ML18078A415

Answers NRC Questions 5.62,5.63,5.82,5.83,5.84 & 5.85 Re Analyses of Main Steam Lines Break within Containment of Unisolated Steam in Main Steam Line & Turbine Plant

ML18078A415
Person / Time
Site:	Salem
Issue date:	11/20/1978
From:	Public Service Enterprise Group
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Shared Package
ML18078A414	List: ML18078A413 ML18078A415
References
NUDOCS 7811240142
Download: ML18078A415 (44)

Text

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QUESTION 5.82 .*,

Provide the results of analyses for a spectrum of main steam line breaks within the containment. You should vary break size and power level to identify the break sizes producing the highest containment temperature and pressure. Provide ~-~

mass an.d energy release data for these two break sizes. Pro- ~ .*

vide a complete description for the method used to calculate the flow from the steam generator into the containment. In-*

elude justification for all equations and assumptions. Pro-vide a comparison of the method used in the above analyses with the method described in WCAP-8843 (MARVEL). The anal-yes should include the results of postulated single failures in the steam and feedwater system.

ANSWER A complete analysis of main steamline breaks inside containment has been performed using the MARVEL code (WCAP-8843) and the Westinghouse containment computer code, COCO, as described in WCAP-8936 and its references. A total of forty-eight (48) different blowdowns covering four power levels and three different break sizes were evaluated. The three break sizes considered at each power level were a full double-ended rupture

~-* upstream of the steamline flow restrictor, a full double-ended rupture downstream of the steamline flow restrictor, and the largest split rupture that will not result in generation of a steamline isolation signal from the primary plant protection

',

t- equipmentl. In the analysis of the third (split) break reactor trip, feedline isolation and steamline isolation are generated by high containment pressure signals. Additionally, all blowdowns used in the analyses were assumed to consist of dry steam.

1 Reference Section 2.3 of WCAP-8822 for a complete discussion of this split break.

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f For each break condition, four different single failures were evaluated. These were (1) failure of a containment safeguards train, (2) failure of a main feed isolation valve, (3) failure of a main steam isolation valve, and (4) failure of the auxiliary feedwater runout protection equipment.

The worst effect of a containment safeguards failure is the loss of a spray pump which reduces containment spray flow by 50%. In all analyses, the times assumed for initiation of containment sp~ays and fan coolers are 59 and 35 seconds respectively following the appropriate initiating trip signal.

These times are based on the assumption of a loss of offsite power, and the delays are consistent with Technical Specifi-cation limits. The delay time for spray delivery includes the time required for the spray pumps to reach full speed and the time required to fill the spray headers and p1ping.

There are two valves in each main feedwater line which serve to isolate main feed flow following a steamline break. One is the main feed regulator valve which receives dual, separate train trip signals from the plant protection system on any safety injection signal and closes within five seconds of receipt of this signal. The second is the feedline isolation valve which also receives dual, separate train trip signals from the pr~t~ction system following a safety inje_ction signal_..___________ _

This valve closes within 30 seconds. Additionally, the main feed pumps receive dual, separate train trip signals from the P78 155 56

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. :- protection system following a steamline break. Thus, the worst failure in this system is a failure of the main feed regulatory~i valve to close. This results in an additional 25 seconds durin~

which feedwater from the condensate feed system may be added to the steam generator. Also, since the feed isolation valve is upstream of the regulator valve, failure of the regulator results in additional feedline volume which is not isolated from the steam generator. Thus, water in this portion of the lines ca~ flash and enter into the steam gener~tor.

Both of these effects were considered in the analyses when failure of a main feed regulator valve was assumed.

Since all main steam isolation valves have closing times of no more than five seconds, failure of one of these valves affects only the volume of the steam and turbine steam piping which cannot be isolated from the pipe ruptur~.

This effect was considered in the analyses as described in I

the response to Question 5.83.

In all the containment evaluations following a steamline rupture, the entire auxiliary feed system was assumed to be operating starting at the time of the break. Flow to the faulted steam generator from this system was determined assuming one steam generator completely depressurized and th_~ __ rem_aining steam generators at the safety valve pre_ssure.

System hydraulic resistances and pump head/flow curves were used in the flow calculation. However, failure of the run-P78 155 57

. r out protection system could result in all auxiliary feed pumps being at their maximum governor speed and flow control valves in th~ auxiliary feedlines being fully open. These changes will result in a higher flow to the steam generator. The effects of such failures were considered in the containment analyses.

All blowdown calculations with the MARVEL code were done assuming the reactor coolant pumps were running, (i.e.

offsite power available) because this increases the primary-to-secondary heat transfer and therefore maintains higher blowdown flow rates (Ref. Section 3.1.7 of WCAP-8822).

Although this is inconsistent with the delay times assumed in containment fan cooler and spray initiations (see above explanation) where loss of offsite power is assumed, the combined effect is extra conservatism in the calculated containment conditions.

Containment initial condition values are given in Table I, and the containment temperatures and pressures resulting from all cases considered are presented in Table 2 along with pertinent trips, trip times, and single failures associated with each. Also shown in Table 2 are four additional entries. These show the results of analyses of the worst temperature and pressure transients as analyzed with the COCO code modified to conform to the NRC interim containment evaluation model and the results P78 155 58

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of the worst pressure transient initiated by a double-ended ~upture when analyzed assuming entrainment in the blowdown as specified in Section 3.2.2 of WCAP-8822. As discussed with the staff previously, these results are pro-vided for comparison of Westinghouse and NRC containment models and for quantifications effects on peak pressure from entrained moisture which is expected to be present in large break blowdowns. As can be seen from the table, the peak pressure for any case using the Westinghouse model is 42.8 psig and the peak temperature for any case using the Westinghouse model is 333.50F. A discussion comparing the results of the NRC and Westinghouse models is provided later.

Mass and energy releases for the two worst cases are pro-vided in Table 3. Graphical results showing containment I

k atmospheric temperature, containment pressure, and other pertinent variables are provided in Figures 1 through 15.

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QUESTION 5.83 Discuss the method by which unisolated steam in the main steam line and turbine plant is added to the containment. frovide a table of unisolated steam mass as a function of power level for when the steam line isolation valve closes and ~hen it is -~

assumed not to close. ":J; ANSWER Steam contained in the unisolated portions of thE steamlines and turbine plant were considered in the containment analyses in two ways. For the large double-ended ruptures, steam in the unisolated steamlines is released to the containment as part of the reverse flow. This is accom~lished by having the reverse flow begin at the time of the break at the Moody'critical flow rate for steam as established by the cross-sectional area of the steamline and the initial steam pressure. The flow is held constant at this rate for a time period sufficient to purge the entire unisolated portion of the steamlines. Enthalpy of the flow is also held constant at the initial steam enthalpy. Following the period of constant flow representing purging of the steamlines, flo~

from the intact steam generators, as calculated by MARVEL, is added to the containment and continues until steamline isolation is complete.

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When considering the split ruptures, steam in the steamlines is included in the analysis by adding the total mass in the

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lines ~o the initial mass of steam in the faulted steam ~-

generator. This is necessary because, unlike double-ended ruptures, the total break area for a split break is unchanged by stearnline isolition; only the source of the blowdown effluent is changed. Thus, steam flow from the ~iping in the intact loops is indistinguishable from steam leaving the faulted steam generator. Bowever, by adding the pi~ing mass to the faulted steam generator mass, and by having dry steam blowdowns, the steamline inventory is included in the total blowdown.

As discussed in the response to Question 5.82, failure of a main steamline isolation valve only results in a larger unisolatec volume in the stearnlines. This failure is con-sidered in the analyses by increasing the duration of the piping blowdown in the large breaks, and by increasing the

!. mass added to the steam generator in the small s~lit breaks i

t by amounts sufficient to cover the additional steam mass in the stearnlines. Table 4 shows the mass in the stearnlines with and without an isolation valve failure at the four power levels considered in the analyses.

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QUESTION 5.84 Discuss the method by which unisolated feedwater in the main and auxiliary feedwater system is added to th~ affected steam generator following a postulated main steam line break. Pro-vide the mass of unisolated feedwater with and without a fail-ure of a feedwater isolation valve.

ANSWER Feedwater in the unisolated portions of the feedlines are added to the faulted steam generator as a result of flashing as the steam pressure drops. Calculation of feedwater flashing is performed by the MARVEL code as described in Section 2.23 of WCAP-8843. For the Salem Nuclear units, the maximum volume of unisolated feedlines is 328.2 ft3 ~ithout a feed regulator valve failure, and incrrases to 868.5 ft3 with a feed regulator valve failure. (See Table 4)

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QUESTION 5.85 Justify by analysis the values assumed for feedwater flow into the ruptured steam generator following a main steam line break.

Flashing of the fluid in the feedwater lines should be con-sidere~ as well as the affect of reduced discharge ~ressure the flow rates through the feedwater and condensate pum~s.

ANSWER Feedwater flow to the faulted steam generator from the main feed system is calculated using the hydraulic resistances of th~ system piping, head/flow curves for the main feed pum~s, and the steam genera~or pressure decay as calculated by the MARVEL code. In the calculations performed to match these systems variables, a variety of assumptions are made to maximize the calculated flows. These include:

1. No credit for extra pressure drop in the feedlines due to flashing -of feedwater (feed flashing is calculated by MARVEL code; see response to Question 5. 84)

2. Feed regulator valve in the faulted loop is full open.

3. Feed regulator valves in the intact loop do not change position prior to a trip signal and close instantly upon receipt of a signal to close.

4. All feed pumps are running at maximum speed.

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5. No credit is taken for flow reduction through the feed regulator or feed isolation valve until they are fully closed.

6. Flow from the pum~s decay linearly following pump triF.

Results of these calculations for the two worst case contain-ment transients (see Table 2) are shown in Figure 16.

Calculation of the flow into the faulted steam generator from the auxiliary feed system is discussed in the res~onse to Question 5.82. Values of the maximum auxiliary feed flow

~ith and without failures of the runout protection system are included in Table 4.

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In order to provide a response to the NRC Staff questions 5.62 and 5.63 parts 2 and 3 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 relating to the containment analysis for a steamline break are addressed by the following writeup.

Analysis Methodology and Introduction Steambreak analyses have been performed for the Salem Nuclear plant.* Various containment models have been utilized in conjunction with the mass and energy releases described in the respnses to questions 5.82 through 5.85.

The majority of the analyses ~erformed utilized the ~estinghouse containment model developed for the IEEE-323-1971 Equipment Qualification program. These models and their justification (experimental and analytical} are detailed in references 1 to 5.

Some major points of the model are as follows:

1. The saturation temperature corresponding to the ~artial pressure of the containment vapor is used in the cal- . 5. 62 (d}

culation of condensing heat transfer to the passive heat sinks and the heat removal by containment fan coolers.

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2. The Westinghouse contajnment model utilizes the analytical approaches described in References 5 and 7 to calculate the condensate removal from the condensate film. Justi-

.;:.f s. 62 f ication of this model is provided in Referenc~s 1, 4, 5, 4. (e) and 7. (For large breaks 100% revaporization of the con- "

densate is used, and a calculated fractional 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 ex~onential decay to the stagnant Tagami 5.62 correlation. The details of these models are given in (c)

Heference 7. Justification of the use of these heat transfer coefficients has been provided in References 1, 4, and 6.

The break spectrum for the MSLB mass and energy releases de.tailed previously in Questions 5.82 through 5.85 have been analyzed using the Westinghouse containment model. Adequate justification has been provided to .the NRC Staff to obtain acce~tance of this model. However, in order to expedite review of the Salem Docket, some additional studies have beem performed.

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In order to resolve any di[~erences between the NRC and the Westinghouse containment models, the worst case small break 5.62 (for which liquid entrainment from the secondary side is not (C 1 d 1 E ex:tiec1:.ed) has been analyzed with both the \twes ti nghouse and the 4.

NRC containment models. For the worst case large break, mass and energy releases which consider liquid entrainment have been provided, along with the respective containment transients.

MSLB Containment Analysis The containment input used in these calculations is provided in Table 1, and Figure 17. The times assumed in the analysis for 5.62 (b) the initiation of the containment sprays and fan coolers are based on the loss of offsite Fower with delays including the following effects:

1. Time for safety equipment to reach full speed

2. Time for the filling of the appropriate SFray header~

and piping Containment analyses assuming both minimum and maximum 5.62 (a) containment safeguards, along with consistent mass and I

energy release data have been performed. The results of all of these analyses are tabulated in Table 2.

I 5.62

(~)

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The large break case resulting in the calculated ~eak

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pressure has been identified as the 1.4 ft2 break at 70% power. This case resulted in a peak pressure of 39.1 psig when dry steam blowdowns were used. ~hen this same case was reanalyzed utilizing blowdowns which included the effect of liquid carryover from the secondary side, the resulting peak pressures were 37.7 and 37.2 using the Westinghouse and NRC containment models respectively.

This indicates the overall conservatism of the Westinghouse containment model when used with dry steam, vs. using the exp~cted mass and energy releases which include the effect of entrainment. The requested transients for the westinghouse inodel with dry stern blowdowns are provided in Figures 1 througb 3.

The case resulting in the calculated peak ~ressure for the I s. 62 (h) small breaks has been identifi.ed as the 0.86 ft2 break at 102% power *. The resulting peak pressure for this case was 42.8 psig. When this case was reanalyzed utilizing the NRC containment model, and the s~me mass and energy release rates, the peak calculated pressure was found to be 43.0 psig. The requested transients for the hestinghouse model are provided in Figures 4 through 6. Similar transients for- the case which used the NRC model are provided in Figures 7 through 9.

I 5.62 (h)

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The case resulting in the calculated peak tem~erature has been identified as the 0.908 ft2 break at 70% power. This case resulted in a peak temperature of 333.5°F. When this same case was analyzed ~ith the NRC containment model a peak tern~-

erature of 3470F was calculated. These results verify that the ~estinghouse and NRC models yield similar results. The requested transients for both of these cases have been provided in Figures 10 through 15.

An:evaluation of the safety related instr~rnentation will be performed to show conformance with the requirements of IEBc-223-1971. This evaluation will be performed by comparing the containment equipment test conditions versus the calculated 5.6 art containment accident environments previously discussed. If a thermal analysis is necessary Westinghouse will use a thermal model similar to that presented in Reference 5. Any differences between the Westinghouse thermal analysis model and the proposed NRC interim model will be discussed and justified. ~ome major points of the model are the following:

1. The condensing heat transfer coefficient will be the same as used in the approved ~estinghouse model for ECCS analysis.

This model is documented in Appendix A of hCAP-8339 and has 5.63 Part been reviewed and appro~ed by the NRC Staff as cornpa~able (a}

to the model recommended in branch Technical Position CSB6-l.

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2. A convective heat transfer coefficient comparable to that 5.63 Part recommended by the NRC will be used. If necessary, sensi- ~r (b) tivity studies will be performed to justify any model ~

differences.

5.0 HEF'ERENCES

1. Letter to Mr. D. B. Vassallo, Chief, Light WAter Reactors Project Branch 6, USNRC from Mr. c. Eicheldinger, Manager, Nuclear Safety, ~estinghouse Electric Corporation, Dated March 17, 1976 (NS-CE992).

2. Letter to Mr. D. B. Vassallo, Chief, Light ~ater Reactors Project Branch 6, USNRC From Mr. C. Eicheldinger, ~anager, Nuclear Safety, westingbouse 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, ~estinghouse Electric tor~oration, Dated April 7, 1976 (NS-CE-1021).

4. Letter to Mr. J. F. Stolz, Chief, Light Water keactors Project Branch 6, USNRC from Mr. c. Eicheldinger, Manager, P78 155 70

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Nuclear Safety, Westinghouse 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 Reactors Project, Branch 6, USNRC from c. Eicheldinger, Manager, Nuclear Safety, Westinghouse Eleetric Corporation, Dated Jun~ 14, 1977 (NS-CE-1453).

7. Bordelon, F. M., Murphy, E. T., *wcAP-8327, Containment Pressure Analysis Code (COCO) July 1974.

RWS/jk P78 155 55/70

TABLE 1 1

INITIAL CONDITI_Qfil) VALUES _.

~ONTAINMENT DESIGN PARAMETERS Containment Design Pressure 47 psig Containment Volume 2,620,000 ft3 Initial Containment Pressure 0.3 psig Initial Air Partial Pressure 14.7 psia Initial Steam Partial Pressure 0.3 psia Initial Containment Temperature 1200F Refueling Water Storage Tank Inventory 350,000 gal Service Water Temperature 850

• SPRAY SYSTEM Number of Spray Trains 2 Number of Spray Trains Operating in Minimum Safeguards Analysis 1 Number of Spray Trains Operating in Maximum Safeguards Analysis 2 Spray Flow Rate per Spray Train 2600 gpm FAN COOLERS

• Number of Fan Coolers 5 Number of Fan Coolers Operating in Minimum Safeguards Analysis 3 Number of Fan Coolers Operating in Maximum Safeguards Analysis 4 INITIATION TIMES/SETPOINT

1. -

System Containment Setpoint used Delay After Setpoint (sec)

Spray __ _ 26.7 psig 59.

Fan Coolers 7.9 psig 35.

TABLE 1 SHEET 1 of 3 P78 155 71

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TABLE 1 (Cont'd)

PASSIVE HEAT SINK Volumetric Thermal Heat (ft2) ft. Cond. Capacity wall t Area La~er Comeosition Thickness BTU/HR-FT-OF BTU/FT3-0F 1 451~9 1 Paint 0.000625 0083 39.6f

' 2 Steel 0.03125 27.0 58.i.

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3 Concrete o.s 0.92 22.6 4 Concrete 4.0 0.92 22.6 2 14206 1 Insulation 0.2083 0.024 3. 9 4 2 Steel 0.03125 27.0 58.8 3 Concrete 0.5 o. 92 22.6 4 Concrete 4.0 0.92 22.6 3 29249 1 Paint 0. 0 00625. 0.083 39.6 2* Steel 0.04167 27. 0. 58.8 3 Concrete 0.5 o. 92 22.6 4 Concrete 3.0 0.92 22.6

• 4 11611 1 Paint 0.0015 .083 39.6 2 Concrete 0.5 o. 92 22.6 3 Concrete 3.0 0.92 22.6 5 6806 1 Paint 0.0015 0.083 39.6 2 Concrete 0.5 -0. 92 22.6

-3 Concrete 0.5 o. 92 22.6 4 Concrete 0.5 o. 92 22.6 6 9424 1 Paint .0015 0.083 39.6 2 Concrete 0.5 0.92 22.6 3 Concrete 1.21 0.92 22.6 7 31660 1 Paint 0.00117 0.083 39.6 2 Concrete 0.5 0.9 2 22.6 3 Concrete 1.0 0 .. 92 22.6 Stainless 8 13279 1 Steel 0.01773 8.0 53.6 2 Concrete 0.5 0.92 22.6 3 Concrete 1.4 0.92 22.6 9 47590 1 Paint 0.000625 0.083 39.6 2 Steel 0.011 27.0 58.8 TABLE 1 SHEET 2 of 3

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RfEULTS FOR SPILT RUP'ruRES Reactor Trip Tt11111 of Hf-1 Tf11111 of Ht-2 Tt 11111 of Fl!fidl hie Th11 of Steam LI ne Th11of ~ak rak Ar..

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Oontal,_nt Preti,

!sec!

Contalnm1mt Pret1, jaec)

Isolation jsec)

Isolation jsec)

... Contal..-nt Model Paak Pressure

!~lg!

Pr ensure

{secJ Pi!ak ~-

(OP')

.1160 102 I IN' ' 14. 7 23.0 95.0 33.0 101.0 w 42.8 809,6 330.9

,860 2 IM' ' 14. 7 23.0 101.0 58,0 107.0 w 37.l ,741.l . 322.2

,860 l HNP' ' 14.65 23.0 97.5 33.0 103,5 w 36.5 701.0 323.6

.860 4 IM' ' 14.95 23.5 97.5 33.5 103.5 w 36.7 696.4 323.8

.908 70 I CPI P 21.0 21.0 89.5 31.0 95.5 w ~2.1 870,4 JJJ.5

,909 2 CPI P 21.0 21.0 95.0 5fi,O 101.0 w 35.5 718.9 324.5

,908 l CPI P 21.0 21.0 91.5 31.0 97.5 w 35.4 674.7 326.4

.908 4 CPI P 21.0 21.0 91.0 31.0 97.0 w 35.6 673.l 326.7

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.942 JO 1 CPI g, 19.0 19.0 89.0 29.0 95.0 w 37.8 870.5 332.3

.942 2 CPI P 19.0 19.0 95,0 54.0 101.0 w 31.6 579.5 323.l

.942 3 CPI P 19.0 19.0 91.5 29.0 97.5 w 31.9 592.3 325.2

.942 4 CPI P 19.0 19.0 90.5 29.0 96.5 w 32.1 574.4 325.5

.960 0 l N/1' 17.0 90.0 27.0 96.0 w 32.7 791.5 329.8

.960 2 Nil' 17.0 96.0 52.0 102.0 w 29.2 156.2 321.2

.960 3 Nil' 17.5 93.0 27,5 99.0 w 29.3 152.4 322.7

.960 4 Nil' 17.0 91.5 27.0 97.5 w 29.6 151.l 32:..3

.90ll 70 I CPI P 21.5 21.5 93.0 31.5 99.0 N 42.4 895.l 347.0

.1160 102 l IM 14. 7 23.5 99.0 33.5 105,0 N 43.0 830.2 344.7 TMLE 2 SHEET l of 4 I HO 416/611 e

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TABLE 2 (Cont'd)

RESULTS POR SPILT RUP'IURF.S Reactor Trip Time of RI-I ----,.iiNt-of~----orrme-of FffdTing-----rJiiiil--of-S-team Line Th'I! of Peak

~nllk Ar* ~r Lewl Single Tl* O:>ntall'lllll!nt Pres. Contalrwn11nt Pres. Isol11tlon l!lollltlon O:>ntal..-nt P9ak Preaure Pressure Peak ,....,.

(ft2) I*! r11llure (sec) (sec) (11ec) (sec) (sec) Model (pslg! (11ec) (OF) 1.4 102 1 HPt.P ' 0.5 -3 -141 10.0 8.o w )8,7 . ,632.8 . 260.0 1.4 2 RP1.P ' 0.5 -3 -133 35.0 8.0 w 37.l 609.0 260.0 1.4 3 HFt.P ' 0.5 -3 -142 10.0 8.o w 35.2 615.3 260.0 1.4 4 HFLP P 0,5 -1 - 62 10.0 8.0 w 37.3 602.0 292.9 4.25 l Hrt.P P 1.0 -4 -112 10,0 8.5 w 35.3 630.3 249.4 4.25 2 RP1.P ' 1.0 -4 - 91 35.0 8.5 w 35.3 313.6 253.8 4.25 3 RP1.P ' 1.0 -4 *111 10.0 8.5 w 32.3 606.5 246,6 4.25 4 llPt.P ' 1.0 -3 - 56 10.0 8.5 w 34.l 214.7 282.4 1.4 1.4 1.4 70 l 2

3 RP1.P ' o.5 llFU' .. 0.5 llFU' ' 0.5

-3

-3

-3

-133

-140

-146 10.0 35.0 10.0 8.0 8.o e.o w

w w

39.l 36.4 35.0 657.0 615.9 617.l 262.l 262.1 262.l e

1.4 4 HFU' P 0.5 -1 - 53 10.0 8.o w 38.l 575.8 296.9 4.25 l HFLP 111 1.0 -4 *110 10.0 8.5 w 36.2 626.0 251.2 4.25 2 HFLP e l.o -4 - 95 35.0 8.5 w 36.7 376.8 256.l 4.25 3 Hn.P P 1.0 -4 -111 10.0 e.5 w 33.15 620.4 248.5 4.25 4 RP1.P ' 1.0 -2 - 54 10.0 8.5 w 35.4 20.0 285.4 1.4 JO 1 RPtP ' 0.5 -3 -141 10.0 8.o w 38.0 747.l 2f'"i.6 1.4 2 llFLP ' 0.5 -3 -146 35.0 8.o w 34.l 617.9 265.6 1.4 3 llFU' ' o.5 -3 -143 10.0 e.o w 33.0 631.5 265.6 1.4 4 llFt.P ' 0.5 -1 - 36 10.0 e.o w 36.95 582.0 301.9 4.25 l Hn.P Ill 1.0 -4 -116 10.0 R.5 w 37.4 614.2 254.2 4.25 2 HFLP Ill 1.5 -4 -102 35.0 9.0 w 38.0 491.0 258.2 4.25 3 llFLP ' 1.5 -4 -113 10.0 9.0 w 34.3 609.0 251.2 4.25 4 llPU' ' 1.5 -2 - 46 10.0 9.0 w 37.1 309.6 290.7 TABLE 2

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Rt!actor Tdp Tl11111 (sec)

NI" Tlriie-of-llT -

Contelnlnl!nt Prll!s.

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Contnlnment Pres.

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tsolntlon (sec) 8.5 Contall"f!ll!nt Modll!l w

Peek Pr4t!1!1Ure (pslg) 35.0 T{lnl! ot Pr~surll!

(sec) 743.4 Peek 1"1111< ~.

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269.8 el

.4 2 NI" -3 -161 35.0 8.5 w 31.3 465.8 269.8

.4 3 N,?. -3 -150 10.0 8.5 w 30.6 434.5 269.8

.4 4 N,?. -1 - 17 10.0 8.5 w 35.65 406.5 305.2

.25 l NI" -3 -111 10.0 9.5 w 36.4 646.6 252.3

.25 2 NI" -3 -111 35.0 9.5 w 36.2 472.0 255.2

.25 3 NI" -3 -112 10.0 9.5 w 32.8 611.0 251.9

.25 4 N,?. -2 - 38 10.0 9.5 w 36.4 2!12 .1 2Q4.I

• 4 70 l RPtP (I 0.5 -3 -136 10.0 8.0 NW 37.2 644.7 273.6

.4 l ffft.P " o. 5 -3 -121 10.0 e.o 37. 7 639.6 253.6

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TABLE 2 NOTES I. Single Failures l = Failure of a containment safeguards train 2 = Failure of a feedline isolution valve 3 = F~ilure of the auxiliary feerlwater runout protection 4 = Failure of a main steam isolation valve II. Trip Symbols HNF = High Nuclear Flux Trip.

HFLP = High Steam Flow/Low Steamline Pressure Trip CPI = High 1 Containment Pressure Trip III. Containment Models W =Westinghouse C~C¢ Model, Dry Steam Blowdowns N = NRC Small Break *Model, Dry Steam Blowdowns WN = Westinghouse C~C~ Model, Wet Slowdowns NW = NRC Model, Wet Blowdowns TABLE 2 SHEET 4 of 4 P78 140 75

.TABLE 3A - MASS AN,!NERGY RELEASES FROM A O.

AT 70% POWlR 1Worst Temterature Case) le P'T2 SPLIT BREAK

• Time Break Flow Energy Flow (sec.) (lb/sec.) (million btu/sec.)

a ,c:

. /~,,e ;'O ~~,,e # Of'~-?~-R/

t!//EJ) ~Jf;/I'/~

.

~ /

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TABLE 3B - MASS AND ENERGY RELEASES FROM A 0.86 FT2 SPLIT RUPTURE AT 102\ POWER - Worst Pressure Case TIME BREAK FLOW ENERGY FLOW (sec) (lb/sec) (MILLION BTU/SEC) a,c

,,-P'&"""a ,;:?$ ~~ ~ Of'.t.v-~ -#/

~.tP 4t:>J/c""/&?z. ~, "97t! .~~

~C*Af'/~ .$ ~M/V~*~

TABLE 3C - MASS AND ENERGY RELEAS~S FROM A 1.4 FT2 DER AT 70\

POhER - INCLUDING ENT!1AINED MOISTUHE }:;1-'rl:.CTS Ti me Break Flow Energy tlow (sec.) (lb/sec.) (million Stu/sec.)

• a,c

~~ ~ ~ .:#c.#W'-7~--B/

~ ,/Vt7t/GM~ ~ /'I~~~

~~*#J'//?L.- ;7b t:?bP4/ ~-~~

TABLE lA 38 4 )C

8 ..

<,

TABLE 4 EFFECTS OF SINGLE FAILURES ON CONTAINMENT ANALYSES  :/..

I. MAIN STEAM ISOLATION VALVES*

Duration of Break Area Power PiEing Blowdown PiEing Blowdown Steam Mass (F' t:Z:) (%) (lb/sec) (sec) (lb)

No MSIV MSIV No MSIV MSIV Forward Reverse Failure Failure Failure Fa i lurel

!

1.4 4. 25 - 102 7047 0.136 2.532 959 17 I 840. I 1.4 4.25 70 7595 0.137 2.541 1038 19,302 1.4 4.25 30 8377 0.137 2.556 1151 21,409 1.4 4.25 0 9002 0.138 2.568 1243 23,115 4.25 1.4 102 2315 0.414 7.706 959 17,840 4.25 1.4 70 2495 0.416 7.736 1038 19,302 4.25 1.4 30 2752 0.418 7.780 1151 21,409 4.25 1.4 0 2957 0.420 7.817 1243 23,115

• Failure of a main steam line isolation valve increases the unisolatable steam line volume from 542 Ft3 to 10,083 Ft3.

II. MAIN FEED LINE ISOLATION VALVE Maximum Unisolatable Feed Line Volume = 328.2 Ft3 Without MFIV Failure Maximum Unisolatable Feed Line Volume = 868.5 Ft3 With MFIV Failure Closing Time of Feed Regulation Valve = <S.O sec.

Closing Time of Feed Isolation Valve = <30.0 sec.

TABLE 4 SHEET 1 OF 2

• TABLE 4 (Continued)

III. Auxiliary Feed System Runout Protection Failure Maximum Auxiliary Feed Flow Without = 1840 gpm Runout Protection Failure Maximum Auxiliary Feed Flow With = 2040 gpm Runout Protection Failure

.... P78 117 37/38 TABLE 4 SHEET 2 OF 2

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FIGURE 16 PEEDWATER FLOW TO THE FAULTED STEAM GENERATOR

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