ML20210T466

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Forwards Large Break LOCA Analysis,Per License Condition 2.C.12.Analysis Performed Using Revised Bart Evaluation Model & Recent Mods in WCAP-9561-P,Addendum 3.Analysis Demonstrates Compliance w/10CFR50.46.Fee Paid
ML20210T466
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 10/01/1986
From: Koester G
KANSAS GAS & ELECTRIC CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
KMLNRC-86-176, TAC-57361, NUDOCS 8610090030
Download: ML20210T466 (61)
v • d • e


Text

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __. _

l KANSAS GAS AND ELECTRIC COMPANY Yt-*E ELECTAC COMPANV October 1, 1986 GLENN L. MOESTER v<t persectasY - hucitan Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 KMLNRC 86-176 Re: Docket No. STN 50-482 Subj: Large Break LOCA Analysis

Dear Mr. Denton:

The purpose of this letter is to transmit the Large Break Loss-of-Coolant Accident (LOCA) Analysis for Wolf Creek Generating Station (WCGS), Unit No.

1. This analysis is required by the WCGS Facility Operating License NPF-42, License Condition 2.C.12. License Condition 2.C.12 requires submittal of the worst large break LOCA analysis for NRC review and approval using an approved ECCS evaluation model, prior to restart following the first refueling outage.

The LOCA analysis was performed using the revised BART Evaluation Model, which considers the effect of core thimbles, and other recent modifications as described in WCAP-9561-P, Addendum 3 Break sizes with the discharge coefficient, C , = 0.4, 0.6, and 0.8 were analysed with minimum safeguards safety injectkoninordertodeterminethelimitingbreaksize. The worst break size was determined to be the C = 0.4 double-ended cold leg guillotine break and resulted inaPeakCladTemperature(PCT)of2099.9 deg-F. The C assuming maximum safeguards safety injectkon=0.6breakwasalsoanalyzed to determine the sensitivity of minimum versus maximum safeguards safety injection flow. The minimum safeguards cases were determined to be limiting.

Present Technical Specification values including Fg (z) and the K (z) curve are bounded by this analysis. This analysis demonstrates conformance for WCGS with the 10 CFR 50.46 requirements for Large Break LOCA analyses. If you have any questions concerning this matter please contact me or Mr. O. L.

Maynard of my staff.

Very truly yours, -

8610090030 861001 /

PDR ADOCK 05000402 PDR

  1. +t 6 A P Glenn L. Koester Vice President - Nuclear GLK:see

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Attachment cc: P0'Connor '

JCummins j pccg [/& OD EJohnson 201 N. Market -Wictsta, Kansas - Mail Address: PO. Box 208 I Wichita, Kansas 67201 - Telephone: Area Code (316) 261-6451

STATE OP KANSAS )

) SS CITY OP WICHITA )

I, John A. Bailey, of lawful age, being first duly sworn upon oath, do depose, state and affirm that I am Director Engineering and Technical Services of Kansas Gas and Electric Company, Wichita, Kansas , that I have signed the foregoing letter of transmittal for Glenn L. Koester, Vice President - Nuclear of Kansas Gas and Electric Company, know the content thereof; and that all statements contained therein are true.

By b John A. Bailey /

Director Engideering and Technical Services SUBSCRIBED and sworn to before me this / de.y of , 1986.

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Expiration Date //27 tros

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Mr. H R. Denton October 1. 1986 Attachment to KMLNRC 86-178 LARGE BREAK LOCA ANALYSIS USING THE BART COMPUTER CODE I

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6.2.1.4.5 Additional Information Required for Confirmatory Analysis No additional information is deemed necessary for the perfor-i mance of confirmatory analyses.

i

'6.2.1.5 Minimum Containment Pressure Analysis for Perfor- s mance Capability Studies on Emergency Core Cooling System The containment backpressure used for the limiting case (CD=0.f) double-endedcoldlegguillotinebreakfortheECCS 'd[

analysis presented in Section 15.6.5 is presented in Figure l

6.2.1-86. The containment backpressure is calculated, using 2E the methods and assumptions described in Appendix A of Ref-erence 9. Input parameters, including the containment initial conditions, net free containment volume, passive heat sink materials, thicknesses, and surface areas, and starting time

' and number of containment cooling systems used in the analysis, are described in the following paragraphs.

6.2.1.5.1 Mass and Energy Release Data The mass and energy releases to the containment during the blowdown and reflood portions of the limiting break transient are presented in Tables 6.2.1-63 and 6.2.1-64.

The mathematical models which calculate the mass and energy releases to the containment are described in Section 15.6.5

. and conform to 10 CFR Part 50, Appendix K, "ECCS Evaluation Models." A break spectrum analysis is performed (see ref-erences in Section 15.6.5) that considers various break sizes,

, break locations, and Moody discharge coefficients for the double-ended cold leg quillotines which do not affect the mass and energy released to the containment. This effect is con- 7 sidered for each case analyzed. During refill, the mass and <

energy released to the containment is assumed to be zero, g

)

which minimizes the containment pressure. During reflood, the ,

effect of steam water mixing between the safety injection water and the steam flowing through the reactor coolant system intact loops reduces the available energy released to the containment vapor spaces and therefore tends to minimize containment pressure.

k 6.2.1-35 n

ev A N

-6NUPPS-cJe f f Creek 6.2.1.5.2 Initial Containment Internal Conditions The following initial values were used in the analysis:

a. A containment pressure of 14.7 psia.
b. A containment temperature of 90 F.

t 6.2.1-35a ,

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l 9NUPPS- Voif CresX i

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d.

A refueling water storage tank temperature of 37 F.

An outside temperature of -60 F.

l i e. A relative humidity. of 99 percent. k These containment initial conditions are representatively low values anticipated during normal full power operation.

6.2.1.5.3 Containment Volume The volume used in the analysis was 2.7 x 108 ft 8.

6.2.1.5.4 Active Heat Sinks The containment spray system and containment air coolers operate to remove heat from the containment. h' Pertinent data for these systems which were used in the analysis are presented in Table 6.2.1-65.

The sump temperature Was not used in the analysis because the i maximum peak cladding temperature occurs prior to initiation

'of the recirculation phase for the containment spray system.

In addition, heat transfer between the sump water and the

containment vapor space was not considered in the analysis.

6.2.1.5.5 Steam-Water Mixing Water spillage rates from the broken loop accumulator are determined as part of the core reflooding calculation and are included in the containment code (COCO) calculational model.

l 6.2.1.5.6 Passive Heat Sinks The passive heat sinks used in the analysis, with their thermo-

physical properties, are given in Table 6.2.1-66. The passive
heat sinks and thermophysical properties were derived in
compliance with Branch Technical Position CSB 6-1, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation."

6.2.1.5.7 Heat Transfer to Passive Heat Sinks The condensing heat transfer coefficients used for heat transfer to the steel containment structures are given in Figure 6.2.1-87

} for the limiting break. The containment pressure transient for the limiting break is shown in Figure 6.2.1-86.

l 6.2.1.5.8 Other Parameters No other parameters, including the operation of the contain-ment minipurge system, have a substantial effect on the minimum containment pressure analysis.

6.2.1-36

.", J 1-9/00 1

eavaF5 LUdf Cmd TABLE 6.2.1-63 MASS AND ENERGY RELEASE DURING BLOWDOWN.

FOR MINIMUM POST.-LOCA CONTAINMENT PRESSURE

  • TIME (SEC) MASS FLOW (LB/SEC)
  • ENERG[ FLOW (BTU /SEC) 0.00000 0.0000000E+00 .6000000E+00 0.05198 5.5471681E+04 3.089781?E+0F 0.10133 5.3182319E+04 2.9591798E+0F
0.15037 5.4812818E+04 0.20038 3.0508679E+0F 5.4747017t+ 04 3.0471567E+0F 0.25177 5.4753027E+04 3.0478936E+ 0F 0.30087 5.4567568E+04 0.35206 3.0377483E+07 5.4480113E+04 ,

3.0332608E+0F 0 4011F 5.4483882E+04 3.0340161E+0F 0.45049 5.4294221E+04 3.0241412E+ 0F 0.50071 5.4092992E+04 3.0134827E+0F 0.55112 5.3426158E+04 .:

0 60040 2.9FF4612E+0F 5.3675149E+04 2.9923747E+0F 0.45172 5.3212098E+04 2.9676127E+0F 0.70092 5.3132940E+04 2.9644813E*07 0.75019 5.2744251E+04 2.9440432E*07 0.80158 5.2780453E+04 2.9477634E*0F 0.85118 5.2530991E+04 2.9354424E+0F 0.90096 5.2288181E*04 2.9236754E+07 0.95069 5.2245975E+04 8 2.9232264E+0F 1.0006L -

' 5.170029FE+04 2.4944708E+0F

. 1 10176*. 5.0824792E+04 2.8491074E+07 1.20169 5.0075073E+04 2.8107821E+07 1.30057 4.9119764E+04 2.7606434E*0F I

1.40130 4.7975369E+04 '

2.6995433E+07 1.50173 4.7807406E+04 2.6941199E+07 1.60103 4.7231284E+04 2.6650928E*0F 1.70081 4.6584725E+04 2.6322652E+0F l 1.80095 4.5742696E*04 2.5880545E+07 i 1.90081 4.4998504E+04 2.5490844E+07 2.00059 4.3848510E+04 2.4862811E+07 l 2.10121 4.2419787E+04 2.4071096E*07

[ 2.20089 4.2111267E+04 2.3930334E+07 i 2.30223 4.0899440E+04 2.3261149E+07 i 2.40179 4.0380546E+04 2.2997010E+07 2.50084

  • 3.9910826E+04 2.2762017E+0F 2.60070 3.9416120E+04 2.2514466E+0F 2.70174 3.9052817E+04 2.2339092E+07 2.80150 3.8502569E+04 2.2057703E*07 2.90101 3.8015348E+04 2.1811596E+0F 3 00112 3.6712053E+0( 2.1100601E+0F .

3.10018 3.5656440E+04 2.0519039E+0F 3.20134 3.476715FE*04 2.0020339E+07 3.30170 3.3803254E+04 1.94FF510E+0F 3.40101 3.3024994E*04 1.9042628E+07 3.50181 3.2901339E+04 1.8983528E*07 3.40120 3.1770630E+04 1.8363673E+0F 3.70140 3.1460784E+04 1.820685FE+07 3.80227 3.1217553E+04 1.8088929E+0F 3.90137 3.038265FE+04 1.7638027E+07 4.00196 3.0042356E*04 1.7467426E+07 4.10231 2.9770775E+04 1.7336461E+07 4.20235 2.9367197E*04 1.713320FE*07 -

4.30165 2.9127611E+04 1.7025421E+0F 4.40170 2.8537560E*04 1.6723076E+0F 4.50008 2.8212997E+04 . 1.6572116E*07 4.40172 2.7839474E+04 1.6392899E+0F

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TABLE 6.2.1-63 (Sheet 2)

TIME (SEC) MASS FLOW (LB/SEC) . ENERGY FLOW (BTU /SEC) 4-4.70122 2.74173F3E+04 4.80111 2.7127097E+04 1.6187338E*07 4.90224 2.6730711E+04 1.6059038t+07 5.0002F 2.6189058E+04 1.5870287E+07 5.25126 2.4897842E+04 1.5599065E+07 5.50168 2.4500986E+04 1.4953615E+0?

3.75091 2.3117509E+04 -

1.4798809E+07 6.000?F 2.2287967E+04 1.4063434E+07 6.25098 2.2180918E+04 1.3624766E+07 6.50148 2.2572270E+04 1.3562947E*07 6.75061 2.2494979E+04 1.3774071E+07 -

F.00189 2.2263395E+04 1.3733525E+0F .

7.25173 2.1358049E*04 1.3621478E+0F

  • 7.50354 2.1068916t+04 1.3138851E*0F 7.75748 2.0736184E*04 1.3009015E*07 8.00503 2 0487904E+04 1.2862887E+0F 8.25640 2.0191603E*04 1.2759719E+0F 8.50922 1.9846557E+04 1.2640470E+07 8.75236 1.9420245E+04 ' 1.2507904E*07 9.00661 1.8911905E+04 1.2342759E+07

' 9.25271- '

1.8560017t+04 1.2134220E+0F

~ 9.50864' #

1.8311603E*04 1.1969198E+0F 9.75110 1.8280752E+04 1.1824228E+0?

10.00335 1.8127259E+04

, 1.1788212E*07 10.25557 1.7841968E+04 1.1690946E+0F 10.50143 1.7498146E+04 1.1571381E*07 10.75477 1.7154096E+04 1.1416162E+07 11.00174 1.6835157t+04 1.1257032E+0F 11.25456 1.6487347E*04 1.1101395E+07 11.50799 1.0934945E+0F 1.6073658t+04 1.0744983E*07 11.75185 1.5625563E*04 12.00606 1.5154172E+04 1.0542684E+07 92.25346 1.4719920E+04 1.0325946E*07 12.50804 1.4320361E+04 1.0114903E+07 l 12.75320 1.3965716E+04 9.9099033E*06 l 13.00699 9.7234253E+06

- 1.3611567E+04 9.5314936E+06 13.2552F 1.3295572E*04 13.50352 1.3021775E+04 9.3511070E+06 13.75701 9.1854583E+06 1.2758802E+04 9.0187561E+06 14.00277 1.2511215E+04

  • l 14.25290 8.8610237E+06 1 1.2267639E*04 8.7017153E+06 14.50633 1.2029078E*04 14.75292 1.1772510E+04 8.5461835E*06 15.00414 8.3815388E+06 1.1519331E+04 8.2185176E+06 15.25181 1.1313300E+04 15.50243 8.0720024E+06 1.1162995E+04 7.9509660E*06 15.75443 1.1041060E*04 l 16.00430 16.25391

% 1.0948689E+04 1.0897746E+04 7.8360011E+06 7.7367441E+06 16.50135 7.6565393E*06 1.0896614E+04 7.6024577E+06 16.75075 1.0847835E*04 17.00030 7.5325571E+06 1.0732635E*04 7.4279312E+06 1 7.25118 1.0592407E+04 17.50111 1.0399776E+04 7.3209811E+06 -

17.75027 7.194420?E+06 1.0143065E+04 '7.0433374E+06 18.00298 9.8723522E+03 18.25066 9.6570097E+03

'6.8842983E+06 18.50178 6.7434150E+06 9.4259227E+03 6.5890321E+06 18.75217 9.1202712E+03 19.00162 6.4127294E+06 8.8032475E*03 6.2338981E+06 19.25272 8.5975171E+03 19.50093 6.0915650E+06 8.5280632E*03 5.9991130t+06 19.75294 8.3641867E+03 20.00088 5.8814290E+06 8.0047030E+03 5.7056319E+06

. --- ._~ ,

C2", "-S- LGog CutA TABLE 6.2.1-63 (Sheet 3)

TIME (SEC) MASS FLOW (LB/SEC) . ENERGY FLOW (BTU /SEC) 20.25144 7.4370674E+03 20.50219 7.5378839E+03 5.5216255E+06 20.75082 7.2500689E*03 5.4304132E+06 21.00178 4.8671034E*03 5.2734818E+06 21.25082 4.5589053E+03 5.0968258E+06 21.50250 4.2767557E+03 4.9416451E*06 21.75129 4.0346490E+03 4.7912035E*06 22.00107 5.8246557E+03 4.6537124E+06 22.25070 5.6323622E+03 4.5255457E+06 22.50093 5.4772775E+03 4.4126039E+06 '

22.75098 5.3080005E+03 4.3147879E+ 06 23.00025 5.1407904E+03 4.2130852E+06 ."~

23.25064 5.0101564E+03 4.1116144E+06 23.50001 4.8465388E+03 4.0221117C+06 23.75150 4.6672789E+03 3.9234639E+06 24.00064 4.4545966E+03 3.8213058E+06 24,25086 3.6879070E+06 4.2498845E+03 24.50160 4.069450?E*03 3.5393754E+06

! ' 24.75125 3.9445860E+03 8 3.3898178E+06 125.00063 - -

4.0039609E+03 3.2329349E+06

.'. 25.25029  ! 3.9730381E*03 3.1293860E+06 25.50024 3.7212447E*03 3.0504008E+06 25.75072 3.4577684E+03 2.8970860E*06 26.00110 3.1852908t+03 ' 2.7384361E+06 26.25091 2.9431291E+03 2.5688543E+06 26.50169 3.3665568E+03 2.4150158E+06 26.75078 4.3836646E+03 2.4925442E+06 27.00052 4.9123266E+03 2.7737051E+06 27.25034 5.0817554E+03 2.9050786E+06 27.50099 5.2777059E+03 2.9201573E+06 27.75164 5.4301659E+03 2.9454353E+06 28.00099 5.5320037E*03 2.9523375E+06 28.25109 5.5975522E+03 2.9381545E+06 18.50137 5.6353159E*03 2.9090508E+06 i

28.75215 5.6512027E+03 2.8688634E+06 i 2.8200543E+06 29.00210 5.6478407E+03 l 29.25023 5.4293904E*03 2.7639809E+06 t

29.50101 5.5966430E+03 2.7022407t+06 l

29.75149 5.5474158E+03 2.6343909E+06 30.00164 . 5.4853462E+03 2.5607102E+06 30.25019 5.4077013E*03 2.4829958E+06 -

30.50113 5.3166917E+03 2.4001150E+06 30.75067 5.2111214E+03 2.3120893E+06 31.00133 5.0973920E+03 2.2216267E+06 31.25113 4.9703102E*03 2.1275869E+06 31.50097 4.8550772E+03 2.0319775E+06 31.75033 4.6732127E*03 1.9440761E*06 32.00069 4.8094215E+03 1.8345207E+06 32.25042 4.7569756E+03 1.8427693E*06 32.50027 4.5402999E*03 1.8094467E+06 32.75027 4.4993140E+03 1.7206751E+06 i

33.00022 4.4501772E*03 1.4766193E+06 1

33.25096 4.3001117E+03 1.6342281E+06 l

33.50114 4.1406660E+03 1.5635108E+06 33.75062 4.0418902E+03 1.4862247E+06 34.00066 3.9376645E+03 1.42462o1E*06 34.25090 3.7840416E+03 1.3643279E+06 l 1.2923804E+06 34.50099 3.7454215E+03 i 34.75021 3.4783349E+03 1.4209963E+06 1

' 35.00079 3.2944506E+03 1.1923785E+06 35.25033 3.0861869E*03 1.1035883E+06 35.50019 ,1.0102918E+06 2.9130371E+03 9.2434722E*05 35.75037 2.8079573E+03 36.00028 2.8446842E+03 8.5953626E+05 8.4270724E+05 l

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

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-SNUPPS~~ Lucy Cr.eek TABLE 6.2.1-63 (Sheet 4) l TIME (SEC) MASS FLOW (LB/SEC) ENERGY FLOW (BTU /SEC) 36.25075 2.8600540E+03 8.3022828E+05 36.50098 2.8322902E*03 8.0865418E*05 36.75098 2.7951990E+03 7.8520833E+05 37.00061 2.7668287E+03 7.6533479E+05 37.25039 4.4476145E*03 1.1524620E+06 37.50001 6.2547636E+03 1.5694592E+06 37.75062 3.7984953E+03 .

9.3433220E+05 38.00098 4.1584691E+03 1.0005845E+06 38.25095 1.4055922E+03 3.2699096E*05 38.50073 2.0597712E+02 2.7118549E+04 38.75073 4.6468470E+02 L.4076078E+04 1 39.00018 -1.2212861E+02 -1.4247354E+05 39.25106 -2.0157203E+02 -2.3515111E+05 39.48105 -2.3064814E+02 -2 4907091E*05

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-suuves uM$ Cwh TABLE 6.2.1-64 MASS AND ENERGY RELEASE DURING REFLOOD FOR MINIMUM POST-LOCA CONTAINMENT PRESSURE TIME 1n(Total) Inh (Total)

(sec) (lbm/sec) (BTU /sec) 40-66 5S*'io 7 0 0 it.?SSSd 7F3 34-+~I 0. 03 3 0.05:4 va. P 7 52-963 CG.2oC 295r46 Eb F(/ 1. 01 '., 46 M d 9'-

00.100 F6 305' 252.77 Cd.J/  ;,;;;3 g.027V .'

Siih4GB /or.cos 9:it4,49 F/ 26 1.00:', / /4 f.f ,

1Gh443 /.?9 cc 7 34L.33 2 y"J.7 7 h 99+5 /. '7/ ".6~

i 3-l 19h403 /.ro.6cs- 345,46.3 Co.c f . 1.00 : 0 / PS'*f 107.000 / 73./05 444,43 JC/.fo, h 69t5 /. # f 74~

Bihh443 6 h6t+5 389,993 010.01 1.22:5 396-993 072.07 0.22:1-  !

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TABLE 6.2.1-65  %

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ACTIVE IIEAT SINK DATA -)

FOR MINIMUM POST-LOCA CONTAINMENT PRESSURE (>

(

Containment Spray System Parameters )

( s Number of pumps operating 2 )

(

Runout flow rate (total), gpm 7754 )

C Temperature of spray, F 37 )

Actuation time (full flow), sec 25

(

)

Containment Air Cooler Parameters (

Number of fan coolers operating 4 ()

Actuation time, cec 35 0

m ,,

1 4/00-

f-

_GNUpps. Wolf Ceak 1

)

s TABLE 6.2.1-66

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STRUCTURAL HEAT SINKS Thickness (ft) Area (ft 2 )

.021 Carbon steel ,64,919 4.0 Concrete

.021 Carbon steel 34,129 3.0 Concrete 1.5 concrete 13,538

.021 Carbon steel 10.0 Concrete 1.0 Concrete 8,564 2.0 Concrete 43,497 2.5 Concrete 17,061 l

.021 Carbon steel 7,821 2.0 Concrete

(-

.021 Stainless steel 2.0 Concrete 8,708

.0001083 Zinc coating 8,081

.005 Carbon steel 2.0 Concrete

.0001083 Zinc Coating 186,183  :

.0104 Carbon steel l1

.0104 Carbon steel 17,746

.0208 Carbon steel 114,205

.0417 Carbon steel 49,101 I

.0833 Carbon steel 31,372 i

.1667 Carbon steel 5,631

.3333 Carbon steel 8,355 i

.6667 Carbon steel 503

.0833 Carbon steel 9,726 k .0104 Stainless steel 19,779

.0417 Stainless steel 10,885 Rcv. I dh4MP- D ll

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O1H SuisuspuoO Ilom gueuJujotuoo M LiolCCreek' FIGURE 6.2.187 CONDENSING HEAT TRANSFER COEFFICIENTS -DECLG (Co = 0.4)

e N - AIC Cre d 15.6.3.3.3.2 Doses to Receptor at the Exclusion Area Boundary and Low-Population Zone Outer Boundary (s The potential radiological consequences resulting from the oc-currence of a postulated SGTR have been conservatively analyzed, using assumptions and models described in previous sections.

The total-body dose due to immersion and the thyroid dose due to inhalation have been analyzed for the 0-2 hour period at the exclusion area boundary and for the duration of the accident (0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) at the low-population zone outer boundary. The results are listed in Table 15.6-5. The resultant doses are well within the guideline values of 10 CFR 100.

15.6.3.4 Conclusions A steam generator tube rupture will cause no subsequent damage to the RCS or the reactor core. An orderly recovery fr'om the accident can be completed, even assuming simultaneous loss of offsite power.

15.6.4 SPECTRUM OF BWR STEAM SYSTEM PIPING FAILURES OUTSIDE OF CONTAINMENT This section is not applicable to SNUPPS.

15.6.5 LOSS-OF-COOLANT ACCIDENTS RESULTING FROM A SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY

() -

~

15.6.5.1 Identification of Causes and Frequency Classification A LOCA is the result of a pipe rupture of the RCS pressure r boundary. For the analyses reported here, a major pipe break (large break) is defined as a rupture with a total cross-

! sectional area equal to or greater than 1.0 square foot (ft 2),

l This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the lifetime of the plant but is postulated as a conservative design' basis j (see Section 15.0.1).

l A minor pipe break (small break), as considered in this section, is defined as a rupture of the reactor' coolant pressure boundary

, ('ee s Section 5.2) with a total cross-sectional area less than

1.0 ft in which the normally operating charging system flow

. is not sufficient to sustain pressurizer level and pressure.

This is considered a Condition III event, in that it is an infrequent fault which may occur during the life of the plant.

(

15.6-16 t

-ONUPPS- (J l9 Cree k.

i l

The Acceptance Criteria for the LOCA is described in 10 CFR s ,

(- 50.46 as follows:

l l

a. The calculated peak fuel element clad temperature is below the requirement of 2,200 F.
b. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
c. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxidation limits of 17 percent are not exceeded during or after quenching.
d. The core remains amenable to cooling during and after the break.
e. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

These criteria were established to provide significant margin in the emergency core cooling system (ECCS) performance following a LOCA. Reference 2 includes a recent study of the probability of the occurrence of RCS pipe ruptures.

(

In all cases, small breaks (less than 1.0 ft2 ) yield results with more margin to the Acceptance Criteria limits than large breaks.

15.6.5.2 Sequence of Events and Systems Operations should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached. A safety injection signal is generated when the appropriate setpoint is reached. These countermeasures will limit the consequences of the accident in two ways:

a. Reactor trip and borated water injection complement the void formation in causing the rapid reduction of power to a residual level corresponding to fiscion product decay heat. However, no credit is taken in the LOCA analysis for the boron content of the in-jection water. In addition, the insertion of control rods to shut down the reactor is neglected in the large-break analysis.
b. Injection of borated water provides for heat transfer

( from the core and prevents excessive clad temperatures.

15.6-17 l

I

CN"PPC Rolf Creek g Description of Large-Break LOCA Transient The sequence of events following a large-break LOCA is presented in Figure 15.6-4.

Before the break occurs, the unit is assumed to be in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. During blowdown, heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling. There-after, the core heat transfer is based on local conditions with transition boiling and forced convection to steam as the major heat transfer mechanisms.

The heat transfer between the RCS and the secondary system may be in either direction, depending on the relative temperatures.

In the case of heat removal from the primary, secondary system pressure increases, and the main steam safety valves may actuate to limit the pressure. Makeup water to the secondary I side is automatically provided by the auxiliary feedwater system. The safety injection signal actuates a feedwater ,

isolation signal which isolates normal feedwater flow by closing the main feedwater isolation valves and also initiates  !

emergency feedwater flow by starting the auxiliary feedwater pumps. The secondary flow aids in the reduction of RCS ('

pressure.

  • When the RCS depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops. Since i the loss of offsite power is assumed, the reactor coolant pumps are assumed to trip at the inception of the accident.

The effects of pump coastdown are included in the blowdown analysis.

{

l The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2,250 psia) falls to a value approaching that of the containment atmosphere. Prior to or at the end of the blowdown, the mechanisms that are responsible for the by-passing of emergency core cooling water injected into the RCS '

are calculated not to be effective. At this time (called '

l end-o f-bypas s ) , refill of the reactor vessel lower plenum begins. Refill is complete when emergency core cooling water has filled the lower plenum of the reactor vessel, which is bounded by the bottom of the fuel rods (called bottom of core recovery time).

The reflood phase of the transient is defined as the time period lasting from the end-of-refill until the reactor vessel l has been filled with water to the extent that the core temper- f l ature rise has been terminated. From the later stage of \

blowdown and then the beginning-of-reflood, the safety injection accumulator tanks rapidly discharge borated cooling water into l

15.6-18 1

-SNUPPG psd Creek  ;

I the RCS, contributing to the filling of the reactor vessel

(-'

downcomer. The downcomer w'ater elevation head provides the driving force required for the reflooding of the reactor core.

The low head and high head safety injection pumps aid in the filling of the downcomer and, subsequently, supply water to maintain a full downcomer and complete the reflooding process.

Continued operation of the ECCS pumps supplies water during long-term cooling. Core temperatures have been reduced to long-term steady state levels associated with the dissipation of residual heat generation. After the water level of the refueling water storage tank reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching to the. cold leg recirculation phase of operation in which spilled borated water is drawn from the engineering

. safety features sumps by the low head safety injection (residual heat removal) pumps and returned to the RCS cold legs. The containment spray system continues to operate to further reduce containment pressure. Approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after j initiation of the LOCA, the ECCS is realigned to supply water to the RCS hot legs in order to control the boric acid con-centration in the reactor vessel.

Description of Small-Break LOCA Transient As contrasted with the large break, the blowdown phase of the small break occurs over a longer time period. Thus, for a

(. small-break LOCA there are only three characteristic stages, i.e., a gradual blowdown in which the decrease in water level is checked, core recovery, and long-term recirculation.

15.6.5.3 Core and System Performance 15.6.5.3.1 Mathematical Model The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50.

Large-Break LOCA Evaluation Model i The analysis of a large-break LOCA transient is divided into three phases: 1) blowdown, 2) refill, and 3) reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the containment, and the fuel and clad temperature transient of the hottest fuel rod in the core. Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.

93% n g 30 The description of the various aspects of the LOCA analysis

, , methodology is given in References 3? This document 3aescribeCf'

( the major phenomena modeled, the interfaEEs among the computer codes, and the features of the codes which ensure compliance 15.6-19

r

-SNttPPS' WIS Cre.s k with the Acceptance Criteria. The SATAN-VI, WREFLOOD, COCO, f and LOCTA-IV codes which are used in the LOCA analysis are (

described in detail in References 4 throuth 7; code modifications ;,y,pp g are specified in References 8 through 10.+ These codes are used to assess core heat transfer and to determine if the core remains amenable to cooling throughout and subsequent to the blowdown, refill, and reflood-phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS during blowdown, and the WREFLOOD computer code is used to calculate this transient during the refill and reflood phases g y ,y.4 g of the accident.W The COCO computer code is used to calculate the containment pressure transient during all three phases of the LOCA analysis. Similarly, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod during the three phases.

SATAN-VI is used to calculate the RCS pressure, enthalpy, density, and the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary and secondary systems as a function of time during the blowdown phase of the LOCA. SATAN-VI also calculates the accumulator water mass and internal pressure and the pipe break mass and energy flow rates that are assumed to be vented to the con-tainment during blowdown. At the end of the blowdown phase, these data are transferred to the WREFLOOD code. Also at the end-of-blowdown, the mass and energy release rates during blowdown are transferred to the COCO code for use in the ('

determination of the containment pressure response during this first phase of the LOCA. Additional SATAN-VI output data from the end-of-blowdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA-IV code.

With input from the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic.model to determine the core flooding rate (i.e., the rate at which coolant enters the bottom of the core), the coolant pressure and temperature, and the quench front height during the refill and reflood phases of the LOCA.

WREFLOOD also calculates the mass and energy flow addition to the containment through the break. Since the mass flow rate to the containment depends upon the core flooding rate and the local core pressure, which is a function of the containment backpressure, the WREFLOOD and COCO codes are interactively linked. t" I F L O O D i s a l a c l i n k e d t o t h e L O C T r.- I V c e d e i n t h a t ReAcc. d, thermal hydraulic parametcra frc= E'nEFLOOD are uced by LOCT?-IV kseWL D in its calculation of the fucl temperature. LOCTA-IV is used throughout the analysis of the LOCA transient to calculate the fuel clad temperature and metal-water reaction of the hottest rod in the core. ( imr4 E-The large-break analysis was performed with the February 1978 Replace 6u MA vercien of the evaluation model, "hich includec modificatione Nac4 c delineated in ncferencca 15 through 10. (

15.6-20

Insert A The BART code is described in References 29 and 30.

Insert B The BART computer code is used to calculate the fluid and heat transfer conditions in the core during reflood.

Insert C The large break analysis was performed with the approved December, 1981 version of the Evaluation Model (Reference 28), with the approved 1984 version of BART (Reference 29).

Insert D With input and boundary conditions from WREPLOOD, the mechanistic core heat transfer model in BART calculates the hydraulic and heat transfer conditions in the core during reflood.

Insert E A schematic representation of the computer code interfaces for large break calculations is shown in Figure 15.6-5.

_c::UFF5 - Wolf Cceek

( .

Thes analysis in this section was performed with the upper head fluid temperature equal to the reactor coolant system cold leg fluid temperature, achieved by increasing the upper head cooling flow (Ref. 23).

Small-Break LOCA Evaluation Model The WFLASH program used in the analysis of the small-break LOCA is an extension of the FLASH-4 code (Ref. 11) developed at the Westinghouse Bettis Atomic Power Laboratory. The WFLASH program permits a detailed spatial representation of the RCS.

The RCS is nodalized into volumes interconnected by flowpaths. +

The broken loop is modeled explicitly with the intact loops lumped into a second loop. The transient behavior of the ,

system is determined from the governing conservation equations of mass, energy, and momentum applied through the system. A .

detailed description of WFLASH is given in Reference 12.  ;

The use of WFLASH in the analysis involves, among other things, the representation of the reactor core as a heated control volume with the associated bubble rise model to permit a transient mixture height calculation. The multinode capability of the program enables an explicit and detailed spatial repre- ,

sentation of various system components. In particular, it

[. enables a proper calculation of the behavior of the cross-over leg during a loss-of-coolant transient.

Clad thermal analyses are performed with the LOCTA-IV code (Ref. 7 ), which uses the RCS pressure, fuel rod power history, steam flow past the uncovered part of the core, and mixture height history from the WFLASH hydraulic calculations as input.

Figure 15.6-44 presents the hot rod power shape utilized to perform the small-break analysis. This power shape was chosen because it provides a conservative distribution of power versus core height, and also local power is maximized in the upper }h regions of the reactor core (10' to 12'). This power shape is skewed to the top of the core with the peak local power occurring at the 10.5-foot core elevation.

This is limiting for the small-break analysis, because of the core uncovery process far small breaks. As the core uncovers, the cladding in the upper elevation of the core heats up and is sensitive to the local power at that elevation. The cladding temperatures in the lower elevation of the core, below the two-phase mixture height, remain low. The peak clad temperature occurs above 10 feet.

Schematic representations of the computer code interfaces are (s given in Figures 15.6-5 and 15.6-6.

15.6-21 -Eev 1

-W&

-GNUPP9 Wo rp Cceg g The small-break analysis was performed with the October 1975 -

version of the Westinghouse ECCS Evaluation Model (Ref. 7, 12, 13, and 14).

15.6.5.3.2 Input Parameters and Initial Conditions Table 15.6-9 lists important input parameters and initial conditions used in the analysis.

s The analysis presented in this section was performed with a reactor vessel upper head temperature equal to the RCS cold leg temperature. The effect of using the cold leg temperature in the reactor vessel upper head is described in Reference 23.

In addition, the analysis in this section utilized the upflow barrel-baffle methodology described in Reference 19.

The bases used to select the numerical values that are input parameters to the analysis have been conservatively determined from sensitivity studies (Ref. 20, 21, 22). In addition, the requirements of Appendix K regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis. The assumptions made a

pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs and include such items as the core-peaking factors, the containment pres-sure, and the performance of the ECCS. Decay heat generated throughout the transient is also conservatively calculated.

15.6.5.3.3 Results

{

Large-Break Results Based on the results of the LOCA sensitivity studies (Ref. 20, 21, and 22), the limiting large break was found to be the double-ended cold leg guillotine (DECLG) break. Therefore, only the DECLG break is considered in the large-break ECCS performance analysis. Calculations were performed for a range of Moody break discharge coefficients. The results of these calculations are summarized in Tables 15.6-10 and 15.6-11.

The worst break in the spectrum of break sizes analyzed was a p) discharge coefficient (Cn) f 0 4S This worst break was analyzed with a 12-seconu diesel generator start time in order /c>

to study the effect of a 2-second delay in the start of the diesel generator. Two cases are presented in Table 15.6-11 <

for the C = 0.6 DECLG break: the minimum safety injection <

case and She maximum safety injection case (Reference 27 g

)

methodology). -The =cximum ccfety injection cccc prczed te bc 2 the =cct limiting. -

The mass and energy release data for the break resulting in the highest calculated peak clad temperature are presented in Section 6.2.1.5.

(,

n e; iT 15.6-22 5/ 9

_ CNUrrs 03 f f Cece k

( Figures 15.6-7 through 15.6-30 presentsthe parameters of 3E

( principal interest from the large-break ECCS analyses. For all cases analyzed, transients of the following parameters are presented:

a'. Hot spot clad temperature

b. Coolant pressure in the reactor core ,
c. Water level in the core and downcomer during reflood
d. Core reflooding rate
e. Thermal power during blowdown The containment pressure transient resulting from a LOCA is presented in Section 6.2.1.5.

For the limiting break analyzed, the following additional transient parameters are presented:

a. Core flow during blowdown (inlet and outlet)
b. Core heat transfer coefficients
c. Hot spot fluid temperature

(' d. Mass released to containment during blowdown

e. Energy released to containment during blowdown
f. Fluid quality in the hot assembly during blowdown
g. Mass velocity during blowdown
h. Accumulator water flow rate during blowdown
i. Pumped safety injection water flow rate during reflood The maximum clad temperature calculated for a large break is L 800-2+74 F , which is less than the Acceptance Criteria limit of 1 2200 F of 10 CFR 50.46. The maximum local metal-water reac-W tion i#-6-le percent, which is well below the embrittlement 3 limit of 17 percent, as required by 10 CFR 50.46. The total core metal-water reaction is less than 0.3 percent for all breaks, compared with the 1-percent criterion of 10 CFR 50.46, and the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. As a result, the core temperature will continue to drop, and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.

7 >

(>

C.

t nc. 19 15.6-23 6 '04 -

l

~

SNUPPS- (JoIf Crea K Small-Break Results As noted previously, the calculated peak clad temperature

(

resulting from a small-break LOCA is less than that calculated for a large break. Based on the results of the LOCA sensi-tivity studies (Ref. 20) the limiting small break was found to be less than a 10-inch-diameter rupture of the RCS cold leg.

Therefore , a range of small-break analyses is presented which establishes the limiting break size. The results of these analyses are summarized in Tables 15.6-10 and 15.6-12.

Figures 15.6-31 through 15.6-44 present the principal parameters of interest for the small-break ECCS analyses. For all cases analyzed, the following transient parameters are presented:

-a. RCS pressure

b. Core mixture height

'c . Hot spot clad temperature For the limiting break analyzed, the following additional transient parameters are presented:

a. Core steam flow rate
b. Core heat transfer coefficient

{'

c. Hot spot fluid temperature The maximum calculated peak clad temperature for all small breaks analyzed is 1790 F. These results are well below all Acceptance Criteria limits of 10 CFR 50.46, and in all cases are not limiting when compared to the results presented for  ;

large breaks.

15.6.5.4 Radiological Consequences 15.6.5.4.1 Method of Analysis 15.6.5.4.1.1 Containment Leakage Contribution PHYSICAL MODEL - Following a postulated double-ended rupture of a reactor coolant pipe with subsequent blowdown, the ECCS limits the clad temperature to well below the melting point and ensures that the reactor core remains intact and in a coolable geometry, minimizing the release of fission products to the containment. However, to demonstrate that the oper-ation of a nuclear power plant does not represent any undue radiological hazard to the general public, a hypothetical accident involving a significant release of fission products to the containment is evaluated.

(

Me'z 9 ? l 15.6-24 4b4M'- d> l

C "JrF3 - WO (Js4 Cred.

15.6.5.4.3.2 Doses to a Receptor at the Exclusion Area ,

Boundary and Low Population Zone Outer Boundary The potential radiological consequences resulting from the occurrence of the postulated LOCA have been conservatively  :

analyzed, using assumptions and models described in previous j sections.

The total-body dose due to immersion and the thyroid dose due ,

to inhalation have been analyzed for the 0-2 hour dose at the (

exclusion area boundary and for the duration of the accident at the LPZ outer boundary. The results are listed in Table 15.6-5. The resultant doses are within the guideline values of 10 CFR 100.

15.6.5.4.3.3 Doses to Control Room Personnel Radiation doses to control room personnel following a pos-tulated LOCA are based on the ventilation, cavity dilution, and dose model discussed in Section 15A.3.

Contzol room personnel are subject to a total-body dose due to immersion and a thyroid dose due to inhalation. These doses have been analyzed, and are provided in Table 15.6-8. The resultant doses are within the limits established by GDC-19.

15.6.6 A NUMBER OF BWR TRANSIENTS This section is not applicable to SNUPPS.

15.

6.7 REFERENCES

1. Burnett, T. W. T., et al., "LOFTRAN Code Description,"

WCAP-7907-A, April 1984. {

2. " Reactor Safety Study - An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants," WASH-1400, NUREG-75/014, October 1975.
3. Bordelon, F. M., Massie, H. W., and Zordan, T. A.,

" Westinghouse Emergency Core Cooling System Evaluation Model - Summary," WCAP-8339, July 1974.

4. Bordelon, F. M., et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant," WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), June 1974.
5. Kelly, R. D., et al., " Calculational Model for Core Reflooding After a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Non- ,

Proprietary), June 1974.

/

Rc'/ . WC-A-15.6-30 0/05

(

GL"J 23 - WACrue k

6. Bordelon, F. M. and Murphy, E. T., " Containment Pressure

( Analysis Code (COCO)," WCAP-8327 (Proprietary) andNWCAP-8326 (Non-Proprietary), June 1974.

7. Bordelon, F. M., et al., "LOCTA-IV Program: Los s-o f-Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974.
8. Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Model - Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-Proprietary), April 1975.
9. " Westinghouse ECCS Evaluation Model - October 1975 Version,"

WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary),

November 1975.

10. Letter NS-CE-924, dated January 23, 1976, Eicheldinger C.

(Westinghouse) to Vassallo D. B. (NRC).

, 11. Porsching, T. A., Murphy, J. H., Redfield, J. A., and

' Davis, V. C., " FLASH-4, A Fully Implicit FORTRAN-IV Program for the Digital Simulation of Transients in a Reactor Plant," WAPD-TM-84, Bettis Atomic Power Laboratory,

, March 1969.

12. Esposito, V. J., Kesavan, K., and Maul, B. A., "WFLASH, A FORTRAN-IV Computer Program for Simulation of Transients 4

( in a Multi-Loop PWR," WCAP-8200, Revision 2 (Proprietary) and WCAP-8261, Revision 1 (Non-Proprietary), July 1974.

i

. 13. Skwarek, R. J., Johnson, W. J., and Meyer, P. E.,

" Westinghouse Emergency Core Cooling System Small Break 3$

October 1975 Model," WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary), April 1977.

14. Letter NS-CE-1672, dated February 10, 1978, Eicheldinger C.

(Westinghouse) to Stolz,J. F. (NRC).

Do.lo.4e 2

15. 3 McIly, n. C., Thc=pcen, C. M., et al., "Mcctinghcure E=crgency Ocre Cccling Sycter Evaluation "Odel for Analycing Large LOCA'c During Opcration eith One Locp Out-Of Service for Plantc Nithout Loop !cclation Valves,"

WCAP-0155, February 1979.

Delehe) 16.hbEicheldinger, C., "Wcatinghouac 5000 Zvaluation Mcdel, y February 1973 "crcion," WCAr-0220-r--A (Proprietary) and <

WCAP-9221-P-n (Non-Proprictary), Echruary 1070. [

Deleted )

17.'SLetter NS-TMA-1001, dated Noveaber 1, 1970, Andersen T. M. <

(Wcctinghcucc) tc Stclc,J. F. (NnC).

k 15.6-31 4,c v . 1

-7/80

-StfttPPt WoIC CreeR beldel - ^ "

d2ted Decc9er 11, 1978, ..ndcracn, T.

(18. -Letter NS ".^.-2014, ..

('eca tinghcusc ) to Tedcsco n. L. (PRC). (-

19. Johnson, W. J. and Thompson, C. M., " Westinghouse Emergency Core Cooling System Evaluation Model - Modified October 1975 Version," WCAP-9168 (Proprietary) and WCAP-9169 (Non-Proprietary), September 1977.
20. " Westinghouse ECCS Evaluation Model Sensitivity Studies,"

WCAP-8341 (Proprietary) and WCAP-8342 (Non-Proprietary),

July 1974.

21. Salvatori, R., " Westinghouse ECCS - Plan Sensitivity Studies, WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July 1974.
22. Johnson, W. J., Massie, H. W., and Thompson, C. M.,

" Westinghouse ECCS-Four Loop Plant (17x17) Sensitivity Studies," WCAP-8565-P-A (Proprietary) and WCAP-8566-A (Non-Proprietary), July 1975.

23. Letter NS-TMA-2030, dated February 12, 1979, Anderson, T. M.

(Westinghouse) to Denton, H. R. (NRC).

24. DiNunno, J. J., et al., " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, Division of Licensing and Regulation, AEC, Washington, D.C., 1962.
25. USNRC NUREG-0409, " Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident," by Postma A. K. and Tam P. S., dated January 1978.
26. Maximum Safety Injection Worst Single Failure, NS-EPR-2538, December 22, 1981, letter from E. P. Rahe of Westinghouse Electric Corporation to R. L. Tedesco, Assistant Director of Licensing, and T. P. Speis, Assistant Director for Reactor Safety of the U.S. NRC.
27. Letter NS-EPR-2538, dated December 22, 1981, Rahe, E. P. P(

(Westinghouse) to Tedesco, R. L. (NRC). y

28. " Westinghouse ECCS Evaluation Model,1981 Ver sion," WCAP-9220-P-A, Rev.1 (Proprietary), WCAP-9221-A, Rev.1 (Non-Proprietary), February 1982.
29. Young, M., et. al., "BART-1A: A Computer Code for the Best Estimate Analyzed Reficed Transients", WCAP-9561-P-A,1984 (Westinghouse Proprietary).
30. Chiou, J. ii. , et al. , "Models for PWR Reflood Calculations Using the BART Code", WCAP-10062.

C

.

15.6-32 -6f84-

i

-ci."Jrrc Wo t$ Cre<.K 2

1A TABLE 15.6-9

.s

(. INPUT PARAMETERS USED IN THE ECCS ANALYSIS Licensed core power *, bA Wt 3411 Peak linear power, includes 102 percent -12.SS 13.439 factor, kW/ft Total peaking factor, F -sh-EH! L 41 Axial peaking factor, F 3 -1.453 LS6I3 Power shape Large break Chopped cosine Small break See Figure 15.6-44 Fuel assembly array 17 x 17 Accumulator water volume, nominal, 850 fta/ accumulator

/

\ Accumulator gas pressure, minimum psia 600

'30 Safety injection pumped flow See Figures 15.6-20 and 15.6-43 Containment parameters See Section 6.2 Initial loop flow, lb/sec 9919- %ch .TI Vessel inlet temperature, F 555.0 559.4'1 Vessel outlet temperature, F -010.0 62T7. l 9 Reactor coolant pressure, psia 2250 Steam pressure, psia 11dr 9 84.07 Steam generator tube plugging level, % He M3

  • Two percent is added to this power to account for calorimetric error.

(

ncc. 1

-3/CC

N (Jslf Geel TABLE 15.6-10 -

s TIME SEQUENCE OF EVENTS FOR

(

LOSS-OF-COOLANT ACCIDENTS Accident Event Time (sec)

Large break LOCA

a. DECLG C =0.8 Start 0.0 D CM9 Reactor trip signal -& 82-Safety injection signal T Accumulator injection beginsi3.co -led End-o f-bypass a5. 531 2-Pump injection begins QS elT 0 End-of-blowdown 2 m i 2 h-O-Bottom of core recovery ~59.M l 7 i Accumulator empty fi. I A to 4-7 r9
b. DECLG C =0.6 Start 0.0 (Minimuk SI Reactor trip signal OYi -Gr83 -',

(

atith 12-- Safety injection signal 5 T2. 1 <

. -second dio el Accumulator injection beginstr.9 lh4 4

-ge".erator End-o f-bypass 2.9.482-26 >

-start) Pump injection begins 21 11 -20: (.

End-of-blowdown 29.412 29.1 Bottom of core recovery M . 495" 4 0 . ^ -  ?, ..

Accumulator empty f 4.24f 2- P,

c. DECLG C =0.6 Start 0.0 C' (Maximuk SI Reactor trip signal o M'l4 -0.03- <

'-f i th 12 - Safety injection signal L .31 1.15-

-second dicact Accumulator injection begins 1r.4 &9- $

generator-- End-o f-bypas s 29.172.,24-4 <

start) Fump injection begins 1932 h >

End-of-blowdown 29.4%1 -2H (,

Bottom of core recovery 42.nto +O-fr (,

Accumulator empty cr. y t,

7 2 r Local Zr/H2O reaction, max. (%) 2 2 o M8 g,7 p efrr38 6,_1g g.73 g,35 y g py j Local Zr/H2 O location, ft G. g on;MS c,, o e +r!> 7,29  ?.5 ,; G,o Total Zr/H 2 O reaction, %. <0.3 <0.3 <0.3 <0.3 Hot rod burst time, aec SR.o439:2 yg, f47 8 37 ,9 yy.c y/A r Gr. p Hot rod burst location, ft 6.0 6.0 6.0 N/A = O 1 j ...

  • Refer to Section 15.6.5.3.3 and Table 15.6-10 for_ a definition of cases a through d. 4 I

i j sW

.- O .

a L BREAK OCCURS REACTOR TRIP ICOMPENSATED PRESSURIZER PRESSUREI g PUMPEO SAFETY INJECTION SIGNAL iMI1 CONT. PRESS. 04 LO PRESSURIZER PRESS I h PUMPED SAFETY INJECTION SEGINS (ASSUMING OFFSITE POWER AVAILASLEl W { $ ACCUMULATOR INJECTION CONTAINMENT NEAT REMOVAL SYSTEM INITIATION (ASSUMING OFFSITE POWER AVAILASLEI l U END OF SYPASS 1 r R END OF SLOWOOWN E y I PUMPED SAFETY INJECTION SEGINS (ASSUMING LOSS OF OFFSITE POWER) L * ) L 1r SOTTOM OF CORE RECOVERY - 1 'l CONTAINMENT HEAT REMOVAL SYSTEM INITIATION (ASSUMING LOSS OF OFFSITE POWERI R - E ACCUMULATORS EMPTY F L 0 -O .n - ~ CORE OUENCHED 'T > L o g SWITCH TO COLO LEG RECIRCULATION ON RWST LOW LEVEL ALARM (SEMI AUTOM ATICI G T. E ' R .% M 4 % j C SWITCH TO LONG TERM RECIRCULATION IMANUAL ACTIONI O O

  • L l 1 N

l 0

l 1

l M woc /L G s;A y FIGURE 15.64 SEQUENCE OF EVENTS FOR LARGE

BREAK LOCA ANALYSIS l

t i l p..... I ---.....)  ; I l \ 1 a l a 8 I i i l 1 l _ 15 t l l _ = E l i b,. _ _ _ _ _l .i _ _ _ _ _ _ _ , 8 h lm , I I 1I -le l , ll 18 --o- I . i o a d I , a g I . m I 3 I n b *I lI I 8 d I E b l g :- m 4 i i 3 8 e I I E ....___.J %___7_ I l- .g ~ l l - lI. gas I o g l

8 a n l J .

l QI = i  ! n-; ,g, l 4 a ~ E -} -- i Eul ' . 3 u. II -SNUROS. LllsM Cru.& l FIGURE 15.64 l CODE INTERFACE DESCRIPTION FOR LARGE BREAK MODEL l l l l l l 3 ) i W . L 1 F

  • L , O l A C .

S T H A CORE PRESSURE, CORE FLOW, MIXTURE LEVL'., AND FUEL ROD POWER . HISTORY - O < TIME < CORE COVERED 1 F i l l I N @lA Cv:u4 FIGURE 15A4 CODE INTERFACE DESCRIPTION FOR SMALL BREAK MODEL l _-.__e...,- - - - . - - . _ , y--- .--- 1 l l i LEGEND: LINE MARKED BY AN

  • IS AT THE PEAK NODE ELEVATION (7. 25 f t FOR ALL CASES)

UNMARKED LINE IS AT THE BURST NODE ELEVATION (6.00 ft FOR ALL CASES) l IC / . . ~ ./ / ,

  • l

= a - f "5m -

  • i 1 / ,

.W ~~ . i ~/ , < ~ \( ~ g k u s33ao3ai con sow . sos o^v N V l -r.N"J7 F,-WP$ LYa k l FIGURE 15.6-7 PEAK CLAD TEMPER ATURE s - DECLG (Cp = 0.6-MIN SI/ .E /  : R ./ ~ - E

  • / /  !
  • / /

./' /  : ( 'e j / RE' ~~

.< \ f

$( -

  • I .. _

%\ ' " I i IJ 53 m 30 5 i  ! com ion mai s u our l N LwQ- Cwh FIGURE 15.6 7a PEAK CLAD TEMPERNTURE DECLG (CD = 0.6- MAX SI) 9 O = 2 g  ;' N G N e-

iF
  • =

2 ~ l , / a E . x e - g . i m., w,, 4M FIGURE 15.64 COREpngssygg_ I DECLG (Co = 0.8) -..___..._--..,-_,..____-..-_,m...____---. _ _ . _ ___.__ _____. 3 I E I 3 I [Gm, w - td g y m n ..[ ; E o ~~ ? 3 (N , y . I . g W ' E - . __ g R $*$*E IAJ113A31 331WA

  • I
  • M LWEO-CYak FIGURE 15.6-9 DECLG (Co = 0.8 MIN St DOWNCOMER AND CORE WATER LEVELS DURING REFLOOD

.__.h.a- - . - __ .m. .. -. _ _ = 1 1

e. .

k l1* i l If M i i i 17.5 i. / b memm 4 DD j h 12 .s Y e se i w I ,,, ,r- C012 E O g i o r 1 <2 0 m # O 2 *5 4 mO m r- [

  • y a 3 j O g ,8 @ 's se Ise ise aos ase ses sse see ese see sse zy 3 tint estes m

EZo gj N P $ l 3O I P ! mO e

mm E e n C" k $

!; E! M m r . l 3 I

t i

t1 l *. t i i 3 I 3 B 3 G [w l lW i i  : I > $ 3 _E l : ., I i 4 " - ' 2 - ~ 13X/Wil3Aug 000 b l -SNtkPPS--- LAM // Cr2/A FIGURE 15.8-10 l 4 DECLG(Co = 0.6 MIN SI) { CORE INLET , VELOCITY DURING REFLOOD l . I 3 R I . E  : ieE l E! > k 2 / _E 3 4 3 s' , 83M/ Nil 3JWII 0001J - SNswFe-WdF CrEn FIGURE 15.610e l ' DECLG (CD = 0.8- MAX SI) CORE INLET VELOCITY DURING REFLOOD i IC 2 2 S EU V

's i 2

= a- < a a . .. .. . j h3AOd

  • M (64 1 C T44/i FIGURE 15.611 CORE POWER TRANSlENT-DECLG (Co = OA)

b " e I h . / >'- / , i * ./ E ~/ f ~ n .I / s .I / / se - . /

3r
  • \ / r i  ;  ; E
  • \ -

.\ / i A \ s , A , i I '"==> I . . a ,, ,, a , . h@/g Cy.,2e FIGURE 1E21g PEAK CLAD TEMPERATURE ~ DECLG (Co = OA) O e K 2 R O / j q - / 3 = - l - / , j J I IWIsol I. . I. s 3mnss3md i e O FIGURE 15.613 CORE PRESSURE - i DECLG (CD=0.8) ( 3 .E I 3 = i Eo $c "W - 9 ut 2 E I;g 2 a e i

: a s i g

+' N3 (, , e 3 . 3 =  ; e  ; e.3 . 3 . iunun ni. . N Mt/gCmd FIGURE 15.8-14 j i DECLG (CD= 0.8) DOWNCOMER AND CORE WATER LEVELS DURING REFLOOD 1 i e e . 3 I a. ta "W l I;- I = E_ 4 3 l 1 . p ' i m unmumaoou e M LUJY LYMk FIGURE 15A 15 1 1 DECLG (Co = 0.8) CORE INLET VELOCITY DURING REFLOOD I d 2 R ,G M. O 9  % e ,. 9 ['d/d)** m wdt-cwa. FIGURE 15A 16 CORE POWER TRANSIENT - DECLG (Co = OA) n . ) .. a w .1/ f / 5

  • I /

./ / ./ / ( 5 's .' / sx ( .s I' Q " (N . g } = A> . i I i IJ 53350308 !  ! a 00m a m e j w m y i l CMTG~WJYCwA FIGURE'15.817 PEAK CLAD TEMPERATURE - DECLG (CD = 0.4) l 9 9 I O R / / n-i - E P . , 2  : ~ / r = . B B B g = 2 R e _I = , swisde 3anssand I - t -SNU998 UL:i/ Cr uA FIGURE 15A 18 CORE PRESSURE - DECLG (Co = 0.4) _ . . . _ ~ _ _ _ _ _ . . . _ .__ _ ._ _ , _ - . . _ - . _ _ . S .E

  • E S

M G . '? . . . g c - E-E "' W Q s 8 = 5 s~ . p) n \ \ ' s N = 5 e  : 2  ; . - = . 11J113A31 333gn = sa y cr a 1 FIGURE 15.819 { DECLG (Co = 0.4) DOWNCOMER AND CORE WATER LEVELS DURING REFLOOD I . _ _ _ -._ _

  • 8

\ S e .E E, a 5 W E- "W e . .. } & n 9 s { \ E = e  :. l i FIGURE 15A40 ) i DECLG (Co = 0.4) CORE INLET VELOCITY DURING REFLOOD 1 e G I 1 \ i I 2 1 l C N 2-V S G  % M (%'en um e MiUJYLYsab ^ FIGURE 15A 21 CORE POWER TRANSIENT - DECLG (CD = 0.4) 't S a  : 8_ E z . W 4

  • Q

.a s . N - E 1 ^ a ~E , II!I taasism I!Ii menw e s.z M Uhl/ CY % . FIGURE 15A-22 CORE FLOW -TOP AND BOTTOM - DECLG (CD = 0.4 3 ~ 4 J L a

L l "

1 1

3

- )

l 0

g l

g- -

1

-g c

~

a I

gg

%==..

h, , a B W

qF jI

  • - m: dL _,..

_] ,,,, #

d 1 i E 3

...u m miai.u- - =. ~

O m- w,2f Crmt FIGURE 15.6-23 CORE HEAT TRANSFER COEFFICIENT -

DECLG (Co = 0.4)

R e/ ..

e

,, J g

/ ~

. / B

~

/

-l ,

=

/ -

. t .-  :

I

/ b

/

~

d E i

$ b  :

h 3 e >

x g D

V R b .

tJ $33WDEI 3trummas 01fild ~

FIGURE 15.6-24 FLUID TEMPERATURE -

DECLG (Co = 0.4)

G gY I v l l 3

. ~

8

/

)

f

. /

/

?  ?  ?  ?  ? .  ?"

~

j FIGURE 15.6-25 BREAK MASS FLOW RATE -

DECLG (Co = 0.4)

---y,.--.-,. . -- , , . - - - . . . - - , ,- -

, w ,, -- - , ,

t

)

, c

$ i

=

c G

s

,e  :

- W

~

/ ,

l. - ~ )

/ ,

/

/

/

e . e-5 4

155 4n5

- ,1 1, ,1 P y g )ADa3N3 m 3ae p.

FIGURE 15.6-26 BREAK ENERGY RELEASE RATE -

DECLG (Co = 0.4)

e I

.An W -

i

. .c -

e A

e -

o  !

x- -

m, , .M  :

e-JJ

. . . 7 >

. ~

,)

1

. - ~ j

/

\m  %'- l x

x N-1 I

i l

l l

7 a 3 .

~ ~

l 1180 343dl Oln1J JO A11'tW10 .

1 i

?= w w - WC UY Y FIGURE 15A47 FLUID QUALITY -

DECLG (Co = 04)

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

b B

  • I  :

~/

=

IC -

G W

2-W

. . 3

.. - ~

. - ~

  • l I

. E.

. .I so m sli nou wuou

.E l .I.

l t

l -stet:teps ab4 e r aA i

l FIGURE 15A-28 ACCUMULATOR FLOW DURING BLOWDOWN - DECLG (Co = 0.4)

1 l

1 1

O

  • \

i n 1 )

I i e  !

  • rw

?

G

. N ~

F G

.e .= -

C e.

a. -a } I b w i

/ -f r r som ewen u nao w ssow -

1 M aja t/- C t a k FIGURE 15.849 MASS VELOCITY -

DECLG (Cp = 0.4)

8 11 t 7 - .

H 6

7 5 s

j h

v 4- /

o

)

E 3 2

mm '

.Sh,

,1 Om 'T 00 m ,

Om o **

  • iQcM h O O ta M . e ,

~

20 40 '

E 60 80 100 120

$ 140 160 180 oj h 9 f me (sec) 80 g Oc e2 Oy bO Q

  • 1 D

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______

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