0.364 4.52 5.88 2.0 113 x 108 *

150 61.2 1.00 0.292 0.477 0.3S4 2.74 6.27 3. '. 440.0 |200 60.1 1.00 0.205 0.341 0.394 1.58 6.55 40 1.11 X 10'250 58.8 1.01 0.15S 0.260 0.306 1.45 6.69 9.8 2.14300 57.3 1.03 0. V.o 0.220 0.305 1.18 6JO 6.0 4.00350 55.6 1.05 '.105 0.159 0.391 1.02 6.69 6.9. 6.24400 53.6 1.08 9.001 0.170 0.381 0.027 6.57 8.0 8.95450 51.6 1.12 0.250 0.155 0.3G7 0 S76 6.34 9.0 12.1500 49.0 1.10 0.071 0.145 0.349 0.87 5.00 10.0 15.3SJ't 45.9 1.31 0.064 0.139 0.325 0.93 5.05 11.0 17.8600 g 42.4 1.51 0.058 0.137 0.292 1.09 4.57 12.0 20.6

Iy a e, s X I@ eX1P &.X tm ArXtrg Ob/ (Bto/ 06/ (so ft/ (3tu/ Pr #gg,) g3fy)su ft) 16 0 ft ase) ese) hrit F) | (1/T eu fu

Commercial Aniline

60 64.0 0.48 325.0 5.0S 0.10 56.0 3.25 I f100 63.0 0.40 170.0 2JO 0.10 30.0 3.24 0.49 21.6 X 108 ;150 61.5 0.505 06.5 1.57 0.008 18.0 3.16 0.492 64.5 t200 00.0 0.515 61.1 1.02 0.096 11.8 3.11 *

300 57.5 0.54 32.5 0.565 0.093 6.8 3.00 '

Ammonia (Saturated Liquid)

- 20 42.4 1.07 17.6 0.417 0.317 2.15 6.940 41.6 1.0S 17.1 0.410 0.316 2.00 7.04

10 40.8 1.00 1G.0 0.407 *0.311 2.07 7.0832 40.0 1.11 16.1 0.402 0.312 2.05 7.03 1.2 238 X 10'50 39.1 1.13 15.5 0.39G 0.307 2.04 6.05 1.3 266

,

80 37.2 1.17 14.5 0.3SG 0.293 2.01 GJ3120 35.2 1.22 13.0 0.355 0.275 1.90 G.40

Frcon 12, CCI,F,, (Saturatnl I.iquid)

-40 04.8 0.211 23.4 0.300 0.010 5.4 2.00- 20 93.0 0.214 25 b o 272 0 0 to 4.8 2.01 1.03 4.6 X 10'

O UI.2 0.217 23.1 0.253 0.041 4.4 2,07 1.05 5.2720 89.2 0.220 21.0 0.23S 0.012 4.0 2.14 1.34 7.8032 87.2 0.223 20 0 0 230 0.042 3.8 2.10 1.72 10.560 83.0 0.231 1S.O 0.213 0 012 3.5 2.10 2.1 14.4

100 78.5 0.240 1G.0 0.2tm 0 040 3.5 2.12 2.5 19.4120 75.0 0.244 15.5 0.204 0.039 3.5 2.12

397

-

--- . . . ... , , _ ,

Krish Q Prtnapres of Heat TrandmCop *f r;[f RSh IhJd Tewdook Co., Ccrowlon FA

M6E il of 11