ML19289C643
| ML19289C643 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 12/27/1978 |
| From: | Crane P PACIFIC GAS & ELECTRIC CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7901220047 | |
| Download: ML19289C643 (36) | |
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.1 q-s PACIFIC G-A S AND E LE C T RI C C C M PANY 2 G / 5 77 BEALE STRCCT. 31ST F iO R r, A N FR ANCiSCO, C ALIFO RNI A 94106 (415) 781 4211 d
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December 27, 1978
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[ P' cy Mr. John F. Stolz, Chief Light Water Reactore Branch No. 1 h
Division of Project Management E/
4{$'hg% g7 h(*
S U. S. Nucicar Regulatory Commincion k;;
ll7 4
Washington, D. C. 20555 6
Re Docket No. 50-275-OL
/
6 Docket No. 50.323--OL D
Diablo Canyon Units 1 and 2 m
Dear Mr. Stolz:
Enclosed are 30 copics of Westinghouse proprietary material and 30 copies of the non-proprietary version of re-sponses to Questions 220.1 through 220.6 from your letter of July 25, 1978.
Also enclosed are an Application for Withhold-ing Proprietary Information Frcn1 Public Disclosure and support-ing letters from Westinghouse.
Kindly acknowledge receipt of the above material on the enclosed copy of this letter and return it to me in che enclosed addressed envelope.
Very truly yours, Philip A. Crane, Jr.
Enclosures CC w/non-proprietary enc.:
Service List 79012200 %
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TABLE 3 (Continued)
MASS AND ENERGY RELEASES II. Peak Temperature Case 0.908 ft2 Split Ruoture at 707. Power with Failure of a Containment Safeguards Train II.A.
Main Feedwater Flow transient - see Figure 1.B.
11.8.
Auxiliary Feed Flow - 1890 gpm II.C.
Steam Mass in Unisolated Steam Lines - 1878 lbm II.D.
Water Mass in Unisolated Feed Lines - 10,175 lbm II.E.
Initial Steam Generator Mass - 129,780 lbm (include item II.C.)
II.F.
Mass / Energy Releases (Dry Steam)
Time (sec)
Mass Flow (lbm/sec)
Eneroy Flow (M8tu/sec) 0.0 1655.6 1.981 1.0 1631.6 1.953 2.0 1610.0 1.927 3.0 1589.8 1.904 4.0 1571.3 1.882 5.0 1554.4 1.862 6.0 1538.9 1.844 7.0 1524.6 1.827 8.0 1511.5 1.812 9.0 1499.4 1.797 10.0 1488.2
.784 12.0 1468.5 1.761 14.0 1451.9 1.741 16.0 1438.1 1.725 18.0 1427.1 1.712
TABLE 3 (Continued)
MASS AND ENERGY RELEASES Time (sec)
Mass Flow (lb/sec)
Eneroy Flow (MBtu/sec) 18.0 1746.7 2.096 20.0 1651.6 1.981 25.0 1488.4 1.783 30.0 1383.4 1.656 35.0 1310.2 1.567 40.0 1256.5 1.502 45.0 1216.1 1.453 50.0 1184.8 1.415 60.0 1140.0 1.361 70.0 1062.3 1.267 80.0 1006.3 1.199 90.0 966.2 1.151 100.0 933.8 1.112 110.0 907.7 1.080 120.0 882.9 1.050 130.0 860.3 1.023 140.0 839.2 0.997 150.0 818.8 0.973 175.0 759.7 0.902 200.0 711.2 0.843 225.0 660.1 0.782 250.0 612.8 0.725 300.0 531.2 0.627 350.0 466.8 0.550 400.0 414.0 0.486 450.0 371.7 0.436 500.0 343.7 0.402 550.0 326.0 0.381 600.0 308.9 0.360 700.0 140.9 0.160
TABLE 3 MASS AND ENERGY RELEASES I.
Peak Pressure Case:
3.69 ft2 DER at 70% Power with Failure of a Main Feed Reculator Valve I.A.
Main Feedwater Flow transient - see Figure 1.A (curve 2)
I.B.
Auxiliary Feed Flow - 1890 gpm I.C.
Steam Mass in Unisolated Steam Lines - 1878 lb, I.D.
Water Mass in Unisolated Feed Lines - 10,175 lbm I.E.
Initial Steam Generator Mass - 127,000 lb, I.F.
Mass / Energy Releases Time (sec)
Mass Flow (lb/sec)
Eneray Flow 'MBtu/sec) 0.0 14,158.0 16.939 0.75 13,835.9 16.558 0.76 8890.9 10.645 1.0 8540.9 10.246 2.0 7448.1 8.948 3.0 6646.7 7.990 4.0 6058.1 7.284 5.0 5649.6 6.792 6.0 5370.9 6.455 7.0 5158.0 6.198 8.0 4984.8 5.994 9.0 4850.5 5.825 9.5 2456.1 2.955 10.0 2396.0 2.883 12.0 2186.4 2.629 14.0 2011.8 2.427 16.0 1865.9 2.241
TABLE 2 (Continued)
DIABLO CANYON CONTA!hMENT RESULTS Double-Ended Split Ruptures Time of Hi-1 Time of Hi-2 Time rf Break Pcner Reactor Trip Containment Contair. ment Time of FL Time of SL Peak Peak Peak Area Level Single and Time Pressure Pressure Isolation isolation Containcent Pressure Pressure Temp.
(ft )
(1)
Failure (sec)
(sec)
(sec)
(sec)
(s,-)
Model (psig)
(sec)
( F) 1.4 0
1 N/A
<2
<150 10.0 8.5 W
35.3 756 253.1 1.4 0
2 N/A e2
- 210 62.3 8.5 W
30.9 484 253.1 1.4 0
3 N/A e2 el60 10.0 8.5 W
30.5 430 253.1 1.4 0
4 N/A el e30 10.0 8.5 W
35.5 412 294.5 3.69 0
1 N/A
<2 e105 10.0 10.0 W
38.4 636 257.6 3.69 0
2 N/A
<2
- 130 62.3 10.0 W
38.9 642 258.9 3.69 0
3 N/A
<2
<107 10.0 10.0 W
33.8 551 250.1 3.69 0
4 N/A
<2 e38 10.0 10.0 W
37.7 337 281.3 3.69 70 2
HFLP 0 1.5
<3 41 62 9.0 W
41.8 515.8 P53.3 3.69 70 2
HFPL 9 1.5 e3 e67 62 9.0 W
41.8 581.8 282.8 Notes:
1.
Single Failures 1 = Failure of a containment safeguards train 2 = Failure of a feedline isolation valve 3 = Failure of the auxiliary feedwater runout protection 4 = Failure of a main steam isolation valve l
II. Trip Syntols HNF High Nuclear Flux Trip
=
j HFLP = High Steam Flow / Low Steamline Pressure Trip High 1 Cortainment Pressure Trip 1
=
i l
!!!. Containment Models I
W
= Westinghouse COCO Model, Dry Steam Blowdowns N
= NRC Sm il Break Model, Dry Steam Blowdowns j
WW
= Westinghouse C0C0 Model, Wet Blowdowns i
NW = NRC Hodel, Wet Blowdowns We,.,
.-w,,.v,.--
-x
~~
~
TABLE 2 (Continued)
OIABLO CANYON CONTAINMENT RESULTS Oouble-Ended Split Ruptures Time of Hi-1 Time of Hi-2 Time of Break Power Reactor Trip Containment Containment Time of FL Time of St Peak Peak Peak Area Level Single and Time Pressure Pressure Isolation Isolation Contaltnent Pressure Pressure Temp.
2 (ft )
(%)
Failure (sec)
(sec)
(sec)
(sec)
(sec)
Model (psig)
(sec)
( F) 3.69 102 1
HFLP 0 1.5 e2 ell 3 10.0 9.0 W
35.5 608.5 252.8 3.69 102 2
ffLP 0 1.5 e2
<82 62.0 9.0 W
44.3 568.8 267.2 3.69 102 3
HFLP 0 1.5 e2
- 114 10.0 9.0 W
32.1 605.5 246.9 3.69 102 4
HFLP 9 1.5
- 2 e56 10.0 9.0 W
33.8 604.3 270.4 1.4 70 1
HFLP 0 0.5
- 2 el45 10.0 8.0 W
38.2 649.9 257.2 1.4 70 2
HFLP 9 0.5 e2
- 140 62.0 8.0 W
38.2 641.8 257.7 1.4 70 3
HFLP 0 0.5
- 2
- 150 10.0 8.0 W
33.9 612.1 250.3 1.4 70 4
HFLP 9 0.5
- 2
- 57 10.0 8.0 W
36.9 519.4 236.8 3.69 70 1
HFLP 9 1.5 e3 ell 2 10.0 9.0 W
36.1 643.2 254.1 3.69 70 2
HFLP 9 1.5
- 3
- 89 62.0 9.0 W
44.9 607.5 268.1 3.69 70 3
HFLP 0 1.5 e3 ell 3 10.0 9.0 W
32.6 606.0 247.8 3.69 70 4
IFLP 9 1.5 e2 e52 10.0 9.0 W
34.6 600.4 273.2 1.4 30 1
HFLP 0 1.0
<2 el40 10.0 8.5 W
38.1 712.9 256.9 1.4 30 2
HFLP 9 1.0
- 2
- 155 62.5 8.5 W
35.0 617.6 252.3 1.4 30 3
HFLP 9 1.0
- 2
- 145 10.0 8.5 W
32.9 603.1 250.6.
1.4 30 4
ffLP 9 1.0 el e40 10.0 8.5 W
37.2 480.6 292.5 3.69 30 1
HFLP 9 2.0
- 2
- 108 10.0 9.5 W
37.6 618.0 256.6 3.69 30 2
HFLP 0 2.0
- 2
- 101 62.5 9.5 W
43.1 623.5 265.3 3.69 30 3
HFLP 0 2.0
- 2
- 110 10.0 9.5 W
33.7 607.2 250.0 3.69 30 4
HFLP 9 2.0
- 2 e44 10.0 9.5 W
36.6 291.2 278.3
d (AOLL 4 DIABLO CANYON CONTAINMENT RESULTS Split Ruptures Time of Time of Hi-1 Time of Hi-2 Break Power Reactor Trip Containment Containment Time of FL Time of SL Peak Peak Pe k Area Level Singis and Time Pressure Pressure Isolation ~ 1 solation Containment Pressure Pressure Temp.
(f t )
(%)
Failure (sec)
(sec)
(sec)
(sec)
(sec)
Model (ps19)
' sac)
{ F) 2 0.86 102 1
HNF 9 13.95 17.0 92.0 27.0 98.0 W
41.7 n 9.6 331.3 0.86 102 2
HNF 9 13.95 17.0 99.5 82.0 105.5 W
36.9 789.5 321.1 0.86 102 3
HNF 9 13.95 17.0 94.0 27.0 100.0 W
35.4 662.8 323.8 0.86 102 4
HNF 9 14.25 17.0 94.0 27.0 100.0 W
35.5 659.3 323.7 0.91 70 1
CPI 9 15.5 15.5 88.5 25.5 94.5 W
40.6 829.6 333.1 0.91 70 2
CP1 9 15.5 15.5 98.0 80.5 104.0 W
34.3 695.2 321.
0.91 70 3
CPI 9 15.5 15.5 91.0 25.5 97.0 W
34.0 633.3 325.7 0.91 70 4
CPI 9 15.5 15.5 90.5 25.5 96.5 W
34.1 632.4 325.9 0.94 30 1
CP1 9 14.0 14.0 89.5 24.0 95.5 W
37.7 871.5 331.8 0.94 30 2
CP1 9 14.0 14.0 103.0 79.0 109.0 W
30.7 550.1 320.6 0.94 30 3
CPI @ 14.0,
14.0 92.5 24.0 98.5 W
31.1 549.0 324.6 0.94 30 4
CPI 9 14.0 14.0 91.5 24.0 17.5 W
31.6 550.5 324.9 0.96 0
1 N/A 13.0 91.0 23.0 97.0 W
32.9 805.6 329.4 0.96 0
2 N/A 13.0 107.0 78.0 113.0 W
28.2 158.9 319.5 0.96 0
3 N/A 13.0 94.5 23.0 100.5 W
29.3 157.4 323.0 0.96 0
4 N/A 13.0 94.0 23.0 100.0 W
29.7 156.5 323.3 0.91 70 1
CP1 9 15.5 15.5 94.0 25.5 100.0 N
41.0 855.6 343.8 1.4 102 1
HFLP e 0.5 s4 el45 10.0 8.0 W
37.8 630.5 256.7 1.4 102 2
HfLP e 0.5
<2
<130 62.0 8.0 W
40.1 636.4 260.7 1.4 102 3
HFLP 0 0.5 e2 el45 10.0 8.0 W
34.0 610.9 250.5 10.0 8.0 W
36.31 577.0 282.9 1.4 102 4
HTLP 9 0.5
<2 e65
TABLE 1 (Continued)
PASSIVE HEAT SINKS Thermal Volumeter Heat Wall #
Area Layer Composition Thickness Cond.
Capacity 2
BTU BTU (ft )
O 3 U HR-FT F FT
-F 13 10460 1
Paint 7.5 mills
.2083 36.85 2
Steel
.15 in 28.0 58.5 14 4325 1
Paint 7.5 mills
.2083 36.86 2
Steel 0.2 in 28.
58.5 15 2300 1
Paint 7.5 milis
.2083 36.86 2
Steel 0.432 in 28.
58.3 16 2800 1
Paint 7.5 mills
.2083 36.86 2
Steel 2 3/4 ir 28.
58.5 17 15525*
1 Paint 7.5 mills
.2083 36.86 2
Concrete 1.0 ft 1.04 23.4 18 10031 1
S. Steel
.438 in 8.6 58.5 19 41300 1
Paint 7.5 mills
.2083 36.86 2
Concrete 1.0 f t 1.04 23.4 20 2639 1
S. Steel 3/16 in 8.6 58.5
- In contact with water
TABLE 1 (Continued)
PASSIVE HEAT SINKS Thermal Volumeter Heat Walli Area Layer Composition Thickness Cond.
Capacity 2
BTU BTU (ft)
O 3
HR-FT F FT
-F 1
92100 1
Paint
- 7. 5 ' mi ll's
.2083 36.86 2
Steel 3/8 in 28.
58.5 3
Concrete 1.0 ft 1.04 23.4 2
9700 1
Paint 7.5 mills
.2083 36.86 2
Steel 1/32 in 28.
58.5 3*
9000 1
Paint 7.5 mills
.2083 36.86 2
Steel 1/16 in 28.5 58.5 4*
18000 1
Paint 7.5 mills
.2083 36.86 2
Steel 3/32 in 28.5 58.5 5
17100 1
Paint 7.5 mills
.2083 36.86 2
Steel 1/8 in 28.0 58 5 6
50500 1
Paint 7.5 mills
.2083 36.85 2
Steel 3/16 28.0 58.5 7
9500 1
Paint 7.5 mills
.2083 36.85 2
Steel 1/4 in 28.0 58.5 8
37000 1
Paint 7.5 mills
.2083 36.85 l
2 Steel 3/8 in 28.0 58.5 9
9500 1
Paint 7.5 mills
.2083 36.85 2
Steel 7/16 in 28.0 58.5 10 26200 1
Paint 7.5 mills
.2083 36.85 2
Steel 1/2 in 28.0 58.5 11 21000 1
Paint 7.5 mills
.2083 36.85 2
Steel 3/4 in 28.0 58.5 12 17862 1
Paint 7.5 mills
.2083 36.85 2
Steel
.20 in 28.0 58.5
-m
I TABLE 1 (Continued) i Number of Spray Trains Operating in Max Safeguards Analysis 2
Spray Flow Rate per Spray Train 2600 gpm FAN COOLERS Number of Fan Coolers 5
Number of Fan Coolers Operating in Min Safeguards Analysis 3
Number of Fan Coolers Operating in Max Safeguards Analysis 4
INITIATION TIMES /SETPOINT Delay After Setpoint Containment Setpoint Accident System Used (sec)
Spray 26.7 psig 48.
Fan Coolers 6.2 psig 64.5
~w,-
l, 9
TABLE 1 INITIAL CONDITIONS VALUES CONTAINMENT DESIGN PARAMETERS Containment Design Pressure 47 psig 3
Containment Volume 2,630,000 ft Initial Containment Pressure 0.3 psig Initial Air Partial Pressure 15.1 psia Initial Steam Partial Pressure 0.3 psia Initial Containment Temperature 120 F Refueling Water Storage Tank Inventory 350,000 gal Service Water Temperature 70 F CONTAINMENT SAFETY FEATURES SPRAY SYSTEM Number of Spray Trains 2
Number of Spray Trains Operating in Min Safeguards Analysis 1
~
TABLE Q222.2-1 (Continued)
III.
Auxiliary Feed System Runout Protection Failure Max. Auxiliary Feed Flow Without 1890 gpm
=
Runout Protection Failured 2080 gpm (assumed)
Max. Auxiliary Feed Flow With
=
Failed Runout Protection
=-
w w-y y
yy w
TABLE Q222.2-1 EFFECTS OF SINGLE FAILURES ON CONTAINMENT ANALYSES l
I.
Main Steam ! solation Valve Duration of Piping Break Area Power Piping 8 lowdown 8 lowdown Steam Mass 2
(ft)
(%)
(1b/sec)
(sec)
(Ib)
No MSIV MSIV No MSIV MSIV Forward Reverse Failure Failure Failure Failure 3.4 3.69 102 6113 0.284 3.039 1736 18578 1.4 3.69 70 6589 0.285 3.051 1878 20100 1.4 3.69 30 7267 0.287 3.068 2083 22295 1.4 3.69 0
7809 0.288 3.083 2249 24671 3.69 1.4 102 2315 0.750 8.026 1736 18578 3.69 1.4 70 2495 0.753 8.057 1878 20100 3.69 1.4 30 2752 0.757 8.103 2083 22295 3.69 1.4 0
2957 0.761 8.141 2249 24071 I
failure of a main steam If ne valve was asstaned to increase the unisolatable steam line voltsee frora 981 2
3 ft to 10,500 ft,
Main Feed Line Isolation Yalve 3
Max. Unisolatable Feed Line Volume 93.7 ft
=
Without MFIV Failure 3
Max. Unisolatable feed Line Volume 187.5 ft
=
With F.. IV Failure Closing Time of Fer:d Regulation Valve
<S.0 sec
=
Closing Time of Fred Isolation Valve
<60.0 sec
=
9 I
approved by the NRC staff as comparable to the model recommended in Branch Technical Position CSB6-1.
I 2.
A convective heat transfer coefficient comparable to that recom-mended by the NRC will be used.
If necessary sensitivity studies will be performed, to justify any model differences.
References 1.
Letter to Mr. D. B. Vassallo, Chief, Light Water Reactors Project Branch 6, USNRC From Mr. C. Eicheldinger, Manager, Nuclear Safety, Westinghouse Electric Corporation, Dated March 17, 1976 (NS-CE-992).
2.
Letter to Mr. D. B. Vassallo, Chief, Light Water Reactors Project Branch 6, USNRC From Mr. C. Eicheldinger, Manager, Nuclear Safety, Westinghouse Electric Corporation, Dated July 10, 1975 (NS-CE-692).
3.
Letter to Mr. D. B. Vassallo, Chief, Light Water Reactors Project Branch 6, USNRC From Mr. C. Eicheldinger, Manager, Nuclear Safety, Westinghouse Electric Corporation, Dated April 7, 1976 (NS-CE-1021).
4.
Letter to Mr. J. F. Stolz, Chief Light Water Reactors Project Branch 6, USNRC From Mr. C. Eicheldinger, Manager, Nuclear Safety, Westing-house Electric Corporation, Dated August 27, 1976 (NS-CE-1183).
5.
T. Hsieh et al, WCAP-8936, " Environmental Qualification Instrument Transmitter Temperature Transient Analysis", February 1977.
6.
Letter to John F. Stolz; Chief Light Water Reactor's Project, Branch 6, USNRC from C. Eicheldinger, Manager, Nuclear Safety, Westinghouse
}
Electric Corporation. Dated June 14, 1977 (NS-CE-1453).
t 7.
Bordelon, F. M., Murply, E. T., WCAP-8327, Containment Pressure Analysis Code (C0CO) July 1974
The results of all of these analyses are tabulated in Table 2.
The case resulting in the highest peak pressure has been identified as This case resulted in a peak con-the 3.69 ft2 Break at 70% power.
When tainment pressure of 44.9 psig when dry steam blowdowns were used.
this same case was analyzed using blowdowns which included the effect of liquid carryover from the secondary side the resulting pressure was 41.8 This f
psig for both the Westinghouse and NRC containment models cases.
indicates the overall conservatism of the Westinghouse large break con-The requested transients for tainment model when used with dry steam.
the Westinghouse model with dry steam blowdowns have been provided in Figures 2 to 4.
The case resulting in the highest calculated peak temperature has been identified as the 0.910 at 70% power. This case resulted in a peak When this same case was analyzed with the NRC temperature of 333 0F.
These containment model a peak temperature of 3440F was calculated.
results verify.nat the Westinghouse and NRC models obtain similar The requested transients for both of these cases have been results.
provided in Figures 5 to 10.
An evaluation of the safety related instrumentation will be performed to This evalua-show conformance with the requirements of IEEE-323-1971.
tion will be performed by comparing the containment equipment test con-ditions versus the calculated containment accident environments previ-If a thermal analysis is neccemry Westinghouse will ously discussed.
use a thermal model similar that in reference 5.
Any differences be-tween the Westinghouse thermal analysis model and the proposed NRC interim model will be discussed and justified.
Some major points of the model are the following:
The condensing heat transfer coefficient will be the same as used in 1.
This model is the. approved Westinghouse model for ECCS analysis.
documented in Appendix A of WCAP-8339 and has been reviewed and
PGE MSLB Analysis In order provide a response to NRC Staff questions relating to the analyses of postulated MSLB breaks inside containment the following information is provided.
The NRC question numbers, which the various portions of the writeup address, are noted in the right hand columns. All NRC questions relat-ing to the containment analysis for a steamline break are addressed by the following writeup.
Analysis Methodolooy and Introduction Steambreak analyses have been performed for the Diablo Canyon plant.
Various containment models have been utilized in conjunction with the mass and energy releases described in Questions 222.1 through 222.6.
The majority of the analyses performed utilized the Westinghouse con-tainment model developed for the IEEE-323-1971 Westinghouse Supplemented Equipment Qualification Program.
These models and their justification (experimental and analytical) are detailed in references 1 to 5.
Some major pcints of the model are as follows:
1.
The saturation temperature corresponding to the partical pressure of the containment vapor is used in the calculation of condensing heat transfer to the passive heat sinks and the heat removal by contain-ment for coolers.
2.
The Westinghouse containment model utilizes the an&iytical approaches described in references 5 and 7 to calculate the conden-sate removal from the condensate film.
Justification of this model as provided in References 1, 4, 5, and 7.
(For large breaks 100%
revaporization of the condensate is used, and a calculated frac-tional revaporization due to convective heat flux is used for small breaks.)
3.
The small steamline break containment analyses utilized the stagnant Tagami correlation, and the large steamline break analyses utilized the blowdown Tagami correlation with an exponential decay to the stagnat Tagami correlation. The details of these models are given in Reference 7.
Justification of the use of these heat transfer coef ficients has been provided in References 1, 4, and 6.
The break spectrum for the MSLB mass and ensegy releases detailed pre-viously in Questions 222.1 to 222.6 have been analyzed using the West-inghouse containinent model. Adequate justification has been provided to NRC staff to obtain acceptance of this model.
However, in order to expedite review of the Diablo Canyon Docket, some additional studies have been performed.
In order to resolve any differences between the NRC and Westinghouse containment models, the worst case small break, for which liquid entrainment have been provided, along with the respective containment transients.
MSLB Containment Analysis The containment input used in these calculations has been provided in Table 1, and Figure 1.
The times assumed in the analysis for the initi-ation of the containment sprays and for coolers arc based on an assumed loss of offsite power with delays including the following effects:
1.
Time for safety equipment to reach full speed.
2.
Time for the filling of the appropriate spray headers and piping.
Containmeni. analyses assuming both minimum and maximum containment safe-guards, along with consistent mass and energy release data have been performed.
quantify the effect of the differences between the Westinghouse contain-ment model and the NRC containment model.
These results are also included in Table 2.
Mass and energy releases for the two " worst" case results are provided in Table 3.
Time histories of containment temperature, pressure, and heat transfer coefficient, for the " worst" case results are provided in Figures 2 through 10.
4 x1 the auxiliary feed system line losses.
For Diablo Canyon this flow is 1840 gpm.
To evaluate failure of the auxiliary feed runout protec-tion system, it was assumed that this flow would be increased by 10% if the auxiliary feed pumps were operating at their limiting govenor speed and all flow control valves completely open.
In this condition the Diablo Canyon auxiliary feed flow would be 2080 gpm.
In the analyses, these flows were assumed to start at time = 0 and were terminated at 10 minutes.
Question 222.4 All blowdowns used in the Diablo Canyon main steam line break contain-ment analyses were assumed to be dry steam.
No credit was taken for moisture entrainment in the large breaks.
Questions 222.c and 222.6 Containment analyses for a total of 50 breaks were performed for the Diablo Canyon Nuclear Unit.
These include consideration of 16 double-ended ruptures upstream of the stream line venturi with area = 3.69 ft2,16 double-ended ruptures downstream of the steam line flow 2
restrictors with area = 1.4 f t, and 16 split ruptures of a size just small enough to prevent initiation of steam line isolation or a primary protection system signal.
For the small split ruptures, feed and steam line isolation result from high containment pressure trips.
The con-tainment input assumption's are presented in Table 1 and Figure 1.
Results for these 48 cases are provided in Table 2 along with identifi-cation of plant power level, single failure, and pertinent trips associ-ated with each.
The table also identifies the case which results in the worst temperature and the case which results in the worst pressure.
Two additional analyses were performed using the C0C0 code modified to agree with the NRC interim proposed containment model.
In these analyses the worst temperature and worst pressure cases were analysed to
total mass in released to the containment during the blowdown tran-sient. A tabulation of these masses for Diablo Canyon are given in Table 1.
In analysis of both the double-ended and split ruptures, failure of a i
main steam isolation valve results in increasing the mass of steam which I
cannot be isolated from the pipe break.
This increase in volume results in an increase in the duration of the initial piping blowdown for the double-ended breaks, and increases the mass added to the faulted loop steam generator in the split rupture cases. A tabulation of the steam line masses and the piping blowdown times with and without an isolation valve failure are given in Table Q222.2-1.
Table Q222.2-1 also includes the effects of other single failures on the main steam line break calculations.
Question 222.3 Calculation of feedwater flows to the steam generator with the broken steam line were done making use of the Diablo Canyon feedwater system line losses and pump head cures.
The forcing function for the increased flow was assumed to be the pressure in the faulted steam generator as calculated by MARVEL.
Thus the calculations are conservative since no credit is taken for pressure losses in the piping due to feedwater flashing.
(Feedwater flashing calculations are performed by MARVEL; reference section 2.23 of WCAP-8843). Other assumptions used in calcu-lating feedwater flows are listed in section 3.1.A of Appendix A of p
The results of the feedwater calculations for the worst case f
containment transients are provided in Figure Q222.3-1.
To determine the auxiliary feedwater flows, all auxiliary feed pumps 4
were assumed running with the faulted steam generator completely depres-scrized and the remaining steam generators at the safety valve pres-The flow rate is then established based on the pump head curves k
sure.
i, I
1a m-q j
I Question 222.1 i
Calculation of the mass and energy releases from a broken steam line were made using the MARVEL code which is described by WCAP-8843.
All blowdowns used in this analysis were assumed to be dry steam and were generated in a manner completely analogous to the blowdowns submitted on the Sundesert application.
Heat transfer models for primary-secondary i
energy transfer are described in Section 2.7, 2.17, 2.19, and 2.24 of that WCAF.
Core heat transfer models and kinetics calculation tech-niques are addressed in sections 2.1, 2.2, and 2.4 Modeling of thick metal stored energy is explained in section 2.34, and the flashing of feedwater in unisolated portions of the feed under lines is described in section 2.23.
The determination of increased feed water flow due to pump runout and steam generator depressurization are addressed in the response to question 222.3.
Question 222.2 Two methods are used to model flow of steam from the unisolated portions of the steam lines. For double-ended ruptures, the steam from these lines is added to containment at a constant rate starting at the time of the break and continuing until the entire unisolted line is completely purged.
The flowrate and enthalpy of this constant flow is established by the initial plant operating condition.
The flow is assumed to be from the intact steam loops and is followed by flow from the steam gen-erators in the intact loops.
This technique is described in greater detail in the sample calculations provided in Appendix A of WCAP-8822.
Values used for these piping blowdowns in the Diablo Canyon analyses are given in Table Q222.2-1.
For the small split breaks, steam from the unisolated steam lines is included in the analyses by adding the mass of steam in the lines to the initial mass assumed in the faulted loop steam generator.
Thus this
TABLE 3 (Continued) l MASS AND ENERGY RELEASES f
Time (sec)
Mass Flow (lb /sec)
Eneray Flow (M8tu/sec)_
l m
'r i
20.0 1462.7 1.754 g
l 25.0 1543.0 1.849 i
30.0 1557.0 1.865 35.0 1534.7 1.839 40.0 1503.3 1.802 45.0 1466.8 1.759 50.0 1429.5 1.715 60.0 1357.2 1.630 70.0 1289.2 1.549 80.0 1225.9 1.474 90.0 1168.5 1.405 94.5 1110.8 1.337 95.0 1090.9 1.313 100.0 940.7 1.133 110.0 777.8 0.937 120.0 681.5 0.820 130.0 620.6 0.747 140.0 580.0 0.698 150.0 551.0 0.663 175.0 504.1 0.606 200.0 474.2 0.570 225.0 451.9 C.543 250.0 429.9 0.516 300.0 391.7 0.470 350.0 360.3 0.432 400.0 334.1 0.400 450.0 311.7 0.373 500.0 292.4 0.349 550.0 275.9 0.329 600.0 261.5 0.312 700.0 218.0 0.259 800.0 182.9 0.217 900.0 143.1 0.169 1000.0 91.4 0.107
.j TABLE 3 (Continued) i k
I MASSANDENERGYRELEA5ES II.G.
Mass and Energy Releases From a 3.69 ft2 DER at 70% Power -
Including Entrained Moisture Effec's l
Time Break Flow Energy Flow (sec)
(1bs/sec)
(Milion Stu/sec)
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TABLE 3 (Continued)
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(1bs/sec)
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Power with Maximurn Safeguards (Westinghouse Model)
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U Figure G Containment Heat Transfer Coef ficient for 0.010 ft2 Split Break at 70% Power witti Minimurn Safeuuards (Westingtiouse Model)
500 1100 CONTAINHENT ATH0 SPHERE E 300 ~
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Figure 7 Contain"ient Temperature Transient for 0.910 f t Split Break at 70% Power h
2 with Minimum Safeguards (Westinghouse Model)
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Figure 8 Containment Pressure Transient for 0.010 f t Split Break at 70% Power 2
with Minimum Safeguards (NRC Model)
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Containment licat Transfer Coefficient for 0.910 Ft2 S
Split Break at 70%
Power witti Minimum Safeguards (NRC Model)
-