ML19344A339

Monthly Operating Rept for Feb 1971

ML19344A339
Person / Time
Site:	Midland
Issue date:	03/11/1971
From:	IDAHO NUCLEAR CORP.
To:
Shared Package
ML19344A338	List: ML19344A337 ML19344A339
References
NUDOCS 8008180398
Download: ML19344A339 (14)

Text

- - .

p.. .,

d IDAHO NUCLEAR CORPORATION Nuclear Safety Program Division 4

4

(-

e 1

MONTHLY REPORT -

February 1971 The information contained in this report '

'is' preliminary d subject to further evaluation.

' 4 q09

~

,. , , ,. . - - - - + = - - - < * ' < - - - * - - " ' - - " - ' ' * " ' ' * * **

~

TABLE OF CONTENTS A. . Page LOFT Integral. Experiments and Radiologic 4 Studies .

......... 1 4

B.

. loss-of-Coolant ' Accident Analysis C.

.................. 2.

' Technical 1.ssistance in Reactor Safety Analysis I ........... 5

. D. .

Separate Effects Testing . . . . . ...

..... . ... . . . . . . . , _ 6 4 ,

i 4

i p

. l

,1 .

l

, 'il i

.e N .

s. ...

s -

February 1971 HIGHLIGHTS LOFT Integral Test Program Develooment: A study that identifies environmental conditions to which the LOFT primary sysuam instrumentation will be subjected during loss-of-coolant testing was completed. These envirenmental conditions are beirg used to develop test criteria for iastrumentation development.

Four LOFT design docu=ents were reviewed including the preliminary design description of the IDET blowdown system.

Loss-of-Coolant Accident Analvsis.: The FPFM (Fission Product Fuel Model) computer code is cperatienal on the IBM 360/75 system. This code, based on the steady-state fission product release model developed by Idaho Nuclear Corporation, calcuJ-tes: (1) noble gas release to the fuel-cladding gap, (2) fuel pin intenal pressure, and (3) noble gas release from ' defective fuel pins.

Separate Effects Testing: Semiscale Test 849 was completed. For this test a quick-opening valve was installed in a bypass line around the system low point to determine whether pressure buildup due to a water L

seal was responsible .in previous tests for emergency coolant expulsion through the break. This test with a high-inlet break configuration also had a nozzle in place of an orifice in the blowdown line. ECC liquid was ejected from the system in Test 849, as in previous tests with accumulator ECC, and at no time did ECC liquid-reach the core.

Semiscale Test 850 was conducted at the same conditions as Test 848, but without emergency core coolant injection. Test 850 was required to provide new baseline information for the later tests of the ECC series, since-several changes in system configuration have been made since earlier

. baseline ,tes ts without ECC.

G b

t iii 4

6 l

' .- c. r

--x 1 7 U February 1971 f

'A. ' LOFT Integral Test Program Development

' l.

~ Highlights

'A study was completed that' identiff as environmental conditions 1 to which the LOFI primary system instrumentation nill be subjected during loss-of-coolant testing. These conditions are being used to'

' develop test criteria for ir.strumentation development.

r Four LOFI design documents .were reviewed.

2. . Technical Activities Program Requirments and Experimental Plan for LOFT Integral Tests:

.On the basis of results fron RELAP-3 and THETAl-B computer code analyses,

~ the environmental ~ conditions at various instrument locations in the LOFT

Primary system during blowdcem have been identified. These conditions, r which include - fluid pressure, temperature, quality, and flow rates will -

=

be used for developing criteria for testing the LOFI experimental instruments.-

Drafts -justifying eight of the LOFT design features were submitted to the LOFI Project for review and inclusion in the LOFr design basis

, report. These drafts discuss the effect of the design parameter on the

LOCA response of the LOFT system. The drafts discuss the following j parameters

{ (1) Fuel geometry .,

~

$ (2) . Core diameter-

.. (3) Core height ..

(4) Initial cladding and fuel' temperature (5). Cladding dimensions (6). Initial DNB ratio

(7)' Core material properties

-(8) . Ratio of system volume to break area.

The following' LOFT design documents were reviewed and comments s'ubmitted to the LOFI Project.

4

. (1) . . .SDD 1.3.9, " Communication and Alarm System" (2) . SDD 1.3.22, " Plant Radiation and Monitoring System"

(3) PSDD 1.1.2A, " LOFT Blowdown System" '

i (4) ' , Appendix B', Revirion E, " LOFT Appendix B, Revision E, Mobile Test Assembly Parameters ' for all SDD 1.1 Series Documents" t

i l'

~ ~ '

, - - . . . . , - . , m~

~

u

. February 1971 B. _ Loss-of-Coolant Accident Analysis

1. _ Highlights The FPFM computer code is operational on the IBM 360/75 system.

This code, based on the steady-state fission product release model developed by' Idaho Nuclear Corporation calculates: (1) noble gas release to the fuel-cladding gap, (2) fuel pin internal pressure. and (3) . noble gas release- from defective -fuel pins.

2. Technical Activities Core Thermal Model Develoe=ent: Development of THETA 1-B code is

' continuing.

A procedure for taking RELAP3 output and preparing it automatically as THETAl-B input was checked and is operational (l).

An analysis of fuel bundle temperatures during blowdown with ECC spray injection following a break in a primary recirculating line has been performed for an early generation BWR (without jet pumps). The calculations were performed with the MOXY computer code employing vendor data ture.

for 'che core heat transfer coefficient and the coolant sink tempera-Figure 1 shows the comparison between MOXY and vendor predictions for temperatures of the axial hot spot of one of the four center rods and of the canister, both for the hottest fuel assembly in the BWR core.

, Identical maximum cladding temperatures (2180 F) were predicted at slightly'different times (238 see by M0XY and 210 see by the vendor).

Agreement between the maximum cladding temperatures is considerea to be due to the fact that the input quantities .for 'the two calculations are virtually the same. ' Hence, the curves in Figure .1 show only that the code calculational logics yield very similar answers for a common problem.-

Code changes are being made in the MOXY code to deceribe the geometry of the LOFT core. MOXY vill be used. for calculating the temperature during blowdown, . of the support tubes in the LOFT central fuel assembly.

View factors for thermal radiation a e being determined for the LOFI core geometry.

Analytical Models Describing Fission Product Behavior: The FPFM code (21 '

based _on the steady-state fission product release model(3) developed by Idaho Nuclear Corporation has b~ e en tested and is operating satisfactorily (1)-

Nuclear Safety Program Div;sion Monthly Report - January (Hai-26-71),

Feb ruatr 12, 1971.

. (2)- l Nuclear Safety Program Div;sion Monthly Report - December (Hai-7-71),

' January 15, 1971.

(3) W. A. ' Yuill, V.- F. Baston, and J. H. McFadden, " Release of Noble Gases from: Operating Fuel Rods," Ouarterly Technical Report Nuclear Safety Program Division October 1 - December 31,1969, IN-1320 (Augus t 1970).

I 2 's

. . . . _ - . . _ . - . - - . - . .~ . .. .. . . . -

i  !

/ *

/ ,.

/ .

l .

-l I I. I I I l l-l l l l

1. 1 l l l l g g i  ; gigi i

\ '

2000 -

Vendor Analysis / _

---

MOXY Results #

f

/ Hotrod f

/-

. /

g 1500 -

,/ _

• . /

w B .- -

/ AN -

- Conister g / / N%m g

a

,--s l E

1. 1000 -

/

f NOTE: These curves show only a j

f check of calculational methods for j j a common problem with the same input

/ parameters.

/

500 =--~-----

O I I I I IIII I I I lIlII! l l l l l 'l 11 i 10' 102 gos

, Time (sec)

INC-A 8FSO3 Fig. 1 Temperatures for a BWR Recirculation Line Break -- MOXY Results Compared with *

, Vendor Analysis Results for Common Input Parameters .

s

d February 1971 -

' on the IBM 360/75 system.- This. code has been developed for use'in radio-

. logical safety analysis of power reactors. Maxi:num noble gas release to the fuel cladding' gap, fuel pin internal pressures, and noble gas release from. defective fuel" pins are calculated for steady-state operating conditions. Fuel pin internal pressures are based on analytical- expres-

sions for the free volume of the fuel-pin as a function of temperature.

r s

4 4

e 1

e 4

I e

a i

6 5

. 4 l

4 1

February 1971

-C. Technical Assistance in Reactor Safety Analysis

. 1. . Highlights None

2. Technical' Activities Containment Resoonse Analvsis: Studies have.been initiated to compare calculations from the most recent version of the CONTEMPT-PS code with experimental" data from the Humboldt and BoNga Bay tests. - This version of the code differs from that described p' ously(2) in that the inlet fluid velocity to an expansion is used in a <pansion pressure drop calcula-tions. Previous versions of the cc de erroneously used outlet fluid velocities for this calculation.

Comparisons made to date are based on assumptions of early air carryover (all air in the- drywell is carried .into the wetwell before any water or steam leaves. the' drywell), .no heat transfer to the drywell or wetwell walls, and homogeneous flow in the vents. For these calculations, mass and energy input rates from the blowdown vessel into the drywell were calculated with the' RELAP3 code, because they were not measured experi-mentally in the tests. For these calculations, the predicted blowdown Pressure history and total blowdown time for the blowdown vessels were - -

r.atched as closely. as possible with the experimentally measured values.

For Humboldt Bay, -peak drywell pressures are underpredict5d by 10 to 25%, depending on the ratio of break-to-vent areas, when zero percent liquid carryover ~ is assumed. Overpredictions of 20 to 30% result when 100% liquid carryover is assumed.

For Bodega Bay, overpredictions.of about 10 to 40% in peak drywell pressures rest.lt for zero ' percent and 100% liquid carryover, respectively.

The greater ov, rprediction of drywell pressures for Bodega Bay (as compared to Hutcboldt Bay)is a result of the larger drywell volume and the overpredict_on of wetwell pressures due to the assumed early and

-complete air' carryover. The effect of the assumed mode of air carryover (early or homogeneous) on predicted containment behavior remains to be investigated.

t 5

5 -

- y g s .- , ,,- .w

, y ,

n- --

t.M .

j J

s .

L t February 1971

' ~

.D.' ' Separate Effects Testing 1~ '

Highlights

\ .l.-

-Semiscale Test;849 was conducted with' a high-inlet break configuration L

-with a nozzle..in place _of an~ orifice:in the bicwdown line. A quick-opening valve was installed in~ a bypass 'Line around the system low point to determine p ~

- whether pressure b'uildup due to a water seal was responsible in previous -

tests forl emergency coolant expu.sion-' through the. break. ECC ~1iquid was s

ejected. from the system!in Test'd49 as in previous tests and at no time did 4 ECC liquid. reach-the' core.

Semiscale Test 850:was' performed without ECC injection to.chtain

! = base-line data for comparison with data from a: previous test that' included emergency . cooling. inj ection. .

i 2.-

~

Technical Activities i

E Single-Loop Semiscale: The fifth semiscale test involving emergency core cooling (ECC), Test 849, was performed with a high inlet break (break area to system volume ratio of 0.007 f t-1) with a nozzle. in place of an

orifice in the blowdown line. The test was conducted with initial conditions of 2260 psig system pressure,,125 gpm fluid flow rate, 547 and 616cF fluid temperatures for the vessel inlet and outlet plenums, respectively, and a core power of 1.lLW.

and 1450F. Fluid conditions for the ECC: accumulator were 678 psig jl Ir addition.to a quick-closing block . valve in the ECC line to prevent i

flow 'of accumulator gas subsequen: to ECC-injection, a quick-opening

!- valve was tinstalled bstween the hot leg and pump suction leg to bypass the system low point (atf the bottom af the . steam generator) during the last part of ECC injection. This-' system modification was :to eliminate

~

r t he possibility of a pressure buildup preventing -ECC from reaching the core.being caused by steata ' generation back-pressure' from a water seal l or liquid ' trap at the system low point.

The' test was performed with E,CC injected directly into the inlet plenum.

4 i- Core power ~was terminated 6.5 seconds .after rupture and decompression was essentially complete 60 seconds'after rupture. The quick-closing ECC block

~

j qvalve and quick-opening bypass valve were~ a'ctuated simultaneously 36 seconds -1 after rupture when'the vessel; inlet plenum pressure had decreased'to'80-

.to.100. psig and ~af ter about 95% of the 'ECC liquid had been ijjected.

o .

Preliminary results1for Test 849 indicate that ae no time did the

![ ECC liquid reach the core. . A buildup offECC liquid o _urred in the inlet

' plenum but-was1 subsequently pushed _ from the vessel after ECC injection ,

t was' terminated (about 38 seconds Tafter rupture). Back pressure frem ,

a liquid trap at -the 'sy~ stem low point has been eliminated as the expulsion

? mechanism.

~~

l 'l '

I

'.e i

t'

'i

,- '6 -

4

=

% \

-y-- 4

-c; - , . .

/ 1 February 1971 i

-Figure 2- shows a comparison of t he vessel inlet nozzle pressure for Test .849,with that from.a prev 1ous te st which utilized an orifice in the blowdown line. Differences in the -pressure traces for the two tests  ;

reflect the higher , initial fluid -temperatures for-Test 849. The blowdown - .

time is- not significantly _ affected by the use of a nozzle in place of an-  ;

orifice. . Analysis to identify the location of the critical pressure

-plane (s) in the nozzle and to deter =ine the ratios of the inlet and outlet pressures for the nozzle -is in progr'ess.

An ' additional semiscale test, Test 850, was conducted without ECC l injection to obtain base-line data for conoa-ison with the results l ECC-injection Test 848 reported previously(II. Data reduction for this  !

, test is -in rcogress. I l

BWR-FLECHT: Analysis of data from the bottom flooding tests performed  !

' with a BWR-FLECRT stainless steel-clad bundle at pressures up to 300 psia  !

indicates that heat transfer efficiency for bottom flooding ECC increases j with increasing system pressure up to at least 200 psia. Figures 3 and 4 show die cladding temperatures and heat transfer-coefficients for tests conducted with essentially the same initial temperatures, bundle power levels, and flooding rates but with five' system pressures ranging from li to 300 psia. A monotonic increase in cooling effectiveness with increasing pressure 'is apparent up to 200 psia. The data show a slight decrease in cooling effectiveness upon further increase in pressure to 300 psia. '

PWR-FLECHT: .A partial explanation for the cause of the apparent increase in heat . transfer coefficient during the Zircaloy-clad bundle Test 8874(4) has been' found to be a result of an instrumentation error

+

in the experiment. Comparison of the heating rate with those of the other Zircaloy-clad bundle tests shows that the energy ~ input to Test

'8874 was 10 to 15% below test specifications. The difference between the energy input to the rod and the change in stored energy in the rod was used in the calculation of heat transfer coefficients. Since the

• specified energy input was used -in the calculation rather than the actual energy input, ex g sive- heat transfer' coefficients were calculated as reported earlier . Additional analysis, including re-calculation of the heat ' transfer coefficients, will be pertorned to determine the effect of _the experimental power input error.

A comparison of the cladding temperature response of a Zircaloy-clad -

bundle (Test 9573) with that of a stainless steel-clad bundle (Test 6553), i and with a pretest prediction was reported previously(2). The temperature response of ~the Zircaloy test was underpredicted when heat transfer coefficients--  ;

I 1

(4)- 'ucicar' Safety Program Monthly Report for September (Hai-384-70),

.0ctober 14,'1970.

7 4

4

1J iIl) \l I lll iIl1

?

.  ?

0

- - _ _ - _ 9 9

9 4*

7 1

B l

i 0 -

8 C N

l l

i 0 9 7 4 8

- d n

98 -

a 44 8 l

88 , i 0 4

, t t 8

. s s _ 6 e e s TT - t

- s

- e

- T

- e

~ l l - t 0 a

-- 5 c

- s i

m

) e

- c S

_ e g s

.- n i

i 0 (e i r

-=

- 4m i d

u

- T _

- s _

~ e _

l -

~ z z

I

~ I 0 o

~ 3 N

. . t

= e

- l

~

% n

' - I

. I

, l 8

I 0 e s

2 s

% e V

s t

,s a t \ e 8

t I 0 r g 1 u 1

\ s _

\ s _

r\ e .

r g

i P s

% 2 t t s\

8 t

. i

' 1 0 g

- i

- F

- - - 0

. 0 0 0 0 0 0 0 1

0 0 0 0-4 0

0 0 0 0 2 6 2 8 4 4 2 1 1

. ^ $ a e a*$ [ -

• . x ,; ,  ;'

mA O

~ D 3

1503 3 .o 2. .:=

• 6

• p ...... ..I....

i i i I -l -l I 15 osio

1. ,[. , . . N

• g%g,

. * %..,, * * . . .... 45 psio e \

~1250 - \ - - - - IOO psio

' ' * * . . * . . ----- * .200 osia

\.N %

Y ., .

,, **.g* ~*-- 300 psia x _

( . .

1000 - \

• I*l I }

+ i.

. g.

y I- 1 I.  !  :

-l  :

2 s-e-

1  :

le  : l l

3 11 5 i '

et Red 17,6 f t Esevotion l *g - i 500- Tg

• B3OO*FNominot l* .

• :* i l (.%

- 4 P 250 hW Peck l *

)

, % N.

%,,,p...,,,,.._.L-.-.~.-.---,,,,,.tr---------w t,

Ftooding Rotc :3 in./sec Cota Tco ,,,, = ISO 'F "*4'= % - 4 ,.,.......... ... '

250 -

0 I '

20 0 20 40 60 80 10 0 120 14 0 16 0 180 Time (sec) ,,c... ,,,,

Fig. 3 Temperature Behavior at the Midolane of a Central Rod as a Function of Press.ure i

GC' I j /I i I e

1. j l l

[

1 1 m -

I I-l s l

4

.e .

i 1

e

• 2 40 pl '

s

. .av .

15 p$io 3 .

)

~

i . -- 4 5 psio

-hc I *

.t g l 1.$

gi*.*

-*= 100 psia 1 30 - I ^'

• - -- 100 psio

/). .l*,

u 8

g i /j - - - 3 00 psio I st .* . .* /

$ @' . f Rod 17,6-f Ele sotion 3 20 -

gYl , ,. ; , #

~

r 1 T; 13OO *F P.sminal -

f f P a 250 hW beok I i ; *

' Flooding RJte t 3 irt/seC Cold to -

Tc,,,,,, s ISO CF t

(

t" 5 o- I- I l l '

l l l l

-20 0 20 {

40 60 80 10 0 120 140 160 18 0 Time (sec) ,ne. .,,,,,

Fig. 4_Iteat Transfer Coefficients for 'the Midplane of a Centr.tl ,

RodasaFunctignofPressure '

.L ,

Febru ry 1970 obtained from the stainless steel' test were used .for predictions. Metallo-was not excessive ~(about 50% of the amount predicted by the B equation) - and does not account for the higher temperatures. Figures 5 and-6 show comparisons of :the heat transfer coefficients determined from data

'for. Zircaloy. bundle Test 9573 and stainless steel bundle Test 6553 at . the _

2 , 4 , 6 , 8 , and 10-f t bundle elevations.

The heat transfer coefficients for. the Zircaloy-clad' bundle are lower at all elevations by about 10%.

Increased steam- superheating and decreased liquid entrain =ent .in the

' Zircaloy-cla.i bundle test appear to be- the causes. of the lower heat transfer coefficients. .

The cladding temperature increases with elevation up to the bundle

- midplane; but 'since the fluid is heated by the rod the fluid temperature.

can not be higher thon the cladding temperature at the. midplane. Negative heat transfer coefficients represent conditions - for which the steam temperatures are higher than the cladding teeperatures. These conditions 1cna flooding rates (Figure 5).are expected at the higher bundle elevations (8 and 1 The calculated heat transfer coefficient

. at the 6-f t elevation for the Zircaloy-clad bundle (Figure 6) does, however, show a negative value for the period from 6 to 12 seconds

'after initiation of flooding. Possible explanations for the negative

-heat transfer coefficient calculated include numerical errors and-inaccurate assumptions of metal-water reaction energy input to the computer program which- calculates the heat transfer coefficients.

i (5) . Louis naKer, Jr. , and Louis C. Just, i

\

at High Temperatures, ANL-6548 (May 1962). -Studies of Metal-Water Reaction i i

i

, 10 t

m 1.

-- ' 20 i . .i. i 1. 1. .

-l Zircoloy Test 9573 -

• ~

- 15

--- Stainless Steel Test 6553 *

, Flooding Rote = 1in,/sec is. -

3 *- -

Ti = 2OOO 'F

~

"e 10 t

E s

i 2

5 5 --

s '

= ~4 - - - ~ ~ .,_,_ _____

3  % :: , - - - _ 8 ft Elevation s*

3 0- ,  !

u  %

3 \ <

t 's,' - - ~ ~ __ i 1

= ~ ~ ~ - - - - - ~ ~ - '

1. -5 - -

3 IO f t Elevation g

Rod Failures Begin Test 9573 ,

j t

f f 1 1 '* *

-15 0 tO 20 30 4O 5O 60 Time of f er Flooding (sec) ,,c...,,,,,

Fig. 5 PWR-FLECHT Heat Transfer Coefficients Above the Midplans

• 30 8 i . . .

I8J l I ,

t I ,

Zircoley Test 9573 25 -

I .

I ---

t ntess Steel Test 6553 h I 2-f t Elevo' ion "c 2 0- s'

• Flooding Rote = lindsec ~

Ti= 2OOO *F s

$ g-2 I ,

D 1 5e 15 -

i, I i\ -

4-f t Elevation __. ..----

j g g 2: # # s

.k 8

u so .- ,./

8 1

\

?o ll - -......

---

f p ,e fi ,es' 6-f t Elevation

-:.,.o// ,

.5 -

x m -

s- -

,. n . Rod Foilure s -

• Begin Test 9573

.,' '5

.* O - 10 . 20 30 40 50 60-Time of ter Flooding (sec) 3 Fig. . 6 PWR-FLECllT Heat Tran.ifer Coefficients at and Below the M'[$~pl,a,n,,e 11

. _ _ ._