Text

Relap 4 MOD 7 Commercial PWR Large Break Transient Analysis Compared W/Loft Data
	ML20037D280
Person / Time
Site:	Zion File:ZionSolutions icon.png
Issue date:	06/30/1981
From:	Deboer T; EG&G, INC.
To:	Mcpherson G; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
	CON-FIN-A-6048 EEG-LOFT-5401, EGG-LOFT-5401, NUDOCS 8107090237
	Download: ML20037D280 (60)
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I EGG-LOFT-5401 Project No. P 394 June 1981 RELAP4/ MOD 7 COMMERCIAL PWR LARGE BREAK TRANSIENT ANALYSIS COMPARED WITH LOFT DATA N30!leSC7Ch805 IOChiliCal Assistance lleport /

Teun C. deBoer A

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's This is an informal report intended for use as a preliminary or working document s

Prepared for the U.S. Nuclear Regulatory Comission Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6048 U

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8107090237 810630 PDR RES 8107090237 PDR

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FORM EG&G 398 (Rev.114 INTERIM REPORT Accession No.

EGG-LOFT-5401 Report No.

Contract Program or Project

Title:

aesearca and Tecmica LOFT Program Office N Assistance Report /

Subject of this Document:

RELAP4/ MOD 7 Commercial PWR Large Break Transient Analysis Compared with LOFT Data Type of Document:

Topical Report Author (s):

Teun C. deBoer o

Date of Document:

June 1981 R:sponsible NRC Individual and NRC Office or Division:

G. D. McPherson, Chief, LOFT Research Branch, Division of Reactor Safety Research, USMC This document was prepared primarily for preliminary orinternal use. it has not received full review and approval. Since there may be substantive changes, this document should not be considered final.

EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 g

Prepared for the U.S. Nuclea'r Regulatory Commission Washington, D.C.

Under DOE Contract No. DE-AC07 76tD01570 o

NRC FIN No. A6048 l

lNTERIM REPORT l

ABSTRACT Data from RELAP4'/ MOD 6 and RELAP4/M007 calculations simulating hypothet-ical, cold leg large break (200% double-ended break) loss-of-coolant acci-dents in the Zion 1 nuclear power plant are compared and evaluated.

The Zion 1 large break calculation using RELAP4/M006 in a previous Loss-of-Fluid Test (LOFT) prototypicality study predicted peak cladding temperature but notthereturntonucleateh$ilingthatoccurredduringtheLOFTlargebreak Experiment L2-3 blowdown.

The large break calculations for Zion 1 in this study using the RELAP4/ MOD 7 computer code with other heat transfer correla-tiens than were used in RELAP4/ MOD 6 and a different core modeling technique, did precict the return to nucleate boiling even at the hottest spot on the not fuel rod. This early rewet phenomenon was also calculated for the more conservative RELAP4/ MOD 7 calculations with 102.4% nominal power, a maximum linear heat generation rate of 50 kW/m, 93% nominal flow, and a more con-servative fuel rod model. Results from the RELAP4/M007 calculation for 100%

nominal power showed similar peak cladding temperature and quenching behavior compared with LOFT Experiment L2-3 data.

The study showed that early rewet occurred with nominal and conservative core hydraulic conditions and fuel peak linear heat generation rates that bracket current commercial plant operating conditions.

The early rewet, which occurred during Experiment L2-3, is predicted to occur in a four-loop commercial pressurized water reactor of the Westinghouse Zion 1 type.

NRC FIN No. A6048 - LOFT Experimental Program ii

SUKk!AR Y o

Results from a study comparing and evaluating the differences among data from calculations of hypothetical, cold leg large break (200% double-ended break) loss-of-coolant accidents (LOCAs) in the Zion 1 nuclear power plant and from Loss-of-Coolant Experiment (LOCE) L2-3 are presented.

The Zion 1 large break LOCA calculations were performed using the RELAP4/M006 and RELAP4/M007 computer codes.

LOCE L2-3 was conducted in the Loss-of-Fluid Test (LOFT) facility. Results from LOFT LOCE L2-3 shor:d a reactor core-wide rewet early during the transient, which was not shown by the Zion I large break LOCA calculation using RELAP4/M006 in a previously reported LOFT prototypicality study.

As a followup of the prototypicality study and in the context of the LOFT experiment pla;,ning, the following objectives relative to the early rewet phenomenon in the core during a large break LOCA were formulated for this study:

e.

1.

Evaluate the differences in the Zion 1 RELAP4/ MOD 6 LOFT prototypi-cality study and the Zion 1 RELAP4/ MOD 7 large break calcui tion results.

Compare these results with the LOFT LOCE L2-3 data.

2.

Evaluate the Zion 1 large break calculation results for a case with more conservative initial conditions (102.4% nominal power, a maximum linear heat generation rate of 50 kW/m, 93% nominal flow, and an evaluation type of fuel rod model).

The RELAP4/M007 calculations conclusively showed a very significant early rewet in the core for the Zion 1 nuclear power plant at the hcttest spot on the average fuel rod and on the hot rod, even in the more conserva-tive calculations. The different behavior in the RELAP4/ MOD 6 large break calculation (no early quenching at the hottest spot at the hot rod) was partly due to differences in the hot rod modeling technique, but mainly due to the differences (improvements in RELAP4/ MOD,) in code versions with emphasis on differences in heat transfer correlations. The hot rod calcu-lation should be performed as an integral part of the system calculation iii t

i

to take in account the total system effects, as done in the RELAP4/M007 calculations, instead of performing a separate hot rod calculation using the btindary conditions from the system calculations.

The RELAP4/M007 100% nominal power case, on the basis of the same initial temperature difference across the core as observed for LOFT LOCE L2-3, showed similar peak cladding temperature and quenching behavior as occurred in LOFT LOCE L2-3.

This study showed that early rewet occurred with nominal and conserva-tive core hydraulic conditions and fuel peak linear heat generation rates that bracket current commercial plant operating conditions. Tne early rewet, which occurred during Experiment L2-3, is predicted to occur in a four-loop commercial pressurized water reactor of the Westinghouse Zion i type.

e e

n iv

CONTENTS e

ABSTRACT.............................................................

ii

SUMMARY

iii 1.

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

1 2.

COMPUTER CODE AND REFERENCE PLANT DESCRIPTIONS..................

4 2.1 RELAP4/M007 Description...................................

4 2.2 Reference Plant Description...............................

5 3.

RELAP4 INPUT M096L..............................................

6 3.1 Nodalization Scheme.......................................

6 3.2 Analytical Models.........................................

10 3.2.1 Discharge Flow Model..............................

10 3.2.2 Stagnation Pressure Option........................

11 3.2.3 Bubble Rise Model.................................

11 3.2.4 He at Trar.s f e r Mode l...............................

11 3.2.5 Nonequilibrium ECC Injection Model................

11 3.2.6 Vertical Slip Model...............................

12 3.2.1 Accumulator Model.................................

12 3.2.8 Gap Conductance Model.............................

12 3.2.9 Metal-Water Reaction Option.......................

13 3.3 RELAP4/ MOD 7 Input.........................................

13 4.

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

18 4.1 General Discussion of Large Break Calculations............

19 4.2 Calculations f'r More Conservative Conditions.............

37 5.

CONCLUSIONS.....................................................

42 l

6.

REFERENCES......................................................

44 APPENDIX A--RELAP4/M007 UPDATE LISTING...............................

45 O

APPENDIX B--ZION 1 POWER DISTRIBUTION DATA...........................

49

~

FIGURES 1.

RELAP4/ MOD 7.nodalization system diagram used for Zion 1 calculations....................................................

7 v

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4

-s-.-.

2.

RELAP4/ MOD 7 heat slab diagram used for Zion 1 calculations......

8 i

3.

Break flow at vessel side from R4-M6-110 and R4-M7-100 and -110 calculations...........................................

20 4.

Break flow at vessel side from R4-M7-100, -102.4, and -G-102.4 calculations.......................................

20 5.

Break flow at pump side with no broken loop ECC injection from R4-M6-110 and R4-M7-100' calculations.......................

21 6.

Break flow at pump side with ECC injection from R4-M7-100, -102.4, and -G-lC2.4 calculations....................

21 7.

Average pressure in lower plenum from R4-M6-110 and R4-M7-100 and -110 calculations.............................

22 8.

Average pressure in lower plenum from R4-M7-100,

-102.4, and -G-102.4 calculations...............................

22 9.

Average temperature in lower plenum from R4-M6-110 and R4-M7-100 and -110 calculations.................................

24

10. Average temperature in lower plenum from R4-M7-100, -102.4, and -G-102.4 calculations.......................................

24 11.

Inlet flow of average channel from R4-M6-110 and R4-M7-100 and -110 calculations...........................................

25 12.

Inlet flow of average channel from R4-M7-100, -102.4, and -G-102.4 calculations.......................................

25 13.

Inlet flow of hot channel from R4-M6-110 and R4-M7-iOO and -110 calculations...........................................

26 14.

Inlet flow of hot channel from R4-M7-100, -102.4, and -G-102.4 calculations.......................................

26

15. Outlet flow of average channel from R4-M6-110 and R4-M7-100 and -110 calculations...........................................

27

16. Outlet flow of average channel from R4-M7-100, -102.4, and -G-102.4 calculations.......................................

27 f

17. Outlet flow of hot channel from R4-M6-110 and R4-M7-100 and -110 calculations...........................................

28

18. Outlet flow of hot channel from R4-M7-100, -102.4, and -G-102.4 calculations.......................................

28

19. Flow in middle of hot channel from R4-M6-110 and R4-M7-100 l

and -110 calculations...........................................

29

20. Flow in middle of hot channel from R4-M7-100, -102.4, and -G-102.4 calculations................................'.......

29 vi

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___..._._.-_-,_-__.._..,.__--.r.__-..

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21. Surface temperature at-hot spot on average pin from R4-M6-110 and R4-M7-100 and -110 calculations...................

30

22. Surface temperature at hot spot on average pin from R4-M7-100, -102.4, and -G-102.4 calculations...................

30 a

23. Surface temperature at hot spot on average pin in hot channel from R4-M7-100, -102.4, and -G-102.4 calculations...............

31

24. Surface temperature at hot spot on hot pin from R4-M6-100 and R4-M7-100 and -110 calculations...................

31

25. Surface temperature at hot spot on hot pin from R4-M7-100, -102.4, and -G-102.4 c alcul at ions....................

32 26.

Core inlet and cold leg mass flow rates from R4-M6-110 calculation....................................................

32

27. Core inlet and cold leg mass flow rates from R4-M7-100 calculation....................................................

33

28. Core inlet and cold leg mass flow rates from R4-M7-110 calculation....................................................

33

29. Core inlet and cold leg mass flow rates from R4-M7-102.4 calculation....................................................

34

30. Core inlet and cold leg mass flow rates from R4-M7-G-102.4 calculation....................................................

34

31. Average quality in middle of hot channel from R4-M6-110 and R4-M7-100 and -110 calculations.................................

36 32.

Heat transfer mode at hot spot on hot pin from R4-M6-110

[

and R4-M7-100 and -110 calculations.............................

36 33.

Surface temperature at hot spot on hot pin from R4-M6-110 and R4-M7-110 calculations and from LOFT LOCE L2-3....

37

34. Store _ ener,v at hot spot on hot pin from R4-M7-100, -102.4, and -G-102.4 :alculations.......................................

38

35. Heat transfer mode at hot spot on hot pin from R4-M7-100,

-102.4, and -G-102.4 calculations...............................

38

36. Average quality in middle of hot cnannel from R4-M7-100,

-102.4, and -G-102.4 calculations...............................

39

37. Pump speed in intact loop from R4-M6-100 and R4-M7-100 calculations....................................................

40

38. Pump speed in intact loop from R4-M7-100, -102.4, and

-G-102.4 calculations...........................................

40 vii

l

39. Pump speed in broken loop from R4-M6-110 and R4-M7-100 calculations....................................................

41

40. Pump speed in broken loop from R4-M7-100, -102.4, and

-G-102.4 calculations...........................................

41 TABLES 1.

NOMINAL OPERATING CONDITIONS FOR ZION 1.........................

5 2.

DESCRIPTION-0F VOLUMES FOR THE RELAP4/ MOD 7 SYSTEM INPUT MODEL...

9

'3.

FUEL ROD GAP THERMAL CONDUCTIVITY PROPERTIES....................

13 4.

INITIAL PLANT CONDITIONS FOR ZION 1 CALCULATIONS................

14 5.

LINEAR HEAT GENERATION RATE AT DIFFERENT AXIAL AND RADIAL LOCATIONS IN THE, ZION 1 CORE....................................

16 6.

COMPARIS0N OF LOFT LOCE L2-3 AND CALCULATION PARAMETERS.........

17 7.

CALCULATED SEQUENCE OF EVENTS FOR ZION 1 CALCULATIONS...........

18 B-1. POWER DISTRIBUTION VERSUS AXI AL POSITION........................

51 l

l l.

i I,

i 1

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viii l

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(-

d REuAP4/M007 COMMERCIAL'PWR LARGE BREAK TRANSIENT ANALYSIS COMPARED WITH LOFT DATA 1.

INTRODUCTION 4

This report presents results of a study comparing and evaluating the differences among data from loss-of-coolant accident (LOCA) calculations

- using the RELAP4/M006 and RELAP4/ MOD 7 computer codes and from Loss-of-Coolant' Experiment (LOCE) L2-3 performed in the Loss-of-Fluid Test (LOFT) i facility.I LOCE L2-3 simulated a large break (double-ended offset shear) in the cold leg of a large pressurized water reactor (PWR) coolant loop.

The RELAP4/M006 calculation used in this comparison study was from a previous LOFT prototypicality study,2 where results from LOFT LOCE L2-3 were com[ared with calculated results from a similar hypothetical large break LOCA in the Zion 1 nuclear power plant. The results showed good, i

- general agreement between the experiment and the calculation; however, an early core-wide rewet occurred in LOFT during'the experiment which was not predicted by RELAP4/ MOD 6. in the Zion 1 LOCA calculation. The RELAP4 computer code has since been corrected and updated to RELAP4/ MOD 7.

RELAP4/

M007 was used for the additional ~ Zion 1 LOCA calculations evaluated in this

_ comparison study.

.The objectives for this study are as follows:

1.

To evaluate the differences in large break calculation results as related to the early rewet occurrence between (a) the hot fuel rod 2

calculation performed with RELAP4/ MOD 6 using the lower and upper plenum boundary conditions from the system calculation with an average channel modeled, and (b) an integral system calculation performed with RELAP4/ MOD 7 with an average channel (average rod) and a_ hot channel (hot rod-and hottest rod) modeled in one run.

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2.

To compare the RELAP4/M007 calculation results with the LOFT LOCE 3

L2-3 data and to evaluate the advantages of RELAP4/ MOD 7 over M006.

3.

To evaluate the influence of a higher maximum linear heat genera-tion rate (MLHGR = 50 kW/m), a lower core flow (93% nominal flow),

and a more conservative gap conductance model, on the early rewet occurrence during a large break LOCA.

The RELAP4/M007 computer code predicts thermal-hydraulic response of a large PWR to hypothetical LOCAs and has some capabilities for the calcula-tion of other transients. As such, it incorporates essentially all the analytical models required to correctly understand the events typical of LOCA plant conditions. A brief synopsis of the results of the analysis follows.

For this study, four Zion 1 LOCA calculations were performed with RELAP4/ MOD 7 for 100, 110, and 102.4% nominal power cases (for the later case, calculations with and without a conservative gap conductance model were performed, both with a higher MLHGR and a lower initial flow). The results of all four calculations showed that early rewet occurred at the hottest spot at the average rod, at the average hot rod in the hot channel, and at the hot rod.

The hot rod calculation performed with RELAP4/M006 for the LOFT proto-typicality study did not show the rewet phenomenon, which could be attributed to differences in hot rod modeling, but probably resulted from differences in RELAP4 versions.

In the MOD 7 version, some minor faults were corrected and some changes and improvements, compared with M006, were made in the following areas:

slip, critical heat flux, nonequilibrium, and stagnation pressure calculation.

Tne RELAP4/M007 calculations showed that the early rewet in the core occurred at s5.5 s in the large break transient at the hottest spot in the large PWR, as was shown in the LOFT LOCE L2-3 experimental data. And even 2

when the more conservative large PWR initial conditions were established, the rewet still occurred but later in time and depended on the initial stored heat.

(If the initial stored heat is higher, the maximum peak clad-ding temperature is higher and the quench takes place later.) The more conservative Zion 1 calculations indicated that early rewet will occur if a higher MLHGR (50 kW/m) is applied to a LOFT large break experiment.

l The computer codes and reference plant (Zion 1 nuclear power plant) used for the evaluation are described briefly in Section 2.

Section 3 describes the RELAP4 input model used in the calculations. Results from the calculations are compared in Section 4.

Conclusions and references are presented in Sections 5 and 6, respectively. Appendix A contains the update listing for RELAP4/ MOD 7.

Appendix B lists power distribution data for the most extreme initial conditions used in these Zion 1 large break LOCA calculations.

O e

3 1

2.

COMPUTER CODE AND REFERENCE PLANT DESCRIPTIONS The data for this comparison study were from computer simulations of hypothetical large break LOCAs in the Zion 1 nuclear power plant using the 2

RELAP4/ MOD 6 (LOFT prototypicality study ) and the RELAP4/ MOD 7 computer codes and from LOFT LOCE L2-3.3 The RELAP4/ MOD 7 computer code and the reference PWR (Zion 1) are described briefly in the following subsections.

l I

2.1 RELAP4/ MOD 7 Description 2

The RELAP4/ MOD 6 code used in the LOFT prototypicality study did not l

calculate the early core-wide rewet observed during the LOFT large break LOCEs.

The code was rerised and updated to RELAP4/ MOD 7 (see Appendix A for additional updates).

The RELAP4/M007 code,4 used for the Zion 1 large l

a break LOCA calculations in this study contained the following additional i

characteristics related to the blowdown phase of a LOCA:

1.

Self initialization capability 2.

Flow-regime-dependent vertical slip model 3.

Nonequilibrium emergency core coolant (ECC) injection model 4.

Stagnation properties option I

5.

The " MOD 7 CHF correlation" 6..

The "FRAP-T best estimate fuel model"; however, the "best estimate fuel rod model with a modified Ross and Stoute model" was not included.

5 The Zion 1 input deck used to check out RELAP4/M007 was changed and adapted for the blowdown phase. The renodalization part of the input deck,

,pecifically set up for the reflood phase, was deleted.

a.

RELAP4/M007 (Update 100), Idaho National Engineering Laboratory Configuration Control Number H01343JB.

s 4

2.2.

Reference Plant Description LThe Zion 1 nuclear power plant, a four-loop PWR designed by Westinghouse Electric Corporation, was chosen as the' reference plant for this study for the following reasons:

2 1.

The plant was used in the LOFT prototypicality study 2.

The Zion 1 plant is considered to be representative of the Westinghouse four-loop PWRs of that size; therefore, the calculations shown can be_ considered as typical for a large PWR of that type 3.

A basic RELAP4/M007 model of the Zion 1 plant,5 based on detailed plant information from the BE/EM study,6 was already available.

The nominal operating conditions for the plant are-given in Table 1.

- The reactor core consists of 193 fuel bundles (each with a 15 x 15 rod arrangement), with 204 fuel rods per bundle.

' TABLE 1.

NOMINAL ~0PERATING CONDITIONS FOR ZION 1 Operation Parameter Value Corepower[MW(t)]

3238.0 Peak power density (kW/m) 35.8 Corecoolantflow(kg/s) 18 296.0 Core coolant inlet temperature (K) 550.0 System pressure (MPa) 15.5 s

O l

l 5

3.

RELAP4 INPUT MODEL In this section, the nodalization scheme, analytical models, and input for the RELAP4/M007 code used in the Zion 1 large break LOCA calculations are described.- Also, differences between the RELAP4/M006 and RELAP4/M007 s

models are discussed.

The RELAP4 calculations are identified in this report as R4-M6-110,2 R4-M7-100, R4-M7-110, R4-M7-102.4, and R4-M7-G-102.4, where R4_is "RELAP4";

M6 and M7 are Modification 6 and Modification 7, respectively; and 100, 102.4, and 110 are percent of nominal power. G is the evaluation Ross and Stoute gap conductance model with rod swelling and flow blockage taken into account.

(In the other runs, best estimate models are used.)

3.1 Nodalization Scheme The R4-M7 Zion 1 nodalization scheme shown in Figure 1 consists of 57 volumes and 78 junctions. The model contains 38 heat slabs as shown in Figure 2.

A brief description of each volume is given in Table 2.

In Figure 2, heat slabs 1 through 6 represent an average rod in an average channel, heat slabs 7 through 12 represent an average hot rod in a hot channel, and heat slabs 13 through 18 represent.a hottest rod in a hot channel. Some specific nodalization characteristics are as follows:

1.

The-intact loop represents three loops of a four-loop PWR, and the broken loop represents one loop.

2.

In the broken loop, the 200% break was simulated by fully opening two valves (Junctions 25 and 26 in Figure 1), each with a full pipe flow area, and closing the valve (Junction 23) between these valves. The coolant was discharged from the vessel side (Junc-tion 25) and from the pump side (Junction 26) through the break to the containment.

3.

The high-pressure injection system (HPIS), accumulator, low-pressure injection system (LPIS), and charging system supplied 6

09 M

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Pressurizer 23 3

16 17 3[

15 5

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16 6

6 3

64, 56 head 14 29 91 2{ 1 63g 15 18 6

7 31 155 l

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13 17 7

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I 33l- ' 57 404 R7 21 20 3

139 h45 l T_

39 4 446 31 (22 k20 52 w

47 l38 H44 1 23 7

8 38 4 445 l37 h43 l Containment Pump Accumulator 541 374 444

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604 49 151 1

2 Break 9[h10) 10 23 594 56 33 Charging Pump 2j V

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26 25 Accumulator intact Loop Broken Loop INEL-A-17 793 Figure 1.

RELAP4/ MOD 7 nodalization system diagram used for Zion 1 calculations.

f Intact Loop Broken Loop msmaum n'

-Secondary Secondary

- www V27 V28 26 24 20 22 V5 V4 V16' V17

[

Steam i

generator 25 V6 V3 23 19 V15 V18l 21 O

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U 7

L37.J l

=lV56 Upper head 7

1 27 V46 /c d legs V55 UP-3 136l i

U To l

. DC-1 hot =

V54 UP-2 I 35 l leg g

V53 UP1 l 34 l U

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4 28 V47 o

6 V39 12 V45 18 sj 5

V38 2

11 e V44 17

> Vessel

' DC-2 yj 4

V37 10 V41 16 p

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9 V42 15 29 V48 e

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g 30 V49 V52 LP-3 l 33 l 4

DC-4 VS1 LP-2 l 52 l 4

=

V50 LP-1 l 31 l INEL-A 17 792 j

Figure 2.

RELAP4/M007 heat slab diagram used for Zion I calculations.

8

TABLE 2.

DESCRIPTION OF VOLUMES FOR THE RELAP4/ MOD 7 SYSTEM INPUT MODEL Control Volew -

Description

+

1, 8, 9, 11, 12 Intact loop piping 2

Inlet plenum of intact loop stec.m generator 3,4,5,6 Intact loop steam generator tubes 7

Outlet plenum of intact loop steam generator 10

~ Intact loop primary coolant pump 13, 20, 21, 23-26 Broken loop piping 14 Inlet plenum of broken loop steam generator 15, 16, 17, 18 Broken loop steam generator tubes 19 Outlet plenum of broken loop steam generator 22 Broken loop primary coolant pump 46 Downconer inlet annulus 47-49 Downcomer 50-52

_ Lower plenum 34-39 Core--average channel 40-45 Core--hot channel 4

53-55 Upper plenum 56 Upper head 57 Core bypass (barrei-baffle) 27 Secondary side of intact loop steam generator 28 Secondary side of broken loop steam generator 4

30 Pressurizer 29 Pressurizer surge line 32 Intact loop accumulator 33 Broken loop accumulator 31 Containment water to the reactor system during the simulated accident, at

-given set-points.

(In the R4-M6 calculation, these systems were closed off in the broken loop.)

I 4..

The following main differences in nodalization are noted l

between the R4-M7 model and the R4-M6 model used in the LOFT prototypicality study:

a.

The split downcomer model was not used in the R4-M7 model.

Theuseoftheimprovebverticalslip~modelintherelevant 4

l vertical junctions resulted in a similar behavior of the 9

i t-

l downcomer compared with the downcomer behavior in the R4-M6 calculation with the split downcomer..

)

b.

The average channel.in the_ core was divided into six volumes 1

in R4-M7 instead of three volumes as'in R4-M6.

i i

c.

The hottest rod was modeled within the R4-M7 system run in a separate hot channel (divided in six volumes) and was divided into sir. heat slabs. The hottest rod calculation in i

R4-M6 was performed with a hot rod model with 11 heat slabs in a hot channel (divided into 12 volumes), using the boundary conditions from the system run.

d.

The behavior of an average hot rod in tim,M t channel was y

calculated in the R4-M7 system run, but was not calculated j

by R4-M6.

k 3.2 Analytical Models The analytical models used in the R4-M6 and R4-M7 calculations of Zion 1 large break LOCAs are identified'in the following subsections.

3.2.1 Discharge Flow Model The Henry-Fauske critical flow model was used for the subcooled region,'

with 0.9 as the flow multiplier in R4-M7 (in R4-M6, 1.0 was used as the flow

. multiplier), in order to match the-R4-M6 break flow which will be explained in more detail in Section 4.1.

The homogeneous equilibrium. critical flow model was used for the satur-

.ated region, with l.0 as the flow multiplier in R4-M7 (in R4-M6, 1.0 was used also).- The transition region ended with a quality of 0.02 (in R4-M6,

~the quality was 0.0025).

(

f t

10 l

3 I

3.2.0 Stagnation Pressure Option This option was used during the blowdown phase of the transient. The-static pressure in the upstream volume of the break was converted to the stagnation pressure using the critical flow tables.

This option was deacti-vated if the pressure in the vessel side discharge volume was less than

-0.5 MPa, due to possible not realistic pressure oscillations.

In the R4-M6 calculations, the cold. leg break flow stagnation pressure i

was simulated through increasing the broken leg flow area by 400%.

~

3.2.3-Bubble Rise Model All the volumes contained homogeneous fluid, except the pressurizer, the accumulators, the contsinment, and the secondary side of the steam f

generators. A tiable rise model was applied to these excepted volumes.

4 1

3.2.4 Heat Transfer Model

^

4 The R4-M6 blowdown heat transfer correlations were used with an implicit wall temperature solution in R4-M7. -Also, for the transition boiling regime, the modified Tong-Young correlation; for the film boiling regime, the Condie-Bengston III correlation; and for the critical heat flux (CHF), the M007 CHF correlation, were used in R4-M7 (in R4-M6 calculations, the W3/Hsu and Beckner CHF correlations were used). At the secondary side of the steam generator, the natural convection option was used.

7 3.2.5 Nonequilibrium ECC Injection Mode 1 The nonequilibrium ECC mixing model was implemented in all of those volumes where-ECC was assumed likely to penetrate. These include the intact loop cold leg to and including the pump, intact loop hot leg, broken loop e

l

. cold leg from vessel to steam generator, inlet annulus, downcomer, lower 11

{

plenum, upper plenum, upper head, and barrel baffle.

This model was acti-vated when the accumulator flow into the intact loop cold leg reached 45.4 kg/s.

0 3.2.6 Vertical Slip Mode 1 The flow-regime-dependent vertical slip model was used in most vertical junctions.

It was also used in the junctions between the intact loop cold leg and inlet annulus (Junction 31) to facilitate nonequilibrium ECC injec-tion.

Individual junction dimensions were specified in all cases.

3.2.7 Accumulator Model The nitrogen gas in the accumulator was modeled as an air head with a polytropic expansion model (P v" = c, n = 1.4).

3.2.8 Gap Conductance Model The best estimate model used to model the fuel rod did take into account any rod swelling or coolant flow blockage. The gap conductance properties were chosen such that the initial fuel hottest centerline tem-perature matched the comparable temperature in the basic BE/EM study.6 The fuel rod gap thermal conductivity properties are listed in Table 3 for the R4-M6 and R4-M7 calculations.

T.,e ct.aiistic gap conductance properties were lower (like those of heiium) and were used in the R4-M7-G-102.4 calculation.

In this calcula-tion, the simplified rod swelling and flow blockage fuel rod model was used which automatically implied a modified Ross and Stoute gap conductance model was included.

In the R4-M6 calculation,2 the best estimate fuel rod model with the MacDonald-Broughton gap conductance model was used, but was deleted in R4-M7.

(The gap conductance properties were higher than realistic.)

12

TABLE 3.

FUEL R00 GAP THERMAL CONDUCTIVITY PROPERilES Thermal Conductivity Properties (with interpolation and extrapolation)

Temperature Conductivity Calculation Fuel Rod Model (K)

(W/m k)

R4-M7-100 Best estimate 605.35 1.817 R4-M7-110 No swelling 641.45 2.077 R4-M7-102.4 No blockage 677.59 2.466 and in Reference 5 R4-M7-G-102.4 Simplified rod swelling 273.15 0.142 and flow blockage model 588.70 0.239 and Ross and Stoute gap 810.92 0.294 conductance model 1088.70 0.358 1366.48 0.417 3255.37 0.746 R4-M6-110 Best estimate with 273.15.

0.719 (prototypigality MacDonald-Broughton gap 3255.37 0.719 study) conductance model

- 3. 2.! Metal-water Reaction Option The Cathcart-Pawell Zr0 reaction model, a best estimate model, was 2

used in R4-M7.

3.3 RELAP4/ MOD 7 Input The initial plant conditions for the different Zion I calculations are shown in Table 4.

i 5

The input deck for R4-M7-100 was derived from the input deck used to check out R4-M7 for Zion 1, which was based on the BE/EM study.6 Furthermore, some data have been chacked with the data mentioned in the draft Zion input notebook for the RETRAN code obtained from Commonwealth Edison Company.9 13

TABLE 4.

INITIAL PLANT CONDITIONS FOR ZION 1 CALCULATIONS Initial Conditions Parameter R4-M6-110 R4-M7-110 R4-M7-100 R4-M7-102.4a Power [MW(t)]

3 540.0 3 540.0 3 238.0 3 315.0 Core mass flow (kg/s) 18 395.0 18 395.0 18 395.0 17 061.5 Pressurizer pressure (MPa) 15.43 15.45 15.45 15.45 Core inlet temperature (K) 549.8 549.8 549.8 549.8 AT acr.;ss core (K) 35.8 35.8 32.44 36.09 Accumulator pressure (MPa) 4.21 4.21 4.21 4.21 Average linear heat genera-24.57 24.57 22.48 23.0 tion rate (kW/m)b MLHGR for average rod (kW/m) 28.79 28.1 25.7 32.43 35.17 32.17 42.41 MLHGR for average rod in hot channel (kW/m)

MLHGR for hot rod (kW/m) 39.40 38.29 35.02 49.93 Total peaking factor <

l.60 1.63 1.63 2.32 Axial peaking factor 1.17 1.23 1.23 1.50 Radial peaking factor:d average rod 1.0 0.97 0.97 0.97 Average rod in hot channel 1.22 1.22 1.31 Hot rod 1.37 1.32 1.32 1.54 istribution, peaking factors, and core flow derived from Power,powerp0(seeAppendixB).

a.

Westinghcuse data b.

Total generated power per total length of fuel rods.

Core length = 3.66 m; number of fuel rods = 39 372 (193 x 204).

c.

With no uncertainties accounted for.

d.

Radial peaking factor = fuel.od power per average fuel rod power.

Average fuel rod power = total power per total number fuel rods (39 372).

14

The linear heat generation rates (LHGRs) at different locations in the Zion 1 core are showr, in TaLle 5.

The LHGR for the 102.4% power cases was derived from the Westinghouse information 0 (see Appendix B).

The applied LHGR at the hot rod was 50 kW/m (15.2 kW/f t) instead of 52.5 kW/m (16 kW/f t) as stated in Appendix B.

This change in LHGR was mainly caused by the por-tion of the power assumed to be generated in the moderator oy gamma heating (best-estimate approach). The other LHGRs were lower than stated in Appendix B for the same reasor.

The parameters for LOFT LOCE L2-3 and the calculations are compared in Table 6.

e M

15

TABLE 5.

LINEAR HEAT GENERATION RATE AT DIFFERENT AXIAL AND RADIAL LOCATI0h5 IN THE ZION 1 CORE Calculation Calculation R4-M6-110a R4-M7-110a R4-M7-100a R4-M7-102.4b LHGR LHGR LHGR LHGR Heat Slab (kW/m)

Heat Slab (kW/m)

(kW/m)

Average Pin in Average Channel (Bottom) 7 23.98 1

19.13 17.5 12.44 2

27.6 25.25 25.71 8

28.79 3

38.1 25.7 32.43 4

27.9 25.5 30.10 9

19.76 5

23.5 21.5 18.23 6

10.5 9.6 8.5 (Top)

Average Pin in Hot Channel (dottom) 7 23.1 21.9 16.27 8

34.5 31.6 33.63 9

35.17 32.17 42.41 10 34.9 31.9 41.78 11 29.44 26.9 25.73 12 13.1 12.0 11.0 (Top)

Hot Pin in Hot Channel (Bottom) i 1

24.35 13 26.0 23.8 19.25 2

29.76 3

35.76 14 37.6 34.39 40.0 4

38.97 5

39.29 15 38.29 35.02 49.93 6

39.40 7

39.26 16 38.0 34.75 46.53 l

8 37.64 l

9 34.59 17 32.05 29.3 30.67 l

10 28.48 11 19.10 18 14.3 13.08 13.07 (Top) l a.

A middle of life power profile.

b.

Power profile derived from Westinghouse data 10 (see Appendix B).

16 l

TABLE 6.

COMPARIS0N OF LOFT LOCE L2-3 AND CALCULATION PARAMETERS i

Calculation l

Parameter L0cE L2-3 R4-M6-110 R4-M7-110 R4-M7-100 3

System volume (m )

7.22 355.7 355.7 355.7 Power [MW(t)]

36.0 3 540.0 3 540.0 3 238.0 Initial core mass flow (kg/s) 199.8 18 395.0 18 s 0

18 395.0 Initialcgremassflux 1 ?15.1 3 867.3 3 867.3 3 867.3 (kg/s m )

AT across core (K) 32.2 35.8 35.8 32.4 Core length (m) 1.68 3.66 3.66 3.66 Number of fuel rods 1 300.0 39 372.0 39 372.0 39 372.0 Average linear heat generation 18.0 24.57 24.57 22.48 rate (kW/m)a MLHGR for hot rod (kW/m) 39.0b 39,40 38.29 35.02 MLHGR for average rod (kW/m) 23.0 28.79 28.1 25.7 Total peaking factor 2.4b 1.60 1.63 1.63 Axial peaking factor 1.4 1.17 1.23 1.23 a.

Total generated power per total length of fuel rods.

b.

Center fuel module.

i 4

i 9

I

?

17

4.

xtdutid The calculated sequence of events for the different Zion I calculations are shown in Table 7.

The results discussed in this section are mainly divided into two comparison studies, with R4-M7-100 as a base case:

l.

Comparison between the R4-M6 results from the LOFT prototypicality study and the R4-M7 calculation results 2.

Comparison of the different R4-M7 calculation results.

Furthermore, the surface temperatures at the hot spot on the hot rod from calculations R4-M6-110 and R4-M7-100 and LOFT LOCE L2-3 data are compared.

TABLE 7.

CALCULATED SEQUENCE OF EVENTS FOR ZION 1 CALCULATIONS Time After Rupture (s) a Event R4-M6-110 R4-M7 Blowdown initiated 0.0 0.01 Reactor scrammed 0.53 0.53 HPIS plus charging flow started 1.4 1.4 Pressurizer 2pty 9.3 9.1 Accumulator flow started 15.0 13.5 LPIS flow started 23.0 22.1 Main coolant pumps remain running during the whole transient Sequence of events is the same for the different R4-M7 runs.

a.

O 18

4.1 General Discussion of Large Break Calculations O

An important region in the large break transient is the first dryout, followed by quenching of the core.

Core quenching was controlled by the hydraulic behavior in the system which is discussed in the following paragraphs.

Calculations of break flow at the vessel side, shown in Figures 3 and 4 for the different runs, followed each other closely, except in the R4-M7-110 case.

In this case, the same discharge flow multiplier (1.0) was used as in the R4-M6-110 case, but it resulted in a hiober subcoole' discharge flow (until s2 s). This was possibly due to the inadequate stagnation pressure calculation in the discharge volume in R4-M6-ll0 which significantly influenced the subcooled discharge flow.

In the other R4-M7 calculations, a subcooled discharge flow multiplier of 0.9 was used in order to match the R4-M6 break flow.

~

Calculations of break flow at the pump side, shown in Figure 5, followed each other closely for the different runs without ECC injection in the broken loop.

The discharge flow at s0.05 s in the transient was at saturated conditions with the same flow multiplier of 0.9 for the different calculations. The difference in stagnation pressure calculation had very little influence in the saturated region. The break flow is shown in Figure 6 for the R4-M7 calculations with ECC ir.jection in the broken and intact loops. The ECC injection in the broken loop did not affect the core behavior, but did influence directly the quality and the magnitude of the break flow.

The pressure in the lower plenum, shown in Figures 7 and 8, decreased very fast until the discharge flow at the pump side changed from subcooled to two-phase flow at 0.05 s.

At s2 s in the transient, the pressure depressurization rate decreased again because the discharge flow regime at the vessel side changed from subcooled to two-phase flow. At $13 s, the depressurization rate increased slightly due to cold water injection from the accumulator to the intact loop.

19

0.5e I

I I

I n4-me-iso 0.25

[_

2 n4-w7-ioo y

e R4-M7-Ilo 0.pc 3-v

i

_l S -0.25 iw X

2 2

3 -0.50 f1 h-

~

h-0.75 2

1

. -1.00 E -1.25 3-1.50

~~

=

-1.75

-2.00

-2.25

's

-2.50 I

I I

I

-]

-2. s O

2 4

6 8

10 12 14 16 10 20 22 24 26 7tae ist Figure 3.

Break flow at vessel side from R4-M6-110 and R4-M7-100 and -110 calculations.

B.se I

I I

a4-w7-ioo 0.25 2 n4-w7-io2.4 8 R4-M7-G-lo2.4

~~

B.00 3--

' s -0.25 1

X 1

j g -0.50

[ 12 1-

}-e.75 1

. -1. m 13

u. -1. 25

~

-1.50 x

-1.75

-2.00

-2.25 p1

-2.50 G

-I I

I I

-2. n 0

2 4

6 0

10 12 14 16 18 20 22. 24 26 Tlae (el Figure 4.

Break flow at vessel side from R4-M7-100, -102.4, and -G-102.a calculations.

20

1.s l

l l

1 a4-us-llo 2 R4-M7-100 s,

R 1.s 21 i

5 m

i e.5 241

~

m.

1

=

E h

$%i s:=1%'

e.e a-2 21

~~

-e.5 e

s to is as as Time (el Figure 5.

Break flow at pump side with no broken loop ECC injection from R4-M6-110 and R4-M7-100 calculations.

26ese I

I I

l 1 R4-M7-100

~

2 R4-M7-lo2.4

_3 R4-M7-G-102.4 14EMe r;2ees

^.

) lease 12

=

2 asse u.

L j

sees 1

4eee 2eso q

N 1-1 I

I l

l l

11 e

c-

~

e 2

4 6

8 is 12 14 16 10 2e 22 24 26 Time tal Figure 6.

Break flow at pump side with ECC injer' ion from R4-M7-100,

-102.4, and -G-102.4 calculations.

21

16 3 I

I I

l l

l R4-Me-Il0 2 n4-m7-ioa S R4-M7-Il0 14 12 h_

~

10 h

8

~

C 13

~el

_g g

1N 4

ig N2 A 2N l

i l

i f

l l

l i

l l

O 0

2 4

6 8

10 12 14 16 18 20 22 24 26 Tire ist

._ Figure 7.

Average pressure in lower plenum from R4-Me-110 and R4-M7-100 and -110 calculations.

16 3 I

I I

i i

1 R4-M7-IOO

~

2 R4-M7-102.4 8 R4-M7-C-102.4 14 12

~

mi 10 g

8 E

1 E

6 1

4 2

1 1%

I I

O 0

2 4

6 8

10 12 14 16 18 20 22 24 26 Tlae is)

Figure 8.

Average pressure in lower plenum from R4-M7-100, -102.4, and

-G-102.4 calculations.

22

i The average temperature in the lower plenum shown in Figures 9 and 10 increased until $1 s in the transient because of the heated reversed flow from the core to the lower plenum (see Figures 11 and 12). After 1 s, the I

lower plenum temperature followed the saturated temperature.

The inlet flow of the average channel shown in Figures 11 and 12 and of the hot channel shown in Figures 13 ano 14 was negative until s2 s for l

all the calculations. However, the outlet flows of the average channel i

shown in Figures 15 and 16 and of the hot channel shown in Figures 17 and 18 were positive.

l Until about 3.3 s, a nearly zero flow condition occurred in the middle of the hot channel as shown in both Figures 19 and 20 as well as in the i

middle of the average channel which caused the core to heat up.

In the different R4-M7 calculations for Zion 1, a very effective early quench occurred af ter the first dryout.

" Quenching" is defined to be when the fuel rod surface temperature decreases nearly instantly to saturated temperature due to the rewet process.

f This early quench phenomenon is shown by the surface temperatures at the hottest spot (a) of the average rod in Figures 21 and 22, (b) of the average rod in the hot channel in Figure 23, and (c) of the hottest rod in i

Figures 24 and 25.

For the different R4-M7 calculations, the values of the peak cladding temperature and the quench time were different which are explained later in this report.

e A better illustration of the rewet phenomenon is shown in Figures 26 c

{

through 30, which show the "rewet area." The rewet area was determined by the difference between the intact and broken loop cold leg flow at the inlet I

annulus location which resulted over a short period of time in an upwards i

~ flow in the core and a partial or entire rewet of the core.

l-23 1

See i

l I

I I

I 1 R4-na-tio 1

2 R4-N7-1DO 2

8 R4-N7-110 560 2%

3 s

s 1

m 2

\\

p g

dee 1

460 I

I I

l 1

I I

I O

2 4

6 8

10 12 14 16 18 20 22 24 26 Time (al Figure 9.

Average temperature in lower plentn from R4-M6-110 and R4-M7-100 and -110 calculations.

See f

I I

I i R4-M7-too 3

4 8

M7 52.4 1

1 550 2F 1

h g

1 530

\\

l 1

l 520 t

s i

I 510 1

l 5m t

c 490 1N l

3 2%

l c

4a0 1

470 460 1

450 I

I I

I

[

0 2

4 6

8 10 12 14 16 18 20 22 24 26 1

Time tal Figure 10. Average temperature in lower plenum from R4-M7-100, -102.4, and

-G-102.4 calculations.

24 t

2.00 l

l l

l 1 a4-me-iso 1.75 3-2 N4-M7-LOO I

9 R4-M7-110 m

1.50 N

e 1.25 X

i.00

) e.75 0.50

.m E 0.25 1

1 la

~

M1 'P14k N-j I

0.00 2

-e.25

-e.50

-0.75 1

-1.00 I

I I

I

-1.25 9

2 4

6 8

10 12 14 16 18 20 22 24 26 Time (s)

Figure 11.

Inlet flow of average channel from R4-M6-110 and R4-M7-100 and

-110 calculations.

2.00 l

l l

l 1 R4-M7-300 1.75 i_

2 R4-M7-102.4 3

8 R4-M7-G-102.4 1.50

's 1.25 52 1.00

) 0.75

> 0.50 0.25 2

! 0.00 5

1- -

1

~

NN-

-e.25

-e.50

-0.75 3

-1.00

-1.25 l

I I

I y

0 2

4 6

8 10 12 14 16 18 20 22 24 26 Trae ist Figure 12.

Inlet flow of average channel from R4-M7-100, -102.4, and

-G.102.4 calculations.

25

~__

0.5 I

i I

I I

i R4-we-Ilo 3

2 n4-ur-loa 0.4 1-'

3 R4-M7-Il0 23

~

0.2

~

01 1

L i

1r 1 F1 4 @ ie-a.0

"[

}

-0.1 i

I

-0.2 e

1

-0.3 1-

~

~0.4 I

I I

I

_g,3 O

2 4

6 8

10 12 14 16 10 20 22 24 26 Tlae ist Figure 13.

Inlet flow of hot channel from R4-M6-110 and R4-M7-100 and -110 calculations.

0.5 I

I I

I R4-H7-loo 2 R4-M7-102.4 l

1,_,

O R4-M7-G-102.4 0.4 3 0.3

~

}

e.2

~

3 2

-[ g1h^

M I

j 1f1 1y1gm-e.0 I

-0.1 l

23 l

-0.2 1H 3

I I

i l

_,,3 0

2 4

6 8

10 12 14 16

.8 20 22 24 26 l

I Time fel Figure 14.

Inlet flow of hot channel from R4-M7-100, -102.4, and -G-102.4 calculations.

l 26

2.0 l

l l

l 1 R4-MS-Il0 1.8 2 R4-M7-300 9 R4-M7-Il0 1.6

g 1.4 1.2 1.0 m

0.8 8.6 O.4

~~

E 1

0.2 2

0.0 3

1 iga 2ap.t3 fAf1 Ara ^^*-

yg vg v v v r

r

-e.2 i

-0.4 1

I I

I 0

2 4

6 0

10 12 14 16 10 20 22 24 26 T!as tal Figure 15. Outlet flow of average channel from R4-M6-110 and R4-M7-100 and n

-110 calculations.

2.0 l

l l

l 1 84-M7-300 1.8 1 2 R4-M7-102.4 3

^8 R4-M7-G-102.4 1.6 h

1.4 X

3 1.2 7

1.0 m

0.8 l

0.6 3

2

=

0.4 1

0.2 3

1

~

0.O F31d 13,1%p i

tj7_

[1

-e.2 4

-0.4 2

_,,g 1

I I

l

.i j

0 2

4 6

8 10 12 14 16 10 20 22 24 26 T!ae (s) i Figure 16.

Outlet flow of average channel from R4-M7-100, -102.4, and 1

-G-102.4 calculations.

27

0.5 l

l l

l R4-MS-Il0 2 R4-M7-100 0.4 2

0 R4-M7-Il0 0.3 0.2 a

Iy 0.1 e

0.0 12 " 1 2k1 M 51 " -

-0.1

-0.2

-0.3

-0.4

-0.5 0

2 4

6 8

10 12 14 16 10 20 22 24 26 Tlee tal Figure 17. Outlet flow of hot channel fron R4-M6-110 and R4-M7-100 and

-110 calculations.

0.45 I I

I l

l l

1 R4-H7-100 1

2 R4-M7-102.4 0.40 2-0 R4-M7-G-102.4 0.3s 0.30 l

~

l 0 0.25 5 0.20

~

u.

0.15 1

0.10 2

=

2

[

0.00 W-1F 1

-0.0s l

I I

-e.10 0

2 4

6 8

10 12 14 16 18 20 22 24 26 Time is)

Figure 18. Outlet flow of hot channel from R4-M7-100, -102.4, and -G-102.4 calculations.

28

0.5 l

l l

l I R4-Me-l!0 3

2 R4-M7-100 0.4 1-0 R4-H7-110 0.3 0.2 8.1 3

d

'1 1*-

?-

2 0.0 1

12 1 P 1 M EC'1 M 1W-S 1

-0.1 F

=

1

~

-0.2

-0.3

-0.4

-0.5 0

2 4

6 8

10 12 14 16 18 20 22 24 26 7tae ist Figure 19.

Flow in middle of hot channel from R4-M6-110 and R4-M7-100 and

-110 calculations.

0.5 l

l l

1 R4-M7-300 2 R4-M7-102.4 1_

8 R4-N7-G-102.4 0.4 3 0.3 O

h 0.2 2

0.1 1

E 1

8.0 id 1- ~

i 1N-

-0.1

~

-0.2 I

-0.3 O

2 4

6 8

10 12 14 16 18 20 22 24 26 7tae (al Figure 20. Flow in middle of hot channel from R4-M7-100, -102.4, and

-G-102.4 calculations.

l 29

7s0 l

I '

I I

i 1 R4-me-tio 760 2 R4-M7-too 3g 8 R4-N7-110 740 r1 m

(

1

!m f

2 t

660 f

m 640 1

1--- -1'i

/

620 a

1 1

g, r-5e0 22 2

560 1

1 2%M'

\\

2e se i

l i

I 12 I I

l I

I g

3 O

2 4

6 8

10 12 14 16 10 20 22 24 26 Tlae ial Figure 21.

Surf ace temperature at hot spot on average pin from R4-M6-110 and R4-M7-100 and -110 calculations.

1000 I

l l

1 R4-N7-100 2 R4-N7-102.4 950 0 R4-N7-G-102.4 900 E;

ese g

fm 2

I 750 21 f

70e -f u

O 3

3 650 1

2 600 550 (iP1 1[

iY_

21 171 2

1

,1

+

l l

I l

l l

0 2

4 6

0 10 12 14 16 18 20 22 24 26 Tlae (si Figure 22.

Surf ace temperature at hot spot on average pin from R4-M7-100,

-102.4, and -G-102.4 calculations.

30

lies i

l I

i l

1 a4-x7-too 1050 G

G b2.4 3

1eGe 5

950 2

~

4, j

m I

800 2

1

(

7se -I u

a 7ee Ne 21 2

600 31#

pi

%1 1

D1"1" 550 1

1 1

I I

50e 0

2 4

6 8

10 12 14 16 18 20 22 24 26 Time ist Figure 23.

Surface temperature at hot spot on average pin in hot channel from R4-M7-100, -102.4, and -G-102.4 calculations.

900 T

I I

I i

1 R4-Ne-ito

/

2 R4-N7-100 8se

_ 3 0 R4-N7-11o 1) g 1

8e8 is

_i o

1 1

,750 7

1 Eg y

-fj O

l i

=

t 1

2 a

g/

6ee

/

2" 33, O gg a

d I

I I

I 500 0

2 4

6 8

10 12 14 16 18 20 22 24 26 7tae tal Figure 24.

Surface temperature at hot spot on hot pin from R4-M6-100 and R4-M7-100 and -110 calculations.

31

1150 I

I I

l l

l I R4-N7-100

{

11M 2 R4*M7-!D2.4 0 R4-M7-C-102.4 1950

- 3 1000

'+

[

m

!=

19 I

k 850 2

-f 1

a 800 3

g 3

a b$

d 21 600 31/

imi bi tj 550 I

I I

I L

I l

0 2

4 6

8 10 12 14 16 18 20 22 24 26 Time (al Figure 25.

Surf ace temperature at hot spot on hot pin from R4-M7-100,

-102.4, and -G-102.4 calculations.

2.5 l

1 I

I I

1 cent INLET 2 B. L. C8LO LEG 8 1. L. CGLD LEG

~

2.0 y

8 1

1.5 i

3' d

1.0 2

2N u-0.5 2

3 2

' h : _'k (m g

W W

24]

A (1

_ y 1

17 1

1 1

1 0.0 1

2 i

-e.5 1f i

1 I

I I

1 l

-1.0 l

0 2

4 6

8 10 12 14 16 18 20 22 24 26 T!ae tal Figure 26. Core inlet and cold leg mass flow rates from R4-M6-110 calculation.

1 32 I

3.s E

I I

1 1

I i cear Isa.sT 2 E. L. CSLD LEG g

2.5 0 1. L. COLD LEG 5

2 2.e 1

[//

1.5 5:!I 3 3

f4

=

1.a j

N I!

t 3% A I

3 1

g

-0.5 2

li 1/

I I

-1.9 0

2 4

6 8

10 12 14 16 18 20 22 24 26 i

Tlae (el Figure 27.

Core inlet and cold leg mass flow rates from R4-M7-100 calculation.

3.s I

I I

I l

cents ris a7 2 8. L. CSLD LEG 2.5 2

8 I. L. COLD LEG v s X

2.0 1 1.5 T

3 2

3

/

I

///

g 1.0 0.0 1

-4.5 2 - 1 l

-1.0

.h/

,1 e

I i

i l

l i

I l

l l

-1.5 d

0 1

2 3

4 5

6 7

8 9

10 11 12 Ttse ts)

Figure 28.

Core inlet and cold leg mass flow rates from R4-M7-110 calculation.

33

u 3.0 I

I I

cenE tm.sT 2 5. L. COLO LEG 2.5 2N O I. L. COLO LEG

's

~

x 2.0

~

k 1.5 3 [

~~

5 4.

g 1.0 00 1

-e.5 2

Li

-1.0 14 1

I I

l i

1 I

I

-1.5 0

2 4

6 8

10 12 14 16 18 20 22-24* 26

.Tjee_.te).

Figure 29. Core inlet and cold leg mass flow rates from R4-M7-102.4 calculation.

3.0 I

I l

l l

ceRE Il0LET

~

2 5.

l.. CeLO LEG 2.5 8 I. L. COLD LEG

'e

~

2.0 13 1.5 lN

//

~

/'

3

/,

5 0.0 1-z

-0.5 2

il

,/

-1.0 1

l l

I l

l I

l 0

1 2

3 4

5 6

7 8

9 10 11 12 Time tel Figure 30.

Core inlet and cold leg mass flow rates from R4-M7-G-102.4 calculation.

34

The R4-M6-llc calculation shows a partial rewet at the average rod in Figure 21, but the surface temperature at the hot spot of the average rod did not reach saturation temperature. However, the R4-M6-110 hot pin cal-culation did not show any of the early quench phenomenon (see Figure 24).

Two explanations can be given for the differences in quench behavior between the R4-M7 and the R4-M6 calculations:

1.

A slight difference in hydraulics (that is, after s6 s, a lower inlet hot channel flow in.ne R4-M6 case than in the R4-M7 case, see Figure 13, and a nearly constant quality in the middle of the hot channel, see Figure 31) which caused the surf ace temperature at the hottest spot to remain above saturation.

2.

The application of the W3-CHF correlation in R4-M6 and the M007 CHF correlation in R4-M7 can also cause the mentioned differences.

The heat transfer modes shown in Figure 32 indicate that the hot

^

spot at the hot rod in the R4-M6 calculation remained in the post-CHF regime (heat transfer Mode 6: Condie-Bengston film boiling correlation), while in the R4-M7 calculation, the hot spot turned back via transition boiling (heat transfer Mode 3: modified Tong-Young correlation) to the nucleate boiling regime (heat transfer Mode 2: Chen correlation).

On the basis of the same temperature difference over the core (at =

32.4 K), the R4-M7-100 Zion I calculation results can be compared with LOFT LOCE L2-3 data. At % s into the transient, in both cases the hottest spot in the core was quenched (see Figure 33) which kept the peak surface cladding temperature low at the hottest spot in the core. On the basis of this early quench phenomenon that occurred during the large break transient, it can be stated that LOFT seems to be prototypic of a large PWR, as far as large break transient system behavior is concerned.

d 35

W 1.2 iiiI l

l l

t R4.Ne-1to Z R4-N7-IDO 1.1 8 R4-NT-110 1.0 Il O.S 1

j 0.8 3

12

/1 y

0.7 C"

e 0.6 f

1 a 1/

I -

0 1

2 h

0.5

~1 0.4 1

1 1"

0.3 3

0.2 2-

\\

0.1 I

I I

I 0.0 e O

2 4

6 8

10 12 14 16 18 20 22 24 26 7tae tal 2-Figure 31. Average quality in middle of hot channel from R4-M6-110 and R4-M7-100 and -110 calculations.

9 l

l l

l 1 R4-MG-110 2 R4-M7-1GD 1

8 R4-N7-It0 8

2" T 7

4 l

1 1

1 21 21 6

12 1 21

i s

4 1

3 2

2

-3 2

2 l

l I

I 1.

I I

I l

1 0

2 4

6 8

10 12 14 16 18 20 22 24 L5 7tae (s)

Figure 32. Heat transf er mode at hot spot on hot pin from R4-M6-110 and R4-M7-100 and -110 calculations.

36

See I

I I

i 1 a4-me-tio 2 R4-N7-100 900

[

8 LDP7 L2-9 DATA

,1% /

950

/*

see

___ t N

\\

3-4 re

,1 65.

k I

^

r~~~~~

f l

i l

l I

I I

500 8

2 4

6 8

10 12 14 16 18 20 22 24 26 Time tal

^

Figure 33.

Surface temperature at hot spot on hot pin from R4-M6-110 and R4-M7-110 calculations and from LOFT LOCE L2-3.

4.2 Calculations for More Conservative Conditions The R4-M7 calculation results from the more conservative initial con-ditions (MLHGR of 50 kW/m and lower flow; and in one case, the more conservative evaluation gap conductance model) showed the early rewet occurrence which keeps the peak cladding temperature relative low in the blowdown phase of a large break transient.

As shown in Figures 22, 23, and 25 for the R4-M7-110, R4-M7-102.4, and R4-M7-G-102.4 cases, the magnitude of the peak cladding temperature increased in the sequence of the cases, and the quenching of the fuel rod occurred later. The increased peak cladding temperature was caused by a higher initial stored heat, as shown in Figure 34 for the hot rod.

The hang-up in the surface cladding temperature, just before it decreased to saturation temperature (the early rewet in the last two cases),

was directly coupled with the heat transfer mode, see Figure 35.

In the 37

0.am I

I I

I R4-M7-100 2 R4-M7-102.4 k"

3 R4-M7-G-102.4 0.055 3

33 0.050 h0.045 3

as

~

0.040 U

s.

0.035 1

0.030

-1 1

E 1

0.025 0.020 N

~

0.015 1* 1-1

~ ~ i~

1 0.010 0

2 4

6 0

10 12 14 16 18 20 22 24 26 Tlae (al Figure 34.

Stored energy at hot spot on hot pin from R4-M7-100, -102.4, and -G-102.4 calculations.

9 i

l l

1 R4-N7-100 2 R4-N7-1DE.4 8

8 R4-N7-G-102.4 7

~

"9 r

igl g

u 2 1 21

' 3 1

1 12 1 J

L 6

3

~

5 2

4

~

i 3

21 i

2 1

1 1

1-l l

I l

l l

1 3-i j

e 2

4 6

8 10 12 14 16 18 20 22 24 26 Time (si l

Figure 35. Heat transfer mode at hot spot on hot pin from R4-M7-100, j

-102.4, and -G-102.4 calculations.

38

case with tne more conservative gap conductance model, the film boiling regime continued over a longer period (heat transfer Mode 6) due to a higher local quality until $9.3 s in the middle of the hot channel, see Figure 36. So the quenching occurred later via the transition to the nucleate boiling heat transfer mode (Mode 2: Chen's heat transfer correlation).

The pump speed in the intact loop is shown in Figures 37 and 38 for the dilferent cases.

In the 102.4% power cases, the pump initial speed was lower than the nominal speed (1100 rpm) which produced a lower flow (93%

nominal flow). Directly after the break occurred, the pump ran at the same speed as in the other cases during the transient (1189 rpm) for a certain period.

In the broken loop, the pump showed a turbine-type behavior immedi-ately after the break, see Figures 39 and 40.

The pump speed increased nearly instantly from 1100 rpm (1189 rpm in R4-M6 calculation) to 1218 rpm, after vhich it fluctuated.

(The increase seems to be nonrealistic.)

.x

1. 2 -

1 I

I I

1 R4-N7-100 2 R4-M7-102.4 1.1 8 R4-N7-G-102.4 1.0

~

/

0.9

~

0.0

- 31 U

0.7 1

-f 0.6 I

k d

lls i

0.5 7

I 1

0.4

[

(

0.3 3

12 2

%1 0.2 l'

i 1

0.1 f

1 L

l I

l l

l I

l l

I I

f g,g 3 0

2 4

6 8

10 12 14 16 18 20 22 24 26 Time (cf Figure 36.

Average quality in middle of hot channel from R4-M7-100,

-102.4, and -G-102.4 calculations.

39 l

l

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

1206 I

I I

a4_we.iio 1204 2 R4-M7-!00 1202 1200 J

12 1198 lyi 11 %

~

1 S 1194 3

f 5 1192 D-2 1190 itse T

1 4

1186 21 1

E I

I I

l l

I l

l l

l 11g 0

2 4

6 8

10 12 14 16 18 20 22 24 26 T!ae (c)

Figure 37.

Pump speed in inta:t loop from R4-M6-100 and R4-M7-100 calculations.

1220 I

l l

l 1 R4-H7-100

" R4-M7-102.4 1210 3 R4-H7-G-102.4 1200

~

1

-1"1 1

2 1

1 1._

13 1190 I v 31-321 332 1 1180 Ef 1170

1160 3-m 1

- 1150 f

114e 1130 1120 1110

&l I

I I

1100 O

2 4

6 8

10 12 14 16 18 20 22 24 26 Time ts)

Figure 38.

Pump speed in intact loop from R4-M7-100, -102.4, and -G-102.4 calculations.

40 k-

1230 E

I I

I I R4-MO-llo 2,

2 R4-M7-300

/

71 1225 1

/

1 1

1220 12 % 3 2

'd 1

1 1215 1

i l12 2

1210 1

1 1205 2x -

a 21 2

1200 2

1195 1

1190 1LI I

I I

13,

O 2

4 6

8 10 12 14 16 18 20 22 24 26 Time (c)

Figure 39.

Pump speed in broken loop from R4-M6-110 and R4-M7-100 calcul at ions.

1260 l

l l

1 l

I 1 R4-M7-100 2 R4-H7-102.4 1243 3 R4-H7-G-102.4 1

1-f1N ak s

1 1 31 2 1 1

1 1~ _

i 1200 13

,i 1180 m

a 1160 a.

1140 1120 3

L I

I I

I lim

' a' O

2 4

6 8

10 12 ~ 14' 16 10 20 22 24 26 Tlae (el Figure 40.

Pump speed in broken loop from R4-M7-100, -102.4, and -G-102.4 calculations.

41

l S.

CONCLUSIONS The conclusions reached from comparing and evaluating the differences among data from large break LOCA calculations using RELAP4/M006 and RELAP4/

MOD 7_ computer codes and from LOFT LOCE L2-3 are summarized as follows:

1.

The 200% double-ended cold leg break calculations using RELAP4/

M007 for the Zion 1 nuclear power plant showed a significant early rewet behavior at all fuel rods that caused good heat removal in the early part of the translent.

2.

The RELAP4/M006 Zion 1 calculatior serformed for the LOFT proto-2 typicality study showed a tendency to rewet the hottest spot at the average rod early in the trrnsient, but did not show an early rewet at the hottest spot at the hot rod.

The differences in results between the RELAP4/M006 and the RELAP4/M007 calculations were partly due to differences in the hot rod modeling technique,a but were mainly due to the differences in RELAP4 code versions, with emphasis on differences in heat e

transfer correlations. The hot rod calculation should be performed as an integral part of the system calculation to take in account the total system effects, as done in the RELAP4/M007 calculations, instead of performing a separate hot rod calculation and using the boundary conditions from the system calculations, as done in the RELAP4/M006 calculation.

l l

3.

The Zion 1, RELAP4/M007 ca..alation for 100% nominal power is comparable with LOFT LOCE L2-3 on basis of the same temperature l

l cifference across the core. The calculation results showed approximately the same_ peak cladding temperature and the same quenching phenorrenon (at $5.5 s into the transient) as in the a.

The RELAP4/M006 hot rod calculation is a separate calculation, using the core F andary conditions from the average rod system calculations. The RELAP4/i4007 hot rod calculation is an integrated system calculation with an

_ average channel and a hot channel (with hot rod) modeled.

42

_a, experiment at the hottest spot on the hot rod. The study showed that early rewet occurred with nominal and conservative core t ut hydraulic conditions and fuel peak linear heat generation rates 49 that bracket current commercial plant operating conditions. The early rewet, which occurred during Experiment L2-3, is predicted.

to occur in a four-loop commercial PWR of the Westinghouse Zion 1 type.

i 4.

Even for more conservative initial conditions in the Zion 1 nuclear power plant (such as 102.4% of the. nominal power,

{

MLHGR = 50 kW/m, 93% of the nominal flow, and the evaluation I

option for the fuel rod model), the early rewet occurred at the hottest spot in the core during a large break transient. How-ever, greater initi i stored energy increased the magnitude of the neak cladding tempe.ature and delayed the cladding' quench.

4 A

1 i

a 6

k

)

i lj f

em.

43

6.

REFERENCES i

1.

D. L. Reeder, LOFT System and Test Description (5.5-Ft Nuclear Core 1 LOCEs), NUREG/CR-0247, FREE-1208, July 1978.

2.

L. Winters, Large Break Transient Calculations in a Commercial PWR and LOFT Prototypicality Assessment, EGG-LOFT-5093, April 1980.

3.

D. L. Reeder, Quick-Look Report on LOFT Nuclear Experiment L2-3, QLR-L2-3, May 1979.

t 4.

S. R. Behling et al., RELAP4/M007--A Best Estimate Computer Program to Calculate Thermal and Hydrualic Phenomena in a Nuclear Reactor or Related System, Draft, March 1981.

5.

T. L. DeYoung, RELAP4/M007 Developmental Checkout:

ZION, EGG-CDAP-5396, May 1980.

6.

G. W. Jennsen, F. W. Childs, T. M. Broughton, A Comparison of "Best-Estimate" and " Evaluation Model" LOCA Calculations:

The BE/EM Study, Report No. PG-R-76-009, December 1976.

~

7.

H. Chow, private communication, EG&G Idaho, Inc., June 1979.

8.

C. E. Hendrix, private communication, EG&G Idaho, Inc., February 1979.

9.

T. C. deBoer, private communication, EG&G Idaho, Inc., June 27, 1980.

10.

T. C. deBoer, private communication, EG&G Idaho, Inc., August 26, 1980.

t I

44

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

?.

OUT 110=*/ THIS LPDATE CORRECTS AN ERROR IF 3N + 1 ENTRIES HRC ENTERED 120=*/ ON A FILL TABLE. TK ERROR CAUSED OtE GMBHGE t&1BER IN 130=*/ TE FILL TABLE 140=*I IFFIG99PH.65 150-IF (IL. NITS.LT. 3) GO TO 617 160=*ID HTS 2 - SB 170=*/ THIS LPDATE CORRECTS AN ERROR INADVERTANTLY IP IfrO IN MOD 7.

~~100=*/~~

~~TE LPDATE IS tECESSARY IF TE VOID FRACTION IN A VOLLFE 190=*/ WITH A EAT SLAB EXCEDES 0.92.~~

~~200=*D HTS 2G99CH.1 t~~

~~i 210=0 DATA ACRIT /O.96E0/~~

~~220=*ID INT:~~

~~OB 230=*/ THIS LPDATE CORRECTS AN ERROR IN CALCLLATING DEAD END VOLtPE 240=*/ PFm RES DLRING SELF-INITIR_IZATION.~~

~~250=*D INTBG99KS.4 260=~~

~~NJSC,IJUND,NDEAD,NJD,FIRST1,LAST1) 270=*/~~
~
280=*/ ERRORS HAVE BEEN FOLND IN TE FILL TABLE ItFLIT ROUTItES MEN 290=*/ SI LfiITS ARE USED.
CORRECTIONS HAVE NOT ' RET BEEN FORitLATED.
300=*/
310=*/ THIS LPDATE CORRECTS AN ERROR IN TK REACTIVITY CALCLLATION.
320=*/ NO ERROR EXISTED ON_Y IF T}E CORE SECTIONS BEGAN 330=*/ WITH TE FIRST SLAB AND ERE t&1BERED CONSECUTIVELY.
340=*/
THIS LPDATE IS tECESSMY OtLY IF REACTIVITY CALCLLATION 350=*/
IS WMTED 360=*ID RINI-Z-SB 370=*I RINI.17 l
300=
II = 0 390=*D RINI.26,29 400=
II = II + 1 410=
RDC4. = RDCR. + DOPWT(II)*POLATE(DOPRO,TM(K),NDOP,IDOP(II))
420=
RWCAL = RWCAL + ALPHTW(II)*TEPCFF(J) 430=
RFCAL = RFCAL + ALPHTM(II)*TM(K) 440=
RYCAL = RVCAL + VOIDWT(II)*POLATE(VOIDRO,1.,NVOID,IVOID(II))
450=*ID REAC-2-SB 460=*I REAC.251 470=C II IS TK TABLE LOCATION FOR DOPWT, ALPHTW, ALPHTM,VOIENT 400=
II = 0 490=*I REAC.254 500=
II = II + 1 510=*D REAC.255,257 520=
RD = RD + DOPWT(II)*POLATE(DOPRO,TM(I),NDOP,IDOP(II))
530-RW = RW + ALPHTW(II)*TEPEFF(J) l 540=
RF = RF + ALPHTM(II)*TM(I) 550=*D REACG6eJT.4 i
560=
RV = RV + VOIDWT(II)*POLATE(VOIDRO, VOID,NVOID,IVOIDCII))
570=*/
THIS LPDATE CORRECTS A PROBLEM WITH TOO MUCH PRINT MEN LIQUID 500=*/ MASS DEPLETIONS OCCLR. WFECTS PRINT ON_Y MEN "tEW" WATER 590=*/ PACKING IS USED (W2-R CMD ormR4.LT. IffINITY).
600=*ID NIFT - DS 61B=*I NIFTG94DS.383 620=
FWSE(1,I) =1 47
w 630=*/ THIS LPDATE CORRECTS AN ERROR IN TWC 21 OECK VALVES IF USED IN A 640=*/ A LEAK JLNCTION MD TK LEM TAILE N.NEER letS GREATER THAN 1.
8 650=*ID CW V CB 66Be*D CWVG98SB.7 670=
IF(N.GT. 0) XSIFK = SIfE(N)
SEEl=*ID PLLL SB 690=*/ THIS LPDATE CORRECTS A RESTART ERROR IF REACTOR KIETICS IS 700=*/ USED. TE ERROR INCORRECTLY EXPECTED A POER VS, TIE TO BE 710=*/ READ IN ON T@E2 IF NODEL (POER TYPE) WAS EQUAL TO 2 OR 3 720=*/ AND TE CALCLLATED NORMALI2ED fEUTRON FLLDC 146 LESS THAN 1E-0.
730=*D PLLL.291 740=
NODELX = 0 750 IF(NODEL.EQ. -1) NODELX = 1 760=
IN2 = MAXBCIN2,NODELX) 770=*/ TE FOLLOWING LPDATE CORRECTS m ERROR IN TE ArMLM FLOW 780=*/ SLIP CORRELATION. TE ERROR CAUSED PROILEMS WITH EXCESSIVELY HIGH 790=*/ SLIP VELOCITIES ON SORE OCCASIONS AND CAUSED DISCONTINJTIES IN 000=*/ TE CALCLLATED SLIP VELOCITIES AS TE MASS FLLb( CHMGED.
810=*/ TE LPDATE CHANGES TE DEFINITION OF TE LIQUID FILM 820=*/ THICKESS FOR RMLAR FLOW.
83B=*/ THIS LPDATE WILL RESLLT IN A Sr5LLER CALCLLATED SLIP VELOCITY (ABOUT 840=*/
O E-THIRD SPALLER) IN TE ArMLM FLOW REGItE. T E USE OF THIS 850=*/ LPDATE WILL CWNGE TE RESLLTS OF NOST CALCLLATIONS UTILIZING TE 060=s/ FLOW REGIE DEPENDENT SLIP CORRELATIONS, TWORE, TE LPDATE IS 870=*/ NOT OPERATIONAL IN THIS FILE.
880=*/ *****TO t%KE LPDATE OPERATIONAL CHmGE "/" TO "D" ON TEXT CMD....
890=*/ SVELG95KF.56 900=*/ *****TO MAKE LPDATE OPERATIOrR. REMOVE "*/" ON TEXT CMD.....
910=*/
DELTA = 0.5*(DHY - SQRT(DHY*DHY - (1.-A63)*(DHY*DR/-DIN
~~DIN)))~~
920=*/ *****TO P54(E LPDATE OPERATIOr5L CHmGE "/" TO "D" ON TEXT CMD...
930=*/ SVELG95HF.09 940=*/ *****TO PAKE LPDATE OPERATIOr5L REMOVE "*/" ON ffXT CARD.....
950=*/
DELTA = 0.5*(DHY - SQRT(DR(*DR( - (1.-ALPH)*(DR(*DR(-DIN
~~DIN)))~~
s D
48
- - _ -.. - _...~. -.-,.
A a
APPENDIX B ZION 1 POWER DISTRIBUTION DATA A
f
'd I
1 f
49 i
APPENDIX B ZION 1 POWER DISTRIBUTION DATA A
U Power distribution data for extreme initial conditions (evaiuation model conditions) used in the Zion I large break loss-of-coolant accident a
calculations were provided by Westinghouse Electric Corporation and are listed in Table B-1.
TABLE B-1.
POWER DISTRIBUTION VERSUS AXIAL POSITION Axial Power Distribution (kW/ft)
Axial Hottest Core Heat Distance a
b c
Elevation Slab (ft)
Average Rod Rod Adjacent Rod Bottom 1
0.00 1.782 2.126 2.023 2
1.50 6.359 7.585 7.217 3
3.00 10.122 12.074 11.688 4
4.00 11.936 14.237 13.566
+
5 5.00 13.067 15.587 14.830 6
5.50 13.355 15.930 15.157 7
5.75 13.427 16.016 15.237 8
6.00 13.451 16.045 15.266 9
6.25 13.427 16.016 15.237 10 6.50 13.355 15.930 15.157 11 6.75 13.235 15.767 15.020 12 7.00 13.067 15.787 14.830 13 7.25 12.852 15.331 14.586 14 7.50 12.591 15.020 14.290 15 7.75 12.286 14.655 13.943 16 8.00 11.936 14.232 13.566 17 9.00 10.122 12.074 11.680 18 10.50 6.359 7.585 7.217 Top 19 12.00 1.782 2.126 2.023 Initial core coolant flow 37 674.48 lbm/sec (+93% nominal flow)
Power level 3 315 MW d
Core average LHGR 6.916 kW/ft Peaking factor 2.32 7
a.
Average rod in hot channel.
V b.
Hottest rod in hot channel.
c.
Adjacent hot rods around hottest rod in hot channel.
d.
LHGR--linear heat generation rate.
a.
T. C. deBoer, private ccmmunication, EG;.G Idaho, Inc., August 26, 1980.
51

ML20037D280

Text

Title:

SUMMARY

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