ML18219D096

Forwarding Westinghouses Response to NRC Question 022.9 Re Postulated Steamline Break Containment Environmental Conditions

ML18219D096
Person / Time
Site:	Cook
Issue date:	09/20/1978
From:	Anderson T Westinghouse Electric Corp., Water Reactor Divisions
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NS-TMA-1946
Download: ML18219D096 (20)

Text

II REGULATORY INFORMATION DISTRIBUTION SYSTEM (AIDS)

DISTRIBUTION FOR INCOMING MATERIAL 50-316 EC:

DENTON I.I R NRC ORO:

ANDERSON T M

WESTINGHOUSE ELEC DOCDATE: 09/20/78 DATE RCVD: 09f27/78 COPIES RECEIVED UBJECT:

OCTYPE:

LETTER NOTARIZED:

NO ORWARDING WESTINGHOUSE" S RESPONSE.

TO NRC QUESTION 022. 9 RE POSTULATED TFAMLINE BREAK CONTAINMENT ENVIRON CONDITIONS.

LANT NAME: COOI~ " UNIT 2 DISTRIBUTION OF TIIIS MATERIAL OTES:

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SEND 3 COPIES OF ALL MATERIAL TO ISE REVIEWER INITIAL:

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Westinghouse Electric Corporation Water Reactor Divisions ttntt~slR'LKi ri"LtkFL Box 355 PlttsburghPemsylvaoia 15230 September 20, 1978 NS-TMA-1946 Mr. Harold R. Denton, Director Office of Nuclear. Reactor Regulation U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue

Bethesda, flaryland 20014

Dear Mr. Denton:

AMERICAN ELECTRIC POWER PROJECTS DONALD C.

COOK UNIT 2 (DOCKET 50-316)

Res onse to uestion 022.9 Ref;., Letter J. F.~5tolz to c

C. Eicgel dinger,

--'ay QQ., 1978>>

'1 cn~

m4P WC

="-'ir,

)

B Enclosed are forty (40) copies of a Westinghouse prepared response to NRC question 022.9 regarding postulated steamline break containment environ-mental conditions.

The attached analysis was performed using the Westing-house LOTIC 3 (Reference) code and is in fulfillment of Facility Operating License (No. DPR-74) condition Item 2.C.(3)'(g).

By a separate letter, American Electric Power Service Corporation is authorizing this submittal on their docket No. 50-316.

Very truly yours, I. C. Ratsep/bek Attachment cc:

R.

W. Jurgensen, 1L (AEP) 5A R. S. Hunter, 1L (AEP R. F. Hering, 1L (AEP S.

H. Horowitz, 1L (AEP)

J. Castresana, 1L (AEP) lA C ~~

~

~

T. M. Anderson, Manager Nuclear Safety Department

Q~tt 022.

In order to confirm that the environmental conditions used to qualify the equipment identified in question 022.8 are approp-riate describe and justify the analytical model used to conserva-tively determine the maximum containment temperature and pressure for a spectrum of postulated main steam line breaks for various reactor power levels.*

(1)

Provide single active failure analyses which specifically identify those safety grade systems and components re-

~ lied. upon to limit the mass and energy release and con-tainment pressure/temperature response.

The single failure analysis should include, but not necessarily be limited to:

main steam and connected systems isolation; feedwater, auxiliary feedwater, and connected.

systems isolation; feedwater, condensate, and auxiliary feedwater pump trip, and auxi15.ary feedwater run-out control system; the loss of or availability of'ffsite power; diesel failure when loss of offsite power is evaluated, and partial loss of containment cooling systems.

(2)

Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.

(3)

Discuss and justify the heat transfer correlation(s)

(e.g.,

Tagami, Uchida) used to calculate the heat trans-fer from the containment atmosphere to the passive heat

sinks, and provide a plot of the heat transfer co-eff3.cient versus time for the most severe steam line break accid,ent analyzed.

\\

(4.)

Specify and justify the temperature used in the cal-culation of condensing heat transfer to the passive heat sinks; i.e., specify whether the saturation temp-erature corresponding to. the partial pressure of the vapor, or the atmosphere temperature which may be superheated

'was used.

(5)

Discuss and justify the analytical model including the thermodynamic equations used to account for the removal of the condensed mass from the containment atmosphere due to condensing heat transfer to the passive heat sinks;

• NOTE:

As a result of our review. of the Westinghouse Report WCAP-83/4, "Long Term Ice Condenser Containment Code-LOTIC Code" we have concluded, that the LOTIC-I code is not acceptable for analyzing the containment long term response to secondary system pipe ruptures.

Therefore, the containment long term response to postulated main steam line breaks should be re-,evaluated.

022.9-1 Appendix 9 Unit 2 AMENDHENT 77 i)ULY, 1977

(6)

Provide a table of the peak values of containment.

atmosphere temperature and pressure for the spectrum-of break areas and power levels analyzed (7)

For the case which results in the maximum containment atmosphere temperature, graphically show the contain-ment atmosphere temperature, the containment liner temperature, and the containment concrete temperature.

as a function of tive.

Compare the calculated con-tainment atmosphere temperature response to the temp-erature profile used in the environmental qualification program for those safety related instruments and mech-anical components needed to mitigate the consequences of the assumed main. steam line break and effect safe reactor shutdown; (8)

For the case which results 'in maximum containment atmosphere

pressure, graphically show the containment pressure as a function of time; and (9)

For the cases which result in the maximum containment atmosphere pressure and temperature, provide the mass and energy release data in tabular form.

Res onse to Question 022.9

Response

is provided in the following pages.

022.9<<2 Appendix (-

Unit 2 September, 1978

RESPONSE

TO OUESTION 022.9 A.

Pi e Break Blowdowns 1.

Spectra and Assumptions A series of steamline breaks were analyzed to determine the most severe break condition for containment temperature and pressure response.,

The following assumptions were used in these analyses:

a.

Double ended pipe breaks were assumed to occur at the nozzle of one steam generator and also downstream of the flow restr ictor.

Split pipe ruptures were assumed to occur at the nozzle of one steam generator.

b.

The blowdown is assumed to be dry saturated steam.

c.

Steamline isolation is assumed complete 8.0 seconds after the break occurs for double ended breaks.

The isolation signal is generated by the steamline pressure signal from the Solid State Protection System.

The isolation time allows 5 seconds for valve closure plus 3 seconds for reach-ing the setpoint and processing the signal.

d.

Steamline isolation is assumed complete 6.5 seconds after reaching the high-high containment pressure setpoint for split pipe ruptures.

The isolation time allows 1.5 seconds for processing the signal and 5 seconds for valve closure.

e.

4.6 and 1.4 square foot double ended pipe breaks were evaluated at 102, 70, 30 and zero percent power levels.

Appendix g Unit 2 022.9-3 September, 1978

f.

Split 'ge ruptures were evaluated at 0!6 square foot 102 percent

power, 0.908 square foot 70 percent

power, 0.942 square foot 30 percent

power, and 0.4 square foot hot shut-down.

g.

Failure of a main steam isolation valve, failure of a feedwater isolation valve or main feed pump trip, and failure of auxiliary feedwater runout control were consi dered individually.

h.

The auxiliary feedwater system is manually re-aligned by the operator after 10 minutes.

2.

Break Flow Calculations a.

Steam Generator Blowdown Break flows and enthalpies from the steam generators are calculated using the Westinghouse MARVEL Code.

Blowdown mass and energy releases determined using the MARVEL Code include the effects of core power generation, main and auxiliary feedwater additions, engineered safeguards

systems, reactor coolant system thick metal heat storage, and reverse steam generator heat transfer.

b.

Steam Plant Piping Slowdown The calculated mass and energy releases include the contri-bution from the secondary steam piping.

For all ruptures, the steam piping volume blowdown begins at the time of the-break and continues until the entire piping inventory is released.

The flow rate is determined using the Moody cor-relation and the pipe cross sectional area.

Appendix g Unit 2 022. 9-4 September, 1978

3.

Sin le Fa ure Effects a.

Main Steam Isolation Valve Failure of a main steam isolation valve increases the volume of steam piping which is not isolated from the break.

When all valves operate, the piping volume capable of blowing down is located between the steam generator and the first isolation valve.

If this valve fails, the volume between the break and the isolation valves in the other steamlines including safety and relief valve headers and other connect-ing lines will feed the break.

b.

Failure of a diesel generator would result in the loss of one containment safeguards train resulting in minimum heat removal capability.

'c.

Failure of a feedwater isolation valve could only result in additional inventory in the feedwater line which would not be isolated from the steam generator.

The mass in this I'olume can flush into the steam generator and exit through the break.

The additional line volume available to flush into the steam generator is that between the feedwater isolation valve and the feedwater regulating valve, includ-ing all headers and connecting lines.

d.

Failure of the auxiliary feedwater runout control equipment could result in higher auxiliary feedwater flows entering the steam generator prior to re-alignment of the auxiliary feed system.

For cases where the runout control operates

properly, a constant auxiliary feed flow of 580 gpm was assumed.

This value was increased to 1400 gpm to simulate a

failure of the runout control.

Appendix g Nit 2 022.9-5 September, 1978

8.1 Metho'd of Analysis Following a steamline break in the lower compartment of an ice condenser

plant, two distinct analyses must be performed.

The first analysis, a short term pressure

analysis, has been per-formed with the TND computer code.

The second

analysis, a long term analysis, does not require the large number of nodes which the TMD analysis requires.

The computer code which performs this analysis is the LOTIC3 computer code.

The LOTIC3 code has been modified for the steam break analysis.

It now includes the capability to calculate the superheat condi-tions and has the abi'lity to begin calculations from time zero The major thermodynamic assumption which is used

.3;33 in the steam break analysis is complete re-evaporation of the condensate under superheated conditions for large breaks.

For the most limiting small breaks, no re-evaporation is assumed;

however, convective heat transfer as detailed in Reference f23 is used.

The version of the LOTIC3 computer code which was used to perform the steamline break analyses for D.

C.

Cook Unit 2 is the version which has been accepted for this use~

8.2 Containment Transient Calculations:

The following are the major input assumptions used in the LOTIC3 steambreak analysis for the D. C.

Cook Unit 2.

1.

Minimum safeguards are employed, e.g.,

one of two spray pumps and one of two air return fans.

2.

The air return fan is effective 10 minutes after the high-high containment pressure signal is read.

3.

A uniform distribution of steam flow into the ice bed is assumed.

Appendix g Unit 2 022.9-6 September, 1978

4.

The in~ ial conditions in the containment are a temperature of 120 F in the lower and dead-ended

volumes, a tempera-ture of 100 F in the upper

volume, and a temperature of 0

32 F in the ice condenser.

All volumes are at a pressure of 0.3 psi g.

5.

A spray pump flow of 2000 gpm is used in the upper compart-ment and 900 gpm in the lower compartment.

The spray initiation time assumed was 45 sec.

after reaching the high-high setpoint.

6.

Containment structural heat sinks as presented in FSAR Table 14.3.4-1 were used.

7.

The air return fan empties air at a rate of 40,000 cfm from the upper to the lower compartments.

8.

The material property data given in FSAR Table 14.3.4-3 was used.

9.

The mass and energy releases given in Tables 022.9-1 and 022.9-2 were used.

Since these rates are considerably less than the RCVS double-ended

breaks, and their total inte-grated energy is not sufficient to cause ice bed meltout, the containment pressure transients generated for the pre-viously presented double ended pump suction RCS break is

,considerably more severe.

10.

The heat transfer coefficients to the containment structures are based on the work of Tagami.

An explanation of their manner of application is given in References l.l,21.

Appendix g Nit 2 022.9-7 September, 1978

8.3 Results

\\

The results of the analysis are presented in Table 022.9-3.

The worst case of the double-ended steamline breaks was a 1.4 ft2 break,, occurring at 102K power with main steamline isolation valve failure.

This temperature transient is shown in Figure 022.9-1.

A temperature

study, using the minimum plant initial temperature instead of the maximum plant initial temperatures, was performed on this case.

A benefit of less than one degree was realized.

The results from the study are also presented. in Table 022.9-3.

The results from the steamline split ruptures (or small breaks) are presented in Table 022;9-4.

The worst case for these cases was a 0.942 ft small break, occurring at 30K power, with 2

failure of auxiliary feed runout protection.

A temperature transient of this case is presented in Figure 022.9-2.

A tem-perature study was performed on this case, using the minimum plant initial temperatures, instead of the maximum plant initial temperature.

A benefit of less than one degree was realized; The results from this study are also presented in Table 022.9-4.

Parameter studies have been performed varying the mass between 2.0 and 2.45 million pounds of ice.

These ice mass parameter studies have shown that the maximum containment calculated tem-peratures are not sensitive

( less than 1

F change) to these ice mass changes.

References 1.

NS-CE-1250, 10/22/76, C. Eicheldinger letter to J.

F. Stolz,

NRC, Supplemental Information on LOTIC3 questions.

2.

NS-CE-1453, 6/14/77, C. Eicheldinger letter to J.

F. Stolz,

NRC, Responses to LOTIC3 questions.

Appendix 9 Unit 2 022. 9-8 September, 1978

t 3.

NS-CE-1626, 1

/77, C. Eicheldinger letter to J.

F. Stolz,

NRC, Responses to LOTIC3 questions.

4.

J.

F. Stolz, NRC to C. Eicheldinger, 5/3/78, "Evaluation of Supple-ment to WCAP-8354 (LOTIC3) 5.

J.

F. Stolz, NRC to C. Eicheldinger, 5/10/78, "Staff Approval of LOTIC3 Code".

appendix g

thit 2 022.9-9 September, 1978

TABLE 022.9-1 4.6 FT DOUBLE ENDED BREAK 102'5'OWER WITH MAIN STEAMLINE ISOLATION VALYE FAILURE.

Forward Fl ow Time (sec)

Mass (lb/sec)

Energy (BTU/sec)--.'.,

~10GQM)1

~1000.401

~3ECGP 01

~

CCG"-+01

~

?GAL:>01

~9CCC">01

.1.CG:-~02

~15GCE>02

~ 1QQG <02

~2"=CGg<02

~ZCQGE+02

~3000K<02

~4GGGE+02

~5GOOK&2

~CGGOE+02

~?SGG 8<02

~77C"""+02 771l, +0~

.KCC".&3

&01L&3

~10GG'+04

~8201K>04

~0201@04 5379E804

~4165".<04

~3555K+0>>

~320?C40>>

~2774"=<04

~24GGi<04

~21358<04

~1M""=+04

~1748K<04

~1639K<04

~1455K<04

~1355K+04

~1201C40>>

.121G'=+04

~ 1 ~625'.>04

~Cl'-'C"'+02 iCQW '<02

~1CEC"+61

~1GGGE+01

~1ZQGK+04

~1iGQ=.+04

~1~iQ>>K~04

~1rGSE~04

~ 1cC>>~~04

~1~048 04

~1(028%4

~ 1~01 C+04

~ 1ZGGZ+0>>

~11'9JE<04

~ 119?C~04

~1196K<04

~1194K<04

~1193K+04

~1192&04

.1191 C+0>>

~1191 C<04

~ 1191 C<04

~1cGSQ04

~1~60:-<04

~12QQ=+04 Reverse flow

~1000K&1

.28?7K+04

~ 1193E+04

~2630K+02 e28??K<04

~1193K+04

~2631 K+02 e1000E&1

~1193K~04 a1000KK%

e1000Ml1

~1193K+04 4

P 1

Appendix 9 Unit 2 September, 1978

TABLE 022.9-2 0.942 FT SPLIT 30K POWER

'AITH AUXILIARY FEED RUNOUT PROTECTION FAILURE Forward Flow Ti~e (sec)

Mass lb/sec Enera~

'BTU/sec)

. 1000E-01

. 1000 E+01

. 2000E+01

.3000E~01

.5000'E+01

.6000E+01

.8000E+01

.9000E+01

.11OOE+02

.1200E+02

~ 1300F+02

.1400E+02

.1600E+02 700E+02

.1800E+02

.2000E+02

.2100E+02

.2300E+02

.2400E+02

.2500E+02

.3OCOE+02 3750E+02

.4000Ei02

.4500E+02

.55GOE+02

.6000E+02

.7000E+02

.9000E+02

.1COOE+03

.1100E+03

.1500E+03

.1700E+03

.1900E+03

.2100E+03

.2260E+03

.2261E+03

.6000E+03

.6001E+03

~ 1000E-01 2148E+04

.2124E+04

.2096E+04

.2040E+04

.2023E+04

.1976E+04,

.1956E+04

.1931E&4

.1889Ei04

.1822E+04

.1762E+04

.1659E+04

.1615E+04

.1576E+04

.1501E+04

.1464F+04,

.1397E+04

.1364E+04

.1335E+04

.1250E+04

.1132E+04

.1105E+04

.1082E+04

.1008E+04

.9810E+03

.9580E+03

.8840E+03

.8560E+03

.8310E+03

.7610E+03

.7210E+03

.6820E+03

.6030E+03

.4980E+03

.1946E+03

.1946E+03 0.

.1188E+04

~

-1189E+04

.1189E+04

.1190E+04

.1191E+04

.1191E+04

.1192E+04

.1192E+04

.1205Ei04

.1205E+04

.1205E+04

-1205E+04

.1205E+04

.1205E+04

.1205E+04

.1205E+04

.1205E+04

.1205E+04

..1205E+04

.1205E+04

.1205Ei04

.1205E+OI

.1205E+Q4

.1205E+04

.1205E>04

.1205E+04

.1205E+04

.1205E+04

.1205E+04 1205E+04

.1205E+04

.1205E+04

.1205E+04

.1205E+04

.1205E+04

~ 1205E+04

.1205E+04

.1205E+04 e Flow

. 1000E-01

.6001E+03

.1000E-01 0.

-1205E+04

.1205E+04 Appendix 9 Unit 2 September, 1978

TABLE 022.9-3 1.4 FT DOUBLE-ENDED STEAMLINE BREAKS Operating

Power, X

102 102 70 70 30 30 Aux. Feed Failure w/o w/o w/o MSIV Failure-w/o w/o w/o 0

max'16.6 314.6 317.2 314.1 317.2 313.3 me of Tmax 11.86 8.16 10.81 8.11 10.81 8.16 4.6 FT DOUBLE-ENDED STEAMLINE BREAKS Operating

Power, X

102 70 70 30 30 0

102*

Aux. Feed Failure w/o w/o w '/o w/o w/o MSIV Failure w

w/o w/o w/o 0

max'ime of Tmax, sec 319.1 316.4 313.6 316.3 313.6 311.9 308.9 319.7 4.41 4.06 3.76 3.96 3.76 4.51 4.26 5.11 4

• Temperature parameter

study, using best possible temperatures.

Mass and energy release data do not change.

Appendix g Unit 2 September, 1978

TABLE 022.9-4 STEAMLINE RUPTURES Size of Break, ft 0.86 0.86 0.908 0.908 0.942 0.942 0.4 9.942*

Operating

Power, X

102 102 70 70 30 30 Hot Shutdown 30 Aux. Feed Failure w/o w/o w/o MSIV Failure w/o w/o w/o w/o 0

max'38.0 328.0 327.8 327.9 328.2 328.2 326.5 328.1 Time of T, sec 46.92 47.92 50.46 35.46 42.51 42.08

- 59.29 37.07 "Temperature parameter

study, using best possible temperatures.

Mass and energy release data do not change.

Appendix q Unit 2 September, 1978

13370-I

%00 DOUBLE-ENDED BREAK)

I.% FT l02fo POWER)

W/0 AUX FEED FAILURE, WORST LARGE BREAK 300 U

O SPRAY TURNS ON DECK FANS TURN ON 200 LOWER COMPARTMENT UPPER COMPARTMENT IOO l00 200 300

%00 500 600 700 800 900 TIME. (SECONDS)

Figure 022.9-1 Compartment Temperature APPENDIX g UNIT 2 September, 1978

I3370-2

%00 SPLIT RUPTURE.

0-9%2 FT 30go POWER W/

AUX FEED FAILURE WORST SMALL BREAK 300 LOWER COMPARTMENT 200 I

UPPER COMPARTMENT IOO 0

0 IOO 200 300

%00 TIME (SECONDS) 500 600 Figure 022.9-2 Compartment Temperature APPENDIX Q

UNIT 2 September, 1978

4