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Evaluation of XN-NE-77-25, Exxon Nuclear Co (ENC) Evaluation of Two-Loop Westinghouse PWR Containment Using WREM-II ECCS Model:Large-Break Example Problem. Rept Acceptable for Ref in Applications for Ols,Cps & Amends
	ML20148C148
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Issue date:	06/27/1978
From:	; Office of Nuclear Reactor Regulation
To:	;
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Topical Report Evaluation Report Number:

XN-NF-77-25 Report Titie:

Exxon Nuclear Company ECCS Evaluation of a 2-Loop I

Westinghouse PWR With Dry Containment Using The I

ENC-WREM-II ECCS Model-Large Break Example Problem Report Date:

August 1977 I

Originating Organization:

Exxon Nuclear Company Summary of Topical Report The ENC-WREM-II model used by Exxon to perform the large break ECCS l

calculations reported in this topical consists of the RELAP4-EM/ ENC 26A code for the blowdown and reflood transients, and the T00DEE2/Apr. 77 I

code for hot pin thermal transients.

Both codes have been modified

{

since NRC approval of the original ENC-WREM-II model described in XN-76-27(l); however, with the exception of post CHF return-to-nucleate l

boiling lockout (2), no model changes have been made among the several I

modifications made to the original ENC-WREM-II.

Topical Report Evaluation The ENC-WREM-II model approved in 1973 carried the Exxon designation 95 ENC Version 20.

Following this approval, modifications reported in l

XN-76-36(4) for Versions 21 and 22, in XN-76-51(5) for Versions 23 to 25, l

and in XN-NF-77-24(6) for Versions 26 and 26A have been made which have consisted of minor program corrections or improvements described in these reports. Of the modifications made, only the modification made in Version 24(2) for post CHF return-to-nucleate boiling lockout, and i

in Version 25 for time step control produce computed output changes.

The first change was incorporated in the Exxon model by NRC directive, and the second change was made for computational efficiency, and had the effect of increasing end-of-bypass temperatures by approximately 12 F.

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F

' Staff Position The ENC WREM-Il modifications contained in the current RELAPa-EM/ ENC 26A, and the T00DEE2/ April 77 versions of these codes are acceptable for use in licensing calculations as no basic model changes are involved with the exception of the nucleate boiling lockout feature required by the NRC.

This change was considered in the review of XN-76-44 and found acceptable.

The ENC-UREM-Il oeneric model would be acceptable for 2-loop PWR's if they injected all ECCS fluid into either the cold legs or the downcomer (which they do not).

This model is not acceptable for the plants as currently 4

configured, which inject ECCS fluid into the upper plenum, unless the fluid interaction is treated in a phenonenologically satisfactory manner or the calculations with the current model can be shown to be demonstrably conservative for all potential ranges of use for the generic model.

This f

was not provided in the report reviewed.

The case by case restrictions specified in the NRC approval of the Exxon Nuclear Company's WREM-Based PWR ECCS Evaluation Model(7) also apply to application of the ENC-WREM-Il model I

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References:

a 1.

L.C. Worley, D.S. Rowe, and K.P. Galbrarth, " Exxon Nuclear Company g

WREM - Based Generic PWR ECCS Evaluation Model Updata ENC-WREM-II,"

XN-76-27, August 2, 1976.

2.

" Revision 1 to Safety Evaluation Report on the Exxon Nuclear Company I

WREM - Based Generic PWR-ECCS Evaluation Model Update ENC-WREM-II for Conformance to Requirements of Appendix K to 10CFR50 by the Office of Nuclear Reactor Regulation," dated January 5,1977.

3.

L.C. Worley, " Revised Nucleate Boiling Lockout for ENC WREM-Based ECCS Evaluation Models," XN-76-44, October 1, 1976.

4.

Exxon Nur'cor Company WREM-Based Generic PWR-ECCS Evaluation Model (ENC-GEM-II) 4 Loop PWR With Ice Condenser Large Break Example Problem," XN-76-36, (pg. 56), August 27, 1976.

5.

Donald C. Cook Unit 1 LOCA Analyses Using the ENC WREM-Based PWR ECCS Evaluation Model (ENC-WREM-II), XN-76-51, (pg. 48), October 26, I'

1976.

6.

W.V. Kayser, "LOCA Analysis for Palisades Type D Fuel at 2530 Mwt i

Using The ENC-WREM-II PWR ECCS Evaluation Model," XN-NF-77-24, (pg.83), July 1977.

'3 7.

" Safety Evaluation Report Regarding Review of the Exxon Nuclear Company E

Pressurized Water Reactor Generic ECCS Codes and the H.B. Robinson ECCS Evaluation Model for Conformance to all Requirements of Appendix K to 10 CFR 50,"

Office of Nuclear Reactor Regulation, September 11, 1975.

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1 m 0, C0 m N1S Page

1.0 INTRODUCTION

AND

SUMMARY

1 2.0 BLOWDOWN CALCULATION......

5 2.1 MODEL DESCRIPTION.................

5 2.2 RESULTS......................

5 3.0 HOT CHANNEL CALCULATION...............

19 3.1 MODEL DESCRIPTION................

19 3.2 RESULTS.....................

20 4.0 REFILL CALCULATIONS.................

28 4.1 RELAP4 POWER..................

28 4.2 RELAP4 REFILL..................

28 4.3 B0CREC CALCULATION...............

29 5.0 RE,LOOD CALCULA110N.................

3, g

5.1 RELAP4-EM/ FLOOD MODEL DESCRIPTION........

31 g

5.2 RELAP4-EM/ FLOOD RESULTS.............

33 6.0 HEATUP CALCULATION..................

40 6.1 MODEL DESCRIPTION................. 40 6.2 R E SU LT S.....................

41 7.0 CONTAINMENT BACKPRESSUR2 CALCULATION.......... 45

8.0 REFERENCES

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i XN-NF-77-25(4)

-ii-LIST OF TABLES Table No.

Page 1.1 ANALYSIS RESULTS OF EXAMPLE PROBLEM (DECLS C = 1.0)....

3 D

1.2 LARGE BREAK RESULTS TIME SEQUENCE OF EVENTS.........

4 2.1 WESTINGHOUSE 2-LOOP PWR DATA................

7 3.1 RELAP4-EM/ HOT CHANNEL RESULTS FOR N0DE 18.........

21 7.1 DRY CONTAINMENT DATA..............

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-iii-XN-NF-77-25fA) i l

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LIST OF FIGURES Figure No.

Page 2.1 RELAP4-EM BLOWDOWN SYSTEM N0DALIZATION FOR WESTINGHOUSE 2-LOOP PWR................

9 2.2 UPPER PLENUM PRESSURE, DECLS (C D

2.3 CORE INLET FLOW, DECLS (CD = 1.0)...........

11 2.4 CORE OUTLET FLOW, DECLS (C =1.0)..........

12 D

2.5 DOWNCOMER FLOW, DECLS (CD = 1.0)...........

13 2.6 PRESSURIZER SURGE FLOW, DECLS (C = 1.0).......

14 D

2.7 INTACT LOOP ACCUMULATOR FLOW, DECLS (CD = 1.0)....

15 2.8 BROKEN LOOP ACCUMULATOR FLOW, DECLS (C = 1.0)....

16 2.9 BREAK FLOW, DECLS (CD = 1.0).............

17 2.10 BREAK VOLUME PRESSURE, DECLS (CD = 1.0)........

18 l

3.1 i

RELAP4-EM HOT CHANNEL CORE VOLUMES AND JUNCTIONS...

22 3.2 H0T CHANNEL HOT R0D HEAT SLAB N0DALIZATION......

23

!E 3.3 CLADDING TEMPERATURE @ N0DE 18 DURING BLOWDPWN,

'W DECLS (CD=1.0)...................

24 3.4 I

DEPTH OF METAL-WATER REACTION FOR N0DE 18 DURING BLOWDOWN.........

25 3.5 HEAT TRANSFER COEFFICIENT FOR N0DE 18 DJRING BLOWDOWN....................... 26 3.6 HEAT TRANSFER REGIME AT N0DE 18 DURING BLOWDOWN, DECLS (C = 1.0)..............

27 D

4.1 NORMALIZED POWER REFILL AND REFLOOD..........

30 5.1 RELAP4-FM/ FLOOD NODALIZATION DIAGRAM FOR WESTINGHOUSE 2-LOOP PWR.

34 5.2 CORE FLOODING RATE, DECLS (CD = 1.0).........

35 5.3 CORE MIXTURE LEVEL, DECLS (CD = 1.0).........

36 I

I XN-NF-77-25(A) i v-LIST OF FIGURES (Continued)

Figure No.

Page 5.4 DOWNCOMER MIXTURE LEVEL, DECLS (CD = 1.0).......

37 5.5 UPPER PLENUM PRESSURE, DECLS (CD = 1.0)........

38 5.0 CORE INLET FLOW RATE, DECLS (CD = 1.0)........

39 6.1 T00DEE2 HOT R0D HEAT SLAB NODALIZATION........

42 6.2 T00DEE2 RADIAL P0 INT AND BOUNDARY ASSIGNMENTS FOR WESTINGHOUSE 2-LOOP PWR HOT CHANNEL ANALYSIS...

43 6.3 CLADDING TEMPERATURE VERSUS TIME, DECLS (CD = 1.0)..

44 7.1 CONTAINMENT BACKPRESSURE FOR DECLS (C 1.0).....

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1.0 INTRODUCTION

AND

SUMMARY

This document is presented as a demonstration of the ENC WREM-II II'2) i ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example).

The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area.

The break was assumed to occur in one pump discharge pipe (DECLS break).

The analyses involved calculations using the ENC WREM-II model. Speci-fically, the following codes eiere used:

RELAP4-EM/ ENC 26A for blowdown and hot channel analyses, RELAP4-EM FLOOD / ENC 26A for core reflood analysis, CONTEMPT LT/22 modified as given in CSB 6-1 for containment backpressure analysis, and T00DEE2/APR77 for heatup analysis.

This set of computer codes represents the latest ENC versions of the ENC WREM-Based models.

The ENC 26A version of RELAP4 incorporates the necessary changes to the time step opti-mization package to allow application to a more general class of problems, including REFLOOD and REFILL.

The evaluation of End-of-Bypass (E0BY) has been updated to reflect available data from cold leg steam-water mixing I

studies.

The refill calculation includes the hot wall delay model presented in ENC WREM-II.

T00DEE2/APR77 incorporates those portions of the ENC WREM-II model which are applicable to Westinghouse 2-loop PWR's, specifically, the l

flow reduction due to the blockage models. The FLECHT heat transfer model remains identical to the ENC WREM-I model since its data base covers the range of conditions experienced in a Westinghouse 2-loop PWR.

The criteria and assumptions used in the evaluation of the LOCA are those listed under Section 50.46 and Appendix K of Title 10 of the Code of Federal Regulations.

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'l XN-NF-77-25(E) lE l

5 Principal results of these analyses for the DECLS break are contained in Tables 1.1 and 1.2.

These results include a Peak Cladding Temperature (PCT) of 1951 F and a ma).imum local Zr/H O reaction of less than 5%. These 2

analyses were performed at 1550.4 MWt which is 102 percent of rated power as indicated in Table 1.1.

Removing the Appendix K required two percent core i

power uncertainty factor, the analysis supports operation of the plant with a total Linear Heat Generation Rate (LHGR) of 13.76 kW/f t and with a total T

peaking (F ) of 2.32.

Also included in Table 1.2 are summaries of the g

transient times calculated for major events. Additional results of this analysis are prest.ted in subsequent sections of the report.

l These results indicate that full power operation with a core peaking l

factor, F, of 2.32 satisfy the Appendix K criteria when the ENC WREM-II q

ECCS model is applied to a Westinghouse 2-loop PWR with dry containment.

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i XN-NF-77-25(A)

TABLE 1.1 ANALYSIS RESULTS OF EXAMPLE PROBLEM (DECLS C = 1.0) g I

Analysis Results DECLS I

Peak Clad Temperature, F 1951 Peak Clad Temperature Location, ft from Bottora 7.17 Local Zr/H 0 Reaction (max), %

< 5%

2 Local Zr/H 0 Location, ft from Bottom 5.92 2

Total H Generation, % of Total Zr Reacted 1.0 2

Hot Rod Burst Time, Sec 43.5 Hot Rod Burst Location, ft from Bottom 5.92 Peak Linear Heat Generation Rate, B0CREC, kW/ft 0.704 Calculation License Core Power MWt 1520 Power Used for Analysis, MWt 1550.4 Peak Linear Power kW/ft 13.76*

Total Peaking Factor 2.32 I

All power generated in fuel at 1520 MWt.

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I XN-NF-77-25(A)

I TABLE 1.2 LARGE BREAK RESULTS TIME SEQUENCE OF EVENTS I

Event Time of Event (seconds)

I DECLS Start 0.0 J

Initiate Break 0.05 I

Safety Injection Signal 0.50 Accumulator Injection, Broken Loop 0.7 Pressurizer Empties 6.1 Accumulator Injection, Intact Loop 6.4 End-of-Bypass 16.80 I

Safety injection Flow, SIS 26.20 Start of Reflood 38.26 Accumulators Empty, Intact Loop 45.25 Peak Clad Temperature Reached 113.1 I

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I XN-NF-77-25(6)

E 2.0 BLOWDOWN CALCULATION 2.1 MODEL DESCRIPTION The input data for the example calculation were chosen to be typical of a Westinghouse 2-loop pressurized water reactor with a dry l

containment.

Table 2.1 gives a general listing of some of the plant i

parameters used to develop the input for this model.

l The system was modeled using 48 volumes, 60 junctions and 41 heat slabs (Figure 2.1).

The model includes two accumulators, one pressur-izer, and two vertical U-tube steam generators, with both primary and secondary sides of the steam generators modeled.

The high and low pressure Safety Injection System (SIS) flows were modeled as fill junctions with typical flow rates given as a function of system backpressure. The reactor core power is calculated by the RELAP4-EM solution of the space independent core kinetics equations with radioactive decay energy (ANS + 20) and actinide contributions. Mass and energy release rates from the primary coolant system to the containment were used as input to the CONTEMPT LT/22 code for containment backpressure analysis. The pump performance curves were taken for the Westing-house pump, specific speed 5200, available in the RELAP4-EM program {3)

The reactor core was modeled radially as an averaged core plus a single hot assembly, each with three axial nodes. The upper head fluid temperature was set equal to the hot fluid temperature. Major hardware components were modeled using 38 heat conductors.

2. 2 RESU_L.TS Results from the application of RELAP4-EM program to the blowdown of the system are presented in Figures 2.2 through 2.11.

The timing of various major blowdown events are listed in Table 1.2.

I

iI XN-NF-77-25(A)

The results are typical of pressurized water reactor large break blowdown analyses.

For large break analyses the system decompresses rapidly to the saturation point, and the pressure then continues to decrease smoothly to an ambient condition. After break initiation the core inlet flow reverses, approaching a zero flow condition for several seconds before reversing again.

The core flow approaches zero at the End-of-Bypass (E0BY). The evaluation of E0BY has been updated to reflect available data from cold leg steam-water mixingstudies(4,5,6,7,8,9) per the procedure presented in Reference 10.

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TABLE 2.1 WESTINGh3USE 2-LOOP PWR DATA **

Primary Heat Output, MWt 1520

7 Primary Coolant Flow, lbm/hr 6.8 x 10 3

Primary Coolant Volume, ft 9534.***

Operating Pressure, psia 2250 Inlet Coolant Temperature, F

544.5 3

Reactor Vessel Volume, f t 2473.

3 Pressurizer Volume, Total, ft 800.

3 Pressurizer Volume, Liquid, ft 480.

Accumulator Volume, Total, f t3 (each of two) 1750.

3 Accumulator Volume, Liquid, ft 1100.

l Accumulator Pressure, psia 714.7 2

Steam Generator Heat Transfer Area, f t 39,987.

6 Steam Generator Secondary Flow, lbm/hr 3.13 x 10 Steam Generator Secondary Pressure, psia 770.

Reactor Coolant Pump Head, ft 252.

Reactor Coolant Pump Speed, rpm 1189.

2 l

Moment of Inertia, lbm-ft / rad 80,000.

Cold Leg Pipe, I.D., in 27.5 Hot Leg Pip,

I.D., in 29.0 Pump Suction Pipe, I.D., in 31.0 I

Primary Heat Output used in RELAP4-EM Model = 1.02 x 1520 = 1550.4 MWt.

- As taken from R. E. Ginna nuclear plant.

- - Includes total accumulator and pressurizer volumes.

I

I XN-NF-77-25(A)

I TABLE 2.1 (Continued)

I Fuel Assembly Rod Diamter, in*

0.424 Fuel Assembly Rod Pitch, in*

0.556 Fuel Assembly Pitch, in*

7.803 5

Fueled (Core) Height, in*

142.0 2

Fuel Heat Transfer Area, ft 28,450 2

Fuel Total Flow Area, ft 26.80 Steam Generator Tube Plugging (Assumed uniform) 10%

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I

. XN-NF-77-25(E) 3.0 HOT CHANNEL CALCULATION 3.1 MODEL DESCRIPTION The RELAP4-EM/H0T CHANNEL model is used 1) to determine the first PCT occurring during the blowdown phase, and 2) to establish the temperature profile and extent of the metal-water reaction at the E0BY for input into the fuel rod heatup code T000EE2.

The hot channel model employed was nodalized to be compatible with both RELAP4-EM/ BLOWDOWN and T000EE2.

The model contains eight volumes and eleven junctions as depicted in Figure 3.1.

Volumes 1 and 8 (lower and upper plenums) correspond to RELAP4-EM/ BLOWDOWN, Volumes 32 and 2, respectively.

RELAP4-EM/ HOT CHANNEL uses the time dependent volume conditions from RELAP4-EM/

BLOWDOWN for these volumes.

The model uses 22 heat slabs.

Six of these represent the average core and hot assembly, and the remaining 16 heat slabs are allocated to the hot fuel rod.

The hot fuel rod slabs are of varying height, with a concen-tration of 3-inch slabs around the point where the peak temperatures are expected.

The heat slab nodalization of the hot fuel rod is detailed in Figure 3.2.

This axial nodalization of the hot fuel rod is identical to the nodalization used in T00DEE2.

A chopped cosine axial power profile was chosen to represent the conditions for LOCA.

From this profile the axial power profile used in the blowdown and hot channel models was developed. This axial profile I

was applied to both average and hot fuel assemblies. Axial peaking was 1.398 and total peaking was 2.32.

I

e I XN-NF-77-25(A) i 3.2 RESULTS Results for the node which ultimately becomes the PCT node during the LOCA transient are given in Table 3.1.

The PCT node temperature history and depth of metal-water reaction are shown in Figures 3.3 and 3.4, respec-tively. Heat transfer coefficients are presented in Figure 3.5 and the I

O,3) corresponding heat transfer regime number as defined in the WREM program is indicated in Figure 3.6.

The PCT node in the hot channel analysis, which is node 18, corre-sponds to node 13 in the T00DEE2 analysis.

I I

I

I XN-NF-77-25(4)

TABLE 3.1 i

RELAP4-EM/ HOT CHANNEL RESULTS FOR N0DE 18 I.

I Node 18 (Blowdown Conditions) 1.0 DECLS Blowdown Peak Temperature, F

1431 Elevation of Peak, Feet 7.17 Time of Blowdown Peak, Seconds 3.68 Clad Temperature at End-of-Bypass, F

1290 I

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i l

' Xti-NF-77-25(%)

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LOWER PLENUM JUNCTIONS I

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FIG'JE 3.1 RELAP4-EM HOT CHANNEL CORE '^ JMES AND TIONS I I

I XN-NF-77-25(A)

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l 21 I

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I XN-NF-77-25(A)

I 6.0 HEATUP CALCULATION 6.1 MODEL DESCRIPTION The time dependent fuel rod thermal analysis program T00DEE2 is used to determine both the PCT and extent of metal-water reaction during the refill and reflood periods of a PWR LOCA. The hot fuel rod from the I

hot assembly is modeled with a total peaking factor of 2.32.

The hot fuel rod is divided into 16 axial nodes as shown in Figure 6.1.

As in the hot fuel rod nodalization used in the hot channel analysis, the nodes are of varying heights with the smaller nodes (3-inch) concentrated in the region of the expected maximum temperature. The fuel rod is divided into ten radial nodes, comprised of two cladding nodes, seven equally spaced fuel nodes, and one fuel / gap node. The radial nodalization is shown in Figure 6.2.

The axial power distribution corresponds to that used in RELAP4-EM/H0T CHANNEL.

The code requires input from two sources, the initial fuel rod temperature distributions and depths of metal-water reaction from RELAP4-EM/

HOT CHANNEL calculated values at the end-of-bypass.

The time dependent fluid conditions (flooding rate, inlet enthalpy, etc.) are taken from RELAP4-EM/ FLOOD results.

During the period from end-of-bypass to beginning of reflood the ENC radiation model (Section 7.0 in Volume I, XN-75-41)(I) is conservatively neg-lected. After reflood, heat transfer coefficients are determined using the heat transfer model from the WREM-I model since its data base covers the range of conditions predicted in the Westinghouse 2-Loop reactor.

The heatup portion of the transient has been calculated by the methodasreportedinXN-75-41(I)andXN-76-27{2) The results of the calcu-lation with this model are shown in Figure 6.3.

I

l XN-NF-77-25(A)

I The peak cladding temperature occurs at the 7.17 foot elevation.

The temperature at this location turns around at 113.1 seconds and continues to decrease throughout the remainder of the transient.

6.2 RESULTS Peak clad temperatures and corresponding times are presented in I

Tables 1.1 and 1.2.

Also included in this table are clad rupture times and clad temperature and linear heat generation rates at initiation of reflood.

The temperature history for the node of peak cladding temperature from the end-of-bypass through temperature turn-around is plotted in Figure 6.3.

The PCT Node in the T00DEE2 analysis, which is node 13, corresponds to node 18 in the hot channel analysis.

I I

^ XN-NF-77-25(nf I

I 142.0 g

17 I

118.333 16 l-3 g -

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I FI?:'RE 6.1 T00DEE2 HOT ROD

EAT SLAD NODALIZATION i
I I
l
, Xft-NF-77-25(A)
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I 10 GRID LII!E LOCATICNS 11 P0liii LOCATIONS I
-FIGRE 3.2 T00 DEEP MDIAL DOIrlT AND 33U'4DARY
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ASSIGt" lS FOR NESTINGHOUSE 2-LOOP I
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I XN-NF-77-25(A)
I 7.0 CONTAINMENT BACKPRESSURE CALCULATION The containment backpressure for the reflood period of the postulated LOCA was evaluated in accordance with the discussion presented in XN-75-41, Supplement 5, Section 4.6.
A containment analysis was performed using the computer code CONTEMPT-LT, Version 22 modified as described in Supplement 5, I
Revision 1, of XN-75-41. (I)
The containment analysis considered the equivalent double-ended cold leg split rupture using the mass and energy release frora the RELAP4-EM blow-down analyses. Table 7.1 summarizes some of the pertinent input data such as containment volume, initial pressure and temperature, heat sink dimen-sions and properties, and capacity and initiation times for safety features.
The condensing heat transfer coefficient is modeled in accordance with Branch Technical Position CSB 6-1, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation."(13) Sixteen passive heat sinks were modeled.
Paint and zinc coatings on both the concrete and galvanized steel were ig-nored, which is conservative.
The mass and energy from the blowdown analysis is input through the end-of-blowdown, then assumed zero for the remainder of the transient.
This is conservative as it neglects energy released during reflood to con-tainment which would result in higher containment pressure.
The predicted containment pressure history for the reflood period is shown in Figure 7.1.
I I
I
I XN-NF-77-25(Z I
The computer code CONTEMPT-LT, Version 22 has undergone two modifi-cations:
1) The Fortran coding has been made compatible with the Control Data Fortran Extended Compiler (FTN).
2) The RELAP4 Environmental Subroutines have been added which permit the input data to be supplied in a free format form.
I I
I I
I I
I I
I I
I I
I
I
. XN-NF-77-25(h)
I TABLE 7.1 DRY CONTAINMENT DATA Containment Physical and Thermal Parameters I
6 3 Net Free Volume 1.033 x 10 ft Outside Air Temperature
-10 F I
Initiation Time for:
Spray Flow 2.0 sec Fan Coolers 20.0 sec Containment Initial Conditions:
Temperature 90 F Pressure 14.7 psia I
Relative Humidity 100%
Containment Spray Water:
I T2mperature 37 F Flow Rate (Total, 2 pumps) 3600 gpm I
Fan Air Cooler Capacity (total 4 coolers)
Vapor Temperature ( F)
Capacity (Btu /hr) 6 165 2.00 x 10 7
210 2.85 x 10 I
7 230 4.00 x 10 7
260 5.90 x 10 7
295 8.00 x 10 Thermal Conductivity and Volumetric Heat Capacity Thermal Volumetric Conductivity HeatCagacity Btu /ft - F)
Materials (Btu /hr-ft-F)
Insulation 0.0208 2.0 Carbon Steel Liner Plate 28 58.8 Structural Concrete 0.9 32.9 I
I
I XN-NF-77-25(A)
TABLE 7.1 (Continued)
DRY CONTAINMENT DATA Containment Passive Heat Sinks SURFACE AREA DESCRIPTION MATERIAL THICKNESS FT2 I
1.
Insulated dome and wall insulation 1.25 in 36,181.
steel 0.375 in concrete 2.5 ft 2.
Uninsulated dome and wall concrete 2.5 ft 12,474.
steel 0.375 in 3.
Basement floor concrete 2.0 ft 7,955.
steel 0.375 in concrete 2.0 ft 4.
Sump walls concrete 5.0 ft 2,342.
I steel 0.375 in concrete 3.5 ft 5.
Sump floor concrete 2.0 ft 297.
I steel 0.375 in concrete 2.0 ft 6.
Refueling cavity walls steel 0.25 in 5,200.
(inside) concrete 2.5 ft I
7.
Refueling cavity floor steel 0.25 ft 1,200.
concrete 2.5 ft 8.
Refueling cavity walls I
(outside) concrete 2.5 ft 6,900.
9.
Steam generator compartment concrete 2.5 ft 14,900.
10.
Intermediate level floor concrete 0.5 ft 6,170.
I
11. Operating floor structures concrete 2.0 ft 9,162.
and operating floor 12.
Heavy steel beam and crane steel 1.5 in 9,174.
I structure 13.
Steel beam steel 1.0 in 5,016.
14.
Cylindrical supports and beam steel 0.5 in 8,586.
15.
Crane support columns steel 0.75 in 5,756.
16. Grating and stairs steel 0.125 in 7,000.
m W
W W
m m
m m
m m
m m
m m
m m
m m
e i
4 e
i 1
?
i i
WESTINGHOUSE 2-LOOP PWR 1.0 DECLS i
i i
i i
i i
l i
a i
i 2
i 1
^
5 m
o-
. o-wea$
we o.
i H O I
z
.c:-
w e
l z
e z
~
i y
o a a
3 l
l t
I t
I i
f I
d 100 200 300 400 500 800 700 800 Xz TIME (SEC) 3 e
N I
ro FIGURE 7.1 CONTAINMENT BACKPRESSURE FOR 9
DECLS (C = 1.0)
D i
l
I XN-NF-77-25(A)
I
8.0 REFERENCES
I 1.
Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model, XN-75-41:
I a.
Volume I, July 1975 b.
Volume II, August 1975 c.
Volume III, Revision 2, August 1973 d.
Supplement 1, August 1975 I
e.
Supplement 2, August 1975 f.
Supplement 3, August 1975 g.
Supplement 4, August 1975 I
h.
Supplement 5, Revision 5, October 1975 i.
Supplement 6, October 1975 j.
Supplement 7, November 1975 2.
Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-II, XN-76-27:
a.
July 1976 b.
Supplement 1, September 1976 c.
Supplement 2, November 1976 3.
U.S. Nuclear Regulatory Commission, WREM, Water Reactor Evaluation Model, Revision 1, May 1975.
4.
Lilly, G. P., Mixing of Emergency Core Cooling Water with Steam:
1/3 - Scale Test and Summary, EPRI 294-2 Final Report, June 1975.
5.
Broderick, J.
F.,
Loiselle, V., Cold Leg Condensation Tests.
Task C. Steam Water Interaction Tests, CENPD-129, March 1974.
6.
Broderick, J. R., Burchill, W. E., Lowe, P. A.,1/5 Scale Intact l
Loop Post-LOCA Steam Relief Tests, CENPD-63 Revision, March 1973.
l g 7.
Flanigan, L.
J., Cudnik, R.
A., Denning, R.
S., Topical Report
' W on Experimental Studies of ECC Delivery in A 1/15 - Scale Trans-parent Vessel Model, BMI-1941, November 1975.
8.
Rothe, P. H., Wallis, G. B., Thrall, D. E., Cold Leg ECC Flow Oscillations, EPRI NP-282, November 1976.
9.
Zender, S.
N., Jensen, M.
F.,
Sackett, K.
E., Experiment Data l
Report for. Semiscale MOD-1 Test S-01-4 and S-01.4A, ANCE-1196, March 1975.
I I
I
I
-bl-XN-NF-77-25
I
10. Antonopoulos, P. T., and Husain, A., Method.for Calculating End of Bypass Time for Yankee Rowe Loss of Coolant Accident Analysis, YAEC-1125, March 1977.
11.
Block, J.
A., and Wallis, G.
B., Effect of Hot Walls on Flow in a Simulated PWR Downcomer During a LOCA, CREARE-TN-188, May 1974.
12. Block, J.
A., and Crowley, C.
J., Hot Wall Experiments in Simulated Multiloop PWR Geometry, CREARE-TN-2DT, February 1975.
13.
U.S. Nuclear Regulatory Commission, Minimum Containment Pressure Model for PWR ECCS Performance Evaluation, Branch Technical Position CSB 6-1.
I lI I
I I
l:I I
lI I
I
i XN-NF-77-25(A)
I EXXON NUCLEAR COMPANY ECCS EVALUATION OF A 2-LOOP WESTINGHOUSE PWR WITH DRY CONTAINMENT USING THE ENC WREM-II ECCS MODEL LARGE BREAK EXAMPLE PROBLEM Distribution K. P. Galbraith J. E. Krajicek I
G. F. Owsley G. A. Sofer RG&E c/o L. J. Federico (5)
I USNRC c/o G. F. Owsley (70)
Document Control (10)
SE Jensen I
WV Kayser FJ Markowski WS Nechodom I
i I
I I
I I
I p
,

ML20148C148

Contents

Text

References:

1.0 INTRODUCTION

SUMMARY

8.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

8.0 REFERENCES

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ML20148C148

Text

References:

1.0 INTRODUCTION

SUMMARY

8.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

8.0 REFERENCES

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