ML19341B399

RELAP4/MOD7 Nodalization & Heat Transfer Sensitivity Calculations for Zion Plant During Small Break LOCA, Assessment Rept

ML19341B399
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	12/31/1980
From:	Lin J EG&G IDAHO, INC., EG&G, INC.
To:	Guttmann J, Odar F Office of Nuclear Reactor Regulation, NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
CON-FIN-A-6047 EGG-CAAP-5320, NUDOCS 8102020053
Download: ML19341B399 (50)

Text

hEGcG..,~.,,.

FORM EG&G 396 inov i1796 INTERIM REPORT Accession No.

4 Report No.

EGG-CAAP-5320 Contract Program or Project

Title:

Code Assessment and Applications Division Subject of this Document: RELAP4/ MOD 7 Nodalization and Heat Transfer Sensitivity Calculations for Zion Plant During Small Break LOCA Type of Document: Assessraent Report Author (s):

J. C. Lin Date of Document:

December 1980 Responsible NRC Individual and NRC Office or Division:

J. Guttmann, NRC-NRR and F. Odar, NRC-RSR I

This document was prepared primarily for preliminary or internal use. It has not received f ull review and approval. Since there may be substantive changes, this documant should l

not be considered final.

EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 i

i 4

Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

l e

Under DOE Contract No. DE AC07-761D01570 NRC FIN No. A6047 j

l lNTERIM REPORT -

8/02.o2.0053 EC Researc1 anc "ec1nica l

A Assistance Report

A3STRACT Core nodalization and neat transfer sensitivity studies were performed for tne RELAP4/MJD7 iermal-nydraulic computer code during a small creak loss-of-coolant accident.

Tne results snowed that the system nydraulic response was almost insensitive to tne core nodalization and critical heat flux (CHF) models. Tne sensitivity of tne fuel rod surface temperature to tne core nodalization and CHF neat transfer was small. From tnis analysis it is recommended that a three-volume basic model with tne Hsu-Beckner critical neat flux model be used for small break LOCA analyses.

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ii

CONIENIS ii ABSTRACT............................................................

Vi

SUMMARY

1.

INTRODUCTION.............................................

1 2.

MODEL DESCRIPTION..............

2 2.1 Code Description.......

2 2

2.2 Nodalizations 2.3 Heat Transfer Models.....................................

3 4

2.4 Options Selection...............

2.5 Initial and Boundary Conditions...............

5 3.

RESULTS........................................................

7 3.1 Core Nodalization Sensitivity...

7 3.2 Core Heat Transfer Sensitivity............................

8 4.

CONCLUSIONS AND RECOMMENDATIONS................................ 10 5.

REFERENCES...................................................... 12 e

Y e

iii

FIGURES 1.

Nodalization for the three-volume model...........

16 2.

Core nodalization for the six-even-volume model...

17 3.

Core nodalization for the six-uneven-volume model...............

18 4

Core nodalization for the twelve-volume model...................

19 5.

Comparison of the Mass flow in the intact loop hot leg for for the four core nodalization models...........................

20 6.

Comparison of the mass flow in the intact loop cold leg for the four core nodalization models...........................

21 7.

Comparison of the mass flow in the broken loop hot leg for the four core nodalization models...........................

22 8.

Comparison of the mass flow in the broken loop cold leg for the four core nodalization models...........................

23 9.

Comparison of the mass flow in the core inlet for the four core nodalization models...........................

24 10.

Comparison of the mass flow in accumulator I for the four core nodalization models...........................

25

11. Comparison of the mass flow in accumulator II for the four core nodalization models...........................

26 12.

Pressure in tne upper plenum for the four core nodalization models..........................................................

27

13. Comparison of the mixture level in the upper one-third of the core for the four core nodalization models..................

28

14. Comparison of the mixture level in the downcomer for the four core nodalization models...........................

29

15. Fuel rod surface temperature of the upper one-third of the average rod for the four core nodalization models........

30

16. Comparison of the mass flow in the intact loop hot leg for the two CHF models..........................................

31 17.

Comparison of the mass flow in the intact loop cold leg for the two CHF models..........................................

32 iv

18. Comparison of the mass flow in the broken loop hot leg for t h e t wo C HF mod e l s..........................................

33

19. Comparison of the mais flow in the broken loop cold leg f o r t h e t wo C HF mod e l s..........................................

34

=

20.

Comparison of the mass flow in the core inlet f or t h e two CHF mod e l s..........................................

35

21. Comparison of the mass flow in accumulator I f or t h e tw o CHF mod e l s..........................................

36

22. Comparison of the mass flow in accumulator II f or t h e two CHF mod e l s.........................................

37

23. Pressure in the upper plenum for the two CHF models.............

38 24 Comparison of the mixture level in the upper one-third of the core for the two CHF models.................................

39 25.

Comparison of the mixture level in the downcomer for the two CHF modeJs..................................................

40

26. Fuel rod surface temperature of the upper one third of the average rod for the two CHF models..............................

41 27 Heat transfer coefficient for the upper one third of the average rod for the two CHF models..............................

42 I

s l

V-

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TABLES 4

l 1.

Normalized Axial Power Profile for EM Basic Case..............., 13 2.

HPI Pump Flow..........................,,,,,,,,,,,,,,,,,,,,,,,,, 14 3.

Inpu t Deck Con f i gura t ion Con tro l Number......................... 15 i

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SUMMARY

Ine RELAP4/M001 thermal-nydraulic computer code was used to calculate tne thermal-nydraulic benavior for a pressurized water reactor (PWR) during a

a small Dreak loss-of-coolant accidents (LOCAs). Core nodalization and neat transfer sensitivity studies were performed and tne results of tnese studies are discussed in tnis report. Tne objective of these sensitivity studies was to find an optimized RELAP4/M007 core nodalization and neat transfer model to be used in future analyses.

Six models were developed for the sensitivity studies (four for core nodalization and two for core ner.t transfer sensitivities). The results show tnat the system hydraulic response was almost insensitive'to the core nodalization and core critical neat flux (CHF) models. The sensitivity of the fuel rod surface temperature to the core nodalization and CHF models was small.

Tne three-volume basic model witn tne Hsu-Beckner CHF model is recommended for small break LOCA analyses in a PWP,,

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vii

1.

INTRODUCTION The RELAP4/M007 tnermal-hydraulic computer code was used to calculate tne thermal-hydraulic benavior for a pressurized water reactor (PWR) during a small break loss-af-coolant accidents (LOCAs) in Reference ?.

Core nndalization and heat transfer sensitivity sterlies were not performed in Reference 2.

Sensitivity studies were performed and discussed in this recort. The obiective of tnase studies was to find an optimized RELAP4/ MOD 7 core nodalization and heat transfer model to be used in future analyses, in tnis report Section 7 prosents tne nndel descriptions. Section 3 shows the results ontained from the core nod.lization and neat transfer sensitivity studies. Section 4 presents tne conclusions and recommendations.

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1

2.

MODEL DESCRIPTION This section describes the RELAP4/M007 computer code, nodalizations, and heat transfer models. This section also presents the option selection and initial and boundary conditions.

?.1 Code Description RELAP4 is a computer code developed to describe the thermal-hydraulic behavior of a light water reactor (LWR) subjected to pnstulated accidents such as those resulting from loss-of-coolant, pump failure or nuclear power excursion. Fundamental assumptions inherent in the thermal-hydraulics are that the two-phase flow is homogeneous and that the two phases are in thermal equilibrium. Models are available in the code to modify these homogeneous assumptions. The program is sufficiently general to be applied to experimental water reactor simulators and other thermal-hydraulic experiments.

RELAP4/ MOD 7 (Reference 1) incorporates a thermal non-equilibrium model into the RELAP4 code to modify the thermal equilibrium assumption such that the code can be used to calculate the condensation phenomena which occurs during the ECC water injection period. Major improvements in RELAP4/ MOD 7 3

are given in Reference 1.

In this analysis, the RELAP4/M007 Update 92 was used.

2.2 Nodalizations The RELAP4/M007 oc-del rif the Zion thermal-hydraulic system was defined with close correspondence to the actual primary system of the plant. A i

schematic of the Zion system basic model is shown in Figure 1.

The model i

RELAP4/M007. Update 92, Idaho National Engineering Laborattry a.

Configuration Control Number H007184B.

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m

consists of 40 control volumes and 49 junctions.

The actual geometric data were used for the control voluces. The calculated loss coefficients were used for the junctions. Measured p'Jmp performance data were input for the pump model.

a A one-dimensional heat conductor (heat slab) model is used in RELAP4 to calculate heat transfer from metal walls and fuel rods to the fluid. To model the fuel rod, six heat slabs were used in the system model.

To study the effect of the core nodalization on the system thermal-hydraul.c response, three models were developed. These models are similar to the basic model shown in Figure 1 except for the core region.

The first model, as shown in Figure 2, simulated the core region with six evenly distributed volumes. The fuel rods were modeled using six heat slabs. The second model, as shown in Figure 3, simulated the core region with six unevenly distributed volumes. The fuel rods were modeled using 8 heat slabs. The third model, which is shown in Figure 4, simulated the core region by 12 volumes (two parallel vertical stacks of 6 volumes). The fuel rods were modeled using 18 heat slabs.

The " average" core channe,

which consisted of 192 fuel assemblies, was modeled by a vertical stac8 of 6 heat slabs (slab 1 through 6). The " hot" core channel was simulated by two separate stacks of 6 heat slabs. One stack of heat slabs represented the average rods in the hot assembly (slab 7 through 12). The other stack represented the hot rod in the hot assembly (slab 13 through 18).

2.3 Heat Transfer Models For small break LOCA, it was shown in Reference i that the fuel rod surface temperature increased rapidly during the period of core ~

this analysis, calculations to investigate the sensi' ity uncovering, t..

of the fuel rod surface temperature to the critical heat flux (CHF) models were performed.

Two CHF madels were used:

i l

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3 L

1.

Tne CHF correlations of Hsu-Beckner for hign flow and modiflad c

Zuber" for low flov.

2.

Tne CHF correlation developed in RELAP4/M007 (Reference 1).

It snould be noted tnat the model of six evenly distributed Core volumes was used for tne neat transfer sensitivity study.

2.4 Options Selection Tnere are many hydraulic options in tne RELAP4/M007 code. Tne options used in this analysis were similar to tnose used in Reference 2 and are given as follows:

1.

Compressible flow with momentum flux (MVMIX = 0) was used at all junctions, except at junctions between the vessel and hot or cold legs, pressurizer and accumulator junctions, core bypass patns and fill junctions where the incompressible flow without momentum flux (MVMIX = 3) was used.

2.

Wilson bubble rise was used in the downcomer, downcomer annulus, lower plenum, upper plenum and naad, and pump suction volumes. A bubble gradient of 0.8 was used.

3.

Complete pnase separation was used in the accumulator volumes.

4.

A constant nuonle rise velocity and bubble gradient were used in the steam generator secondaries. Tne value for the bucole rise velocity and tne buoble gradient were calculated oy tne code from an initial energy oalance.

5.

Tne lienry-Fauske and Moody critical flow models were used for subcoole-old saturated regions, respectively.

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6.

Pump Dearing friction equal to 2.5 parcent of rated torque was used.

7.

Tne RELAP4/M007 self-initialization routine was used to initialize tne system. During tne initialization, the enthalpy transport model was used but it was not used during tne transient.

8.

The vertical slip model was used in all calculations at junctions netween the inlet annulus and downcomer, downcomer and lower plenum, lower plenum and core, between core volumes, and between core and upper plenum volumes.

2.5 Initial and Boundary Conditions Tne initial and boundary conditions were similar to those specified in Reference 2.

Tne conditions were:

1.

Normalized axial power profile is given in Taole 1.

2.

ANS + 20 percent decay heat.

3.

Total safety injection flow vs. pressure for four loops is given in Table 2.

4.

Steam generator safety valve full rated flow (FRF) was 524.3 kg/s per steam generator. Tne valve opened at 9.207 MPa allowing 8 percent FRF to pass and the flow rate increased linearly to 100 percent FRF at 8.62 MPa.

5.

The feedwater full rated flow was 509.2 kg/s per steam generator. Feedwater remained at full flow until 5 s af ter SCRAM and remained at 0 until 59 s after SCRAM. Auxilitry feedwater was turned on linearly from 50 to 60 s after SCRAM and remained on at 1.928% of tne full flow. Maia feedwater entnalpy was 975.8 kJ/kg and the auxiliary feedvater enthalpy was 212.13 kJ/kg.

5

6.

Accumulators

/

l P

= 4. 38 MPa init T = 325 K 3

Initial water volume = 23 m i

7.

Scram occurred 3.4 s af ter scram signal was received when tne pressurizer blew down to 12.8?7 MPa. Pumps were tripped off upon reactor scram.

P.

Pressurizer initial mass was 21869 kg of wnicn 2140 kg was vapor.

9.

Turbine trip coincirient with SCRAM signal closed the steam generator outlet valves.

10.

System operating conditions Mass flow rate per loop :

4428 kg/s Upper plenum pressure 15.76 MPa S. G. secondary pressure -

6.9 MPa Cold leg te.nperature 569.3 K Hot leg temperature 604 K Core power 3649 MW Input decks for each calculation are stored at INEL under Configuration Control as detailed in Table.3.

6 4

6

.. -.,, - - +

A 3.

RESULTS T

i This section presents the results obtained from the core nodalization and heat transfer sensitivity studies.

3.1 Core Nodalization Sensitivity As discussed in previous sections, four core nodalization models (three-volume, six-even-volume, six-uneven-volume and twelve-volume models) were developed for the core nodalization sensitivity study. A comparison j

of the important parameters for these cases is shown in the following sections.

Figures 5 through 9 present the mass flow in the intact loop hot leg, intact loop cold leg, broken loop hot leg, broken loop cold leg and core inlet, respectively. The mass flow calculated by the four models was almost identical until the accumulator injection was actuated. A comparison of the mass flow in the accunolator for the four models is1shown in Figures 10 and 11 and the results were similar to those for the loop a.id core mass flows. The 'ime for accumulatur injection initiation for these cases were 961, 967, 981 and 990 s for the twelve-volume, six-uneven volume, six-even volume, and three-volume models, respectively. This observation is consistent with the upper plenum pressure comparison as shown in-Figure 12 which shows some difference in system pressure after-about 700 s.

TheLdepressurization rate for the four cases was almost the same until about 700 s when the core uncovered.

Figure 13 shows a comparison of dxture level in the~ upper one third:

of the core. The core uncovering rate for the twelve volume case was the f astest of the four models : investigated and the core uncovering rate. for the th-ce volume ct te was the slowest.of the four models. The core uncovaring rate frr the six-uneven volume case was faster than that for the six-even volume case at the-beginning of the core uncc ering period.

However, during the latter part of the' core uncovering, the' core uncovering 7

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rate for the six-uneven volume case was slower than that for the six-even volume case. Similar comparison of the downcomer mixture-level for the four cases is shown in Figure 14. The mixture level remained at a constant i

level (5.547 m) except for the three volume case in which the downcomer mixture level dropped to 4.115 m and then rose to 5.547 m between 770 and 900 s.

This dropping ot downcomer liquid level for the three volume case

~

was caused by ater packing of the top core volume when the core. uncovering occurred. The mixture level smoothing option was not turred on until 900 s i

for the basic model. Therefore, water packing before 900 s was not j

smoothed. The pressure disturbance caused by the water packing propagated to the downcomer and caused the mixture level in the downcomer to drop.

t Figure 15 shows the fuel rod surface temperature for.the upper one third of the core for the four cases. The fuel red surface. temperatures i

were identical until the core uncovering occurred. For all the four case, j

during the fuel uncovering period, the fuel rod surface temperature j

increased almost adiabatically. The fuel rod surface temperature started to decrease shortly after accumulator injection was initiated. The fuel j

rod surf ace. temperature decreased to the saturation temperature l

9rresponding to the system pressure when the upper part of the core was refilled with-liquid. The maximum difference in the turn around temperature for these cases was about 56 K.

i, 3.2 Core Heat Transfer Sensitivity

~

l l

As discussed previously, two CHF correlations (Hsu-Beckner and MOD 7).

-were used to study the effect of the CHF heat transfer corre'ation on the system thermal-hydraulic _ response. Figures'16 through 20 show.the mass flow in the hot and cold legs of the intact and broken loops and in the core inlet. The mass flow for'the two cases was almost identical until injection.of ECC water was' initiated. A comparison of the accumulator' flow for these two cases is shown in Figures 21 and 22. The accumulator flow.

for the M007 case was initiated earlier than for the Hsu-Beckner case.

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Figure 23 presents the upper plenum pressure calculated with the M007 and Hsu-Beckner CHF models. The system for these two cases depressurized at almost the same rate until 700 s.

After 700 s the depressurization rate for the M007 case was somewhat faster than that for Hsu-Beckner. The upper plenum pressure with the M007 model reached the accumulator set point The-pressure, 4.138 MPa, a little earlier than for the Hsu-Beckner.

pressure transients were consistent with the initiation of the accumulator flow as was previously discussed.

Figure 24 shows a comparison of the mixture level calculated with the Hsu-Beckner and M007 CHF models for the upper one third f the core. The

~

core uncoverirg rate for the Hsu-Beckner model was faster than that with the M007 model and th core uncovering for the Hsu-Beckner CHF model-occurred somewhat earlier than occurred with the ROD 7 CHF model. A comparison of mixture level in the downcomer is shown in Figure 25. The dowacomer mixture level for the two cases remained a. bout full through the entire blowdown transient.

Figure 26 shows the fuel rod surface temperature calculated for the Hsu-Beckner and MOD 7 CHF models for the upper one third of the core. The fuel rod surface temperature calculated by the two models was the same-until core uncovering occurred. After the core'was uncovered, the fuel rod was heated up almost adiabatically. When accumulator water injection was initiated, the fuel rod temperature started t0 decrease to the saturation temperature corresponding to the core pressure. The. difference in the maximum turn around temperature calculated by the two models was-about 30 K.

i.gure 27 shows the heat transfer coefficient calculated by the Hsu-Beckner and M007 CHF models for1the upper one-third lof the core. The hect transfer coefficients calculated by the two models were-almost the same until the core uncovering occurred. After the core was uncovered, the heat transfer coefficient dropped drastically from about 142 to:

0.06 kW/m -K and was kept at a low heat transfer coefficient until Jthe accumulator injection was initiated.

9

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4.

CONCLUoIONS AND REC 0ffiENUATIONS The conclusions for the core nodalization sensitivity studies were:

l.

The system hydmulic resporwe v<w insensitive to the cora nodalisation.

!iowever, the core nodalization had a small effect on the system pressure response after the core was uncovered.

This effect caused a small difference in the time of accumulator injection initiation. The maximum difference in the time of accumulator injection initiation between the fine and coarse nodalizations was about 30 s.

a.

The core nadalization had some effect on the core uncovering rate.

The core uncovering rate for the fine nodalization was faster than for the coarse nodalization.

3.

The douncomer mi.rture Ecual was insensitive to the core nodalisation.

However, it was quite sensitive to the use of the mixture level smoothing option.

4.

The core nodalisation had an effect on the fuel rod themal response. The maximum difference in the turnaround temperature between fine and coarse nodalization was abcut 56 K.

The conclusions for the core heat transfer sensitivity study were:

L.

The system hydraulic response was incensitive to the CHF heat transfer models used in the core. However, the CHF heat transfer models used in the core had a small effect on the system pressure response after the core was uncovered. This effect caused a small difference in the time when accumulator injection was initiated.

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2. ' The core uncovery rate vas censitive to the core CHF heat tmnsfer model.

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The douncemer mixture Icoat uas incerwitive to the core CHF heat-transfer mbiet.

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-The CHF heat transfer model used in the core had some effect on I

the fuel rod themal response. The maximum difference in the turnaround temperature between the Hsu-Beckner and MOD 7 CHF heat transfer models was about 30 K.

In general, the computing time for the fine nodalization.was about five times longer than that for the coarse nodalization. The computing _

time with the MOD 7 CHF heat transfer was about' twice that with the Hsu-Beckner CHF-heat transfer model.

i L

From the above conclusions, the effect of core nodalization and heat transfer model on the system thennal-hydraulic response is small. To save computing time, it.is reconinended that the three-volume basic model with the Hsu-Beckner CHF correlation for the small break LOCA analyses should be used..The mixture -level smoothing option should be; turned on before the core uncovering occurs.

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REFERENCES 1.

G. W. Jonnsen et al., RELAP4/M007 (Version 7) Users Manual, EG&G Idano, Inc., CDAP-TR-78-036, August 1978.

2.

C. D. Fletcher, Additional Audit Calculations for Westingnouse PWR Small 3reaks, EG&G Idaho, Inc., EGG-CAAP-5052, November 1979.

3.

K. V. Moore and W. H. Rettig, RELAP4 - A Computer Program for Transient Thermal-Hydraulic Analysis, ANCR-ll27, March 1975.

i 4.

Y. Y. Hsu and W. D. Beckner, "A Correlation for tne Onset of Transient CHF," cited in L. S. Tong and G. L. Bennett, "NRC Water Reactor Safety Researcn Program," Nuclear Safety, 18. 1 (January / February 1977).

S.

N. Zuber, " Hydrodynamic Aspects of Boiling Heat Transfer," USAEC Report AECU-4439 (1959).

(PnD Inesis, University of California, Los Angeles, 1959).

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TABLE 1.

NORMALIZED AXIAL POWER PROFILE FOR EM BASIC CASE r

Top t

0.16718 0.22200 0.18383 1

0.15733 0.14783 0.12183 4

Bottom i

a a

4 1

1 i

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TABLE 2.

H?1 PUMP FLOW Pressure Flow (MPa)

(m /ne) 0.0 16110 0.24 13662 O.38 11280 0.62 8832 0.72 6146 0.88 3408 1.48 3284 2.86 2994 4.24 2677 5.62 2360 7.00 1977 8.38 1563 9.07 1259 9.97 542 11.14 414 12.62

?24 14.31 0

17.24 0

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TABLE 3.

INPUT DECK CONFIGURATION CONTROL NUMBER INEL Configuration Run Control Number RELAP4/ MOD 7 six-uneven-volumes H013485B RELAP4/M007 six-even-volumes H013585B RELAP4/ MOD 7 Hsu-Beckner Model H0136858 RELAP4/ MOD 7 M007 CHF Model H013785B RELAP4/M007 12-Volume Model H013885B O

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Nodalization for the three-volume model.

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200 400 600 000 1000 Tim.

tsi Figure 7.

Comparison of the mass flow in the broken loop hot leg for the four core nodalization models.

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Figure 8.

Comparison of the mass flow in the broken loop cold leg for the four core nodalization models.

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6-UNEVEN CORE VOLUME A

6-EVEN CORE VOLUME o

12-CORE VOLUME 10000

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0 200 400 600 800 1000 Time (si Figure 9.

Comparison of the mass flow in the core inlet for th

'our core nodalization models.

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= :

r =r

m I

? e f 0 200 400 6r 0 000 1000 Trm. tsi Figure 11. Comparison of the mass flow in accumulavor 11 for the four core nodalization models. f l l O O (f -o t b Y c aw o .J E

• e O3

+2 ->s e ow O N CL E 3 O c)w>r 4 o w _J g eo -uao O o> C Zu ww w O v1>Z1 h wwwo u mz>u Dwe L E i iN o waae O O Ox40 O o E a +* e L ) ~~ O c 3 E g i t -c. i O L O, g. o o \\. 3 l > f a < i j g +' C r= o I L J 3 O in O g N L c. 1 m o L 3 C1 u t i O O 'O O O O O O O O O O O O O LD O LD e W gggy3 sanssesy 27 + + 2.0 O EM BASE IML33: X 6-UNEVEN CORE VOLUME a 6-EVEN CORE VOLUME o 12-CORE VOLUME 1.5 r N w w " p w-w W N ~ W f c i 0 1.0 e e w r 0.5 [ \\ l l o k _. 0.0 0 200 400 603 8CO 1000 Ttm. tI Figure 13. Comparison of the mixture level in the upper one-third of the core for the four core nodalization models. 1 l I 1 1 1 -~ 4 k 4 i 4 1 i j 6.0 o EM BASE IML3Si X b-UNEVEN CORE VOLUME A 6-EVEN CORE V O L t' ME i o 12-CORE VOLUME = = 2 ==

==

rA ] .E I C 0 4 5.0 i to e ~ w. I = ~ 4.5 I a,o O 200 400 600 800 1000 Time Isl l Figure 14. Comparison of the mixture level in the downcomer for the four i core nodalization models. 4 1500 i O EM BASE ISR5 1 X 6-UNEVEN CORE VOLUME j A 6-EVEN CORE VOLUME o 12-CORE VOLUME 1300 x o L 3 1100 i L 5 Il e e ~ t' 900 1 + / oe I / 3 700 Y A J _j r n-- 2 500 O 200 400 600 80C COD T, as Figure 15. Fuel rod surface temperature of the upper one-third of the average rod for the four core nodalization models. l l l t -. _ ~. + i 4-20000 i e i i i o MOD 7 CHF MODEL 4JW41 j r HSU-BECKNER CHF MODEL 15000 w s j 10000 3 4 E 5000 -. e c .I 2 ~ 0 2 = ^ = = r -5000 O 200 400 600 800 1000 Time isi Figure 16. Comparison of the mass flow in the intact loop hot leg for the two CHF models. l llJ I ll, l 0 0 0 v 1 ,b 7 r 2L o f WE JD g 1 O e M l 0 LF 0 d i EH 8 l DC o c O MR p E o rN o HF l CC E t 7 D ca t DU n OS 0 i MH 0 6 e h i t Om s n ( i r e w m o l i f T ss 0 a 2 0 m. 4 s el he = td o f m o = F nH oC s io = rw at 0 p i 2 0 me 2 oh Ct ~ 7 i eru g i F O 0 0 i 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 4 2 2 4 .e%e x-20 ' e@oE t ym

i

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si!

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i

g i 2 i I i 3 6000 f O MOD 7 CHF MODEL 4JW40s I x HSU-BECKNER CHF MODEL t I th 4000 e 1 ~ cn i .x j -2000 w { l' W e ,e 4 E 3 0 ^ l t i i i - 2000 0 200 400 600 800 1000 i Time ts) 3 Figure 18.. Comparison of the mass flow in the broken loop hot leg for r the two CHF models. j t an m ...--.--..-m.. .. u - m.m m 10000 i a i i o MOD 7 CHr MODEL (JW201 ) HSU-BECkNER CHF MODEL 5000 o e w .e ~ l .g j 0 = 2 gi, e 61o 'r -5000 -10000 O 200 400 600 800 1030 l Time ist Figure 19. Comparison of the mass flow in the broken loop cold leg for the two CHF models. w i n n. i i 30000 i i o MOD 7 CHF MODLL isw30i x H S U - D E C k N E f4 Car MODE, 20000-- l e j 10000 I O __, o it. 1.a + - ri iT P r i r--"- ~ 0 E

1. e l

r -10000 -20000 O 200 400 600 800 1000 Time Isi Figure 20. Comparison of the mass flow in the core inle : for the two CHF models. 4 + 1 t i ,i 2000 i i t o MOD 7 CHr MODEL tJW248 !~ x HSU-DECKNER CHF MODEL I i 4 i, 1000 s t m /, l x f 2 o L w f

• e-i 1

e e i e C: 2 = = = = = = = = = = = = I i -1000 f 0 200 400 600 800 1000 l-Ttme ts) ~ Figure 21. Comparison of the mass flow in irccumulator I for the two CHF models. + i + i 4 f. I 4 g 4 h 8 i 9 6 i ? f q i + 800 i i o MOD 7 CHF MODEL i )W25' X H SU - B E C x t4E R C+d MODEL ( 600 - i I t e s j 400 ) 1 W O w N + ~200 e C I s O' ^' ^ ~ ^ ~ ^ 1 -200 5 -0 200 400-600 800 100; T:me ist i Figure 22. Comparison of the mass flow in accumulator II i for the two CHF models. r + i 4 O O J O / e ..f e 8 J W p sf =t G -a 1 r C O F ll O ~ w : C 9 e ll r= u. u o5 a k O Q OD .C OO O L: ~ O L O O-E 3 C O m E ~ L t-Qa a o l On 0 C .1 +3

r 4

C e hL 3m M Q O a' O N m N e U L 1 3 t G w b l t l O O O 'O O O o O O O O O O O O O O O O m O to N N e v edy3 a ; ri s s o a g 38 _... _ = l e i L 3 I s i I 2.5 i O MOD 7 CHF MODEL 4ML33i X HSU-dECKNER CHF MODEL i 2. 0: 2 c E 1.5 - L c \\ .? we e I 1.0 i i w v, 1 I . ) o.5 i / i I d9 0.0 0 200 400 600 800 1000 j. T;me (s' i Figure 24. Comparison of the mixture level in the upper one-third of the i core for the two CHF models. i 4 5.7 i i i o MOD 7 CHF MODEL e Mt 35 : x HSU-BE Cc NE A CHT MODEL 5.6 j i~ in w, ~ ~ ~ g j 7 , g. Ff T c I o R 5.5 I e w 5.4 5.3 D 200 400 600 800 1000 T: me tsi Figure 25 Comparison of the mixture level in the downcomer for the two CHF model;. + + T g by W + C C s . O. -. _ + a - - ~ ' <s,'-_ e \\ 'k %% '-- I ~, T ' 4, N 's v =.,. _ 4., x_ e s _ss ~- N s ' ~ h h x\\, h ~ s: c x ~ IG Q L O I2 9. Q u-G A N 3 C3 y O a E4 O w O Q Oa o L S 3 ~ sa C e* Q E e n 5 a O O u O e v w L 3 e VO L h 3 LA. O O c4 c N Q L 3 O -e b + Q O O O O O O O O O O O O O O o e m 5 4 0 4 gy) aangesadwal a3eJan$ o 41 + t s P i I 80 i i i i i o MOD 7 CHF MODEL tCASI x HSU-DECKNER CHF MODft t C i e 60 ~ v s C f 1 e -- O Y 8 u .Sm c 40 e,- 0 i. 2 e4 c-L y e 20 ~ e ] 1 I w ^ =^ - O t 0 200 400 600 800 1000 IIme iSI t t . Figure 27. Heat transfer coefficient for the upper one third of the l average rod for the two CHF rnodels. i +. i + ) i s ,,...m._ ._._.,.x.__.. _,