ML17299A635
| ML17299A635 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/30/1985 |
| From: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| ANPP-33610-EEVB, NUDOCS 8510020061 | |
| Download: ML17299A635 (126) | |
Text
SUBJECT:
Forwards results of reanalysis of FSAR,Chapter 15'e containment
~ Results presented as, proposed changes to FSAR Sections 1 ' '-,4i6' '
5 6 ~ 3 ~ 1 ~ 4 ~ M ~ Changes wil'1 be incor popat d into next FSAR amend.
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ACCESSION NBR:8510020061 DOC ~ DATE: 85/g9/30 NOTARIZED:, YES DOCKET ¹ FACIL:STN"-50 528 Palo Verde Nuclear Stationi Unit ii Arizona Pub))
05000528 STN"50-529 Palo Verde Nuclear Station~
Unit 2i Arizona Publi 05000529 STN>>50-530 Palo Ve'rde Nuclear Station~
Unit 3i Arizona Publi 05000530 AUTH,NAME AUTHOR AFFILIATION VAN BRUNTiE'.E.
Ar izona Nuclear Power ProJ ect (formerly Arizona Public-Serv
'RECIP ~ NAME RECIPIENT AFFILIATION KNIGHTONiG ~ N ~
licensing Branch 3
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Arizona Nuclear Power Project P.o. BOX 52034
~
PHOENIX, ARIZONA85072-2034 Director of Nuclear Reactor Regulation Mr. George W. Knighton, Chief Licensing Branch No.
3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D. C.
20555 September 30, 1985 ANPP-33610-EEVB/KLM
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2, an'd 3
Docket Nos.
STN 50-528(License No. NPF-41)/529/530 Containment Analyses Results Chapter 15 Reanalyses File:
85-056-026'.1.01.10
Reference:
Letter to G.
W. Knighton, NRC, from E. E.
Van Brunt, Jr.,
ANPP, dated April 15, 1985 (ANPP-32401);
Subject:
Revised Chapter 15 Analyses
Dear Mr. Knighton:
The attached supplies the.results of the following analyses:
FSAR Section 6.2 6.2 Event CTMT LOCA DEDLS 9.82 FT2 Max ECCS CTMT SLB 102% Power 8.78 FT2 Slot with Cooling Failure 6.2 CTMT SLB 75X Power 8.78 FT2 Slot with Cooling Failure 6.2 CTMT SLB 50X Power 8.78 FT2 Slot with Cooling Failure 6.2 6.2 CTMT SLB 25X Power 8.78 FT2 Slot with Cooling Failure CTMT SLB OX Power 4.0 FT2 Slot with Cooling Failure ANPP committed to the reanalysis and submittal of these events in the referenced letter.
8510020061 850~0 PDR ADOCK 050oo528 P
0
/
P~
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G.
W. Knighton, Chief Containment Analyses Resu~
'- Chapter 15 Reanalyses ANPP-33610 Page 2
The results are presented as proposed changes to FSAR Sections 1.9.2.4, 6.2.1 and 6.3.1.4.M and are all within NRC acceptance criteria.
Upon NRC approval, these FSAR changes will be incorporated into the next PVNGS FSAR Amendment.
The analyses were performed with the following parameter changes:
Pump Flow Rate.
A lower CS pump flow rate of 3525 gpm, instead of 3740 gpm, was utilized to allow for additional pump operating margin over pump life.
Main Steam Line Break (MSLB) Blowdown Data This data was revised due to the use of PVNGS specific as-built MSL pipe I.D. of 25.5 in. rather than 28 in. which is used in CESSAR.
Using the 25.5 in. diameter reduces the maximum break size from 8.78 ft'o 7.16 ft.
The reduced diameter also decreases the size of the OX power MSLB from 4.0 ft to 3.0 ft.
The 3.0 ft'reak is now the maximum break size which does not result in mixed flow out the break.
Please contact Mr.
W. F. Quinn, of,,my staff, if you should have any questions on this matter.
Very:truly
- ours, E, E, Van Brunt, Jr.
Executive Vice Presient Project Director EEVB/KLM/slh Attachment cc:
E. A. Licitra R. P.
Zimmerman A. C. Gehr
0 l~
STATE OF ARIZONA
)
) ss.
COUNTY OF MARICOPA)
I, Edwin E.
Van Brunt, Jr.,
represent that I am Executive Vice President, Arizona Nuclear Power Project, that the foregoing document has been signed by me on behalf of Arizona Public Service Company with full authority to do so, that I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true.
Edwin E.
Van Brunt, Jr.
Sworn to before me thie~Odey of
, 1985.
Notary Public My Commission Expires:
M Commission Expires April 6, 1987
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"P 1.9.2.4.1 STANDARD DES SIGNS Containment.-Spray System (CESSAR Appendix 6A, Section 7.14/FSAR section 6.5.2.8 (RA) 7.1.4)
CSS actuation and flow delivery occurs on a pre-established schedule given in Section 6.3 of CESSAR Appendix 6A.
This schedule determines flow conditions in various analyses that take credit for a train of the CSS functioning to mitigate the H4x>Mv M consequences of postulated transients.
The~time to establish rated flow given in CESSAR Appendix 6A, Section 7.1.4, is 58 seconds after receipt of CSAS.I, PVNGS~CSS rated flow will pot 6 RgA'TGg 'TswAQ 68 56<>~DQ T~iE be delivered in The exception to the CESSAR interface requirement for the CSS flow delivery time is only significant in the determination of the in-containment equipment qualification temperature parameters.
- However, under
~Mph&'vRp'.i4 the worst caseI,conditions (i.e. Main Steam line 4-.Mft slot break at 102% power, loss of one containment spray train, and no loss of offsite power), the calculated temperature transient of section 6.2.1.8 is bounded by the equipment qualification environment of table 3E-1 (sheet 1 of 7).
C-E has also committed to qualify safety-related equipment within their scope to the same environment.
By this analysis, the delay in delivery of rated flow to the CSS has been shown to have no adverse con-sequences under postulated transients.
1.9.2.4.2 Containment Spray System (CESSAR Appendix 6A, Section 7.13.14/FSAR section 6.5.2.8 (RA) 7.13.14)
The head loss requirements of CESSAR Appendix 6A are not applicable to the PVNGS design due to the addition of the auxiliary spray headers.
The analyses of both train A and train B of the containment spray system have shown that either train of the system will operate satisfactorily and meet the design flow requirements independent of the other train during all modes of operation.
The design flow includes the additional flow for auxiliary spray headers.
December 1980 1.9-9 Amendment 3
/ p
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-1 POSTLLATZD AC IDEN.S FOR CONTAINMENT DESIGN (Sheet 1 of 2)
C"ntainment Design Parameter Peak pressure/temperature r
P s,,p, Postulated Accidents Considered Loss-of-coolant accidents (LOCA)
Double-ended hot leg slot (DEHLS)g 19 '4 ft2 area Double-ended suction leg slot (DESLS),
9.82 ft
- area, Maximum ECCS flow Double-ended sue~ion leg slot (DESLS),
9.82 ft
- area, Minimum ECCS flow Double-ended discharge leg slot (DEDLS), 9. 82 ft
- area, Maxim m ECCS flow Double-ended discharge leg slot (DFDLS), 9. 82 ft
- area, Minimum ECCS flow Main steam line breaks (MSLB)
V-b ft slot area MSLB, 102$
power with cooling <<.ra'
='"ilure ft slot area MS'B, 75%
power with cooling train failure 7-I ft slot area MSLB, 508 power with cooling train failure 70 2
ft slot area
- MSLB, 25%
power with cooling t"ain failureft slot area MSLB, 0.0%
power with cooling train failure I+1e
- 6. 2. l-3
I l ~
PVNGS F SAR CONTAINMENT SYSTEMS Table 6.2.1-2 CALCULATED VALUES FOR CONTAINMENT PRESSURE PARAMETERS Parameter Design B~gje Calculated Value (psid)
Peak pressure Peak subcompartment pressure DEDL slot, Max.
ECCS Reactor cavity 35P in 2
DLG 99.8 Steam generator compartment 592 in.2 SLG 29.4 External pressure loading Pressurizer compartment wall 161 in.2 surge line guillotine Inadvertent operation of the containment spray system 72.4 2.6 lpga a.
See table 6.2.1-1 for definition of abbreviations used.
February 1985
- 6. 2. 1-5 Amendment 14
I ~
Pi
PVNGS FSAR CONTAINMENT SYSTEMS the design pressure (60 psig).end the 'caaaulated peak pressure
/
of the as-constructed design~(+&~ psig) kesults in a design margin of approximately 20%. 4 6.2.1.1.1.2 Containment Internal Structures Accident Conditions.
The RCS breaks defined in CESSAR Section 3.6.2 are analyzed and form the design basis for the loads on containment internal structures and equipment.
The simultaneous occurrences are the same as discussed in section 6.2.1.1.1.1.
The design pressures for the reactor cavity, steam generator and pressurizer compartments are given in table 6.2.1-3.
The cu rently calculated values for the as-built design are listed in table 6.2.1-2.
n-6.2.1.1.2 Design Features The design features of the containment structure and internal structures are provided in sections 3.8.1 and 3.8.3, respec-tively.
6.2,1.1.2.1 Protection From The D namic Effects of Postulated Accidents.
The containment structure, subcompartments, and
~
~ sw'.res
~ a.ii4 engineered safety feature systems are protected from loss of safety function due to the dynamic effects of postulated accidents.
Containment design has provided separation and inclusion of barriers, restraints and supports when required to protect, essential structures,
- systems, and components from accident generated missile, pipe whip, and jet impingement forces.
The detailed criteria, locations, and descriptions of devices used for protection are given in section 3.6.
September 1980 6.2.1-7 Amendment 2
PVNGS FSAR CONTAINMENT SYSTEMS
-.able 6.2.1-4 MASS AND ENERGY RELEASE FOR CONTAINMENT PEAK PRESSURE/TEHPERATVRE ANALYSIS (Sheet 1 of 4)
A.
Most severe hot leg break Break type:
Double-ended hot leg slot break Pipe ID, 42 in.
Break area 19.24 ft Accident Phase Time (s)
Mass Release Rate (ibm/s)
Energy Flow Rate Million Btu/s Enthalpy (Btu/ 1 bm)
Reactor Vessel Pressure psia Blowdown Refer to CESSAR Table 6.2.1-10 a.
70 psia containment backpressure case b.
Recirculation begins c.
Safety injection realigned; 50% to hot legs and 50'4 to cold legs.
Excess hot leg safety injection water carries significant decay and sensible heat to containment sump.
d.
RCS sensible heat addition to containment completed.
e.
Includes direct floor spillage for discharge leg break.
f.
r a+st to table 6.2.1-for PVN specific mass/
energ lease data used for equipment qua ication peak temperature n
sis in section 6.2.1.8.
February 1985
- 6. 2. 1-13 Amendment 14
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PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-4 MASS AND ENERGY RELEASE FOR CONTAINMENT PEAK PRESSURE/TEMPERATURE ANALYSIS (Sheet 4 of 4) pppgpcf NI7H ]P/SF<7 R D.
Most severe secondary system break (pressure)
Break type:
Main steam line slot break 25 5
Pipe ID: ~ in.
3 db Break area: ~ ft Reactor power level:
0%
Time (s)
Mas~elease Rate.-
(1bm/4)~
Enthalpy (Btu/ibm)
Energy Release
'Rate (million Btu/s)
Refer to CESSAR Tab
.1-19 E.
Most severe secondary system break (temperature)
Break type:
Main steam line slot break 2S.S Pipe ID: ~ in.
P./O 2
Break area: ~ ft Reactor power level:
102%
Time (s)
Mass.
ase Rate
.(ibm/s)
,.-Energy Release Enthalpy Rate tu/ibm),
(mi 1 1ion. Btu/s )
14I Refer to CEN'.&R Table gzpl~cg
~luau
/~
Amendment 14 6.2.1-16, February 1985
TllAe (s)
Ma ss Re 1 ea se Rate
( ibm/s)
Enthalpy (Btu/ibm)
Energy Release Rate (million Btu/s) 0.0 0.2 O.C o.6 0.8 1.0 2.0 3.0 4,0 5.0'.0 7.0 8.0 9.0 IC.O II.0 12.0 13.0 14.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 130.0 140.0 150.0 160.0 170.O 180.0 l90.0 SN.O 857 I.5 7169.9 7059.9 7040.0 655' 6536.3 6715.7 6535.0 6040 0 4854.6 4755.9 4655.6 455C.C 4528.6 26:.3.2 2568.8 25 7.0 2507 6
2475.2 2452.5 2306.6 2173.7 20'76.4 1995.9 1928.1 1868.1 1813. 7 1715.1 1625.3 1539.9 1460.6 1383.9 1310.9 123C.3 1156.2 1072.0 971.4 831.6 624.9 354.0 25.0 3.0 1185.0 1185.9 1186.4 Ile:.8 1187 2
1187.5 IIS8.9
'1190.0 1151.1 1152. I 1152.9 II93.2 1153.S 1194.4 119'.7 1195.1 1155.5 1196.0 1196.4 1196.7 1198.3 1199.3 120". 6 1201.4 1201.7 1202.3 1202. 5 1203.4 1203.5 1204.0 1204.3 1204.6 1204.6 1204.7 1204.8 1204.3 12C3.4 1202.1 1199.4 1241.5 1262.4 1261.3 10.640 8.503 8.423 8.35-
$.294 8.237 7.965 7,777 7.585 5,7.";
5.573 5.572 5'85 5.40:
3.110 3.070 3.0 3
2,955 2.965 2.935 2.764 2.6C7 2.493 2.399 2.317 2.245 2.181 2.064 1.95:
1.85C 1.755 1.667 1.579
- l. 487 1.393 1.251 1.169 0.9997 0.7489 0.4395 0.03156 0.003784
t
Time (s)
Mass Release '-
Rate (ibm/s)
Enthalpy (Btu/ibm)
Energy Release Rate C.
(million Btu/s) 0.0 0.2 0.4 0.6 0.8 1.0 2.0 3.0 4.0 5.0'.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 14.0 15.0 20.0 II.o 40.0 45.0 50.0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 130.0 14tI.O 150.0 160.0 170.0 180.0 190.0 200.0
~ o 9858.0 7189.5 7031.2 6901.0 6790.8 6695.4 6350.4 6118.6 5938.2 4532.2 4593.1 4635.6 4666.8 4692.9 2608.8 2622.8 2633.5 2637.8 2633.2 2636.7 2524.9 2408.1 2321.0 2225.7 2146.5 2078.3 2009.6 1882.3 1764.4 1649.8 1509.2 1341.4 1182.3 1105.7 10¹5 867.3 707.1 531.2 383.9 193.9 6.6 3.8 1189.9 1190.8 119I.5 1192.1 1192.6 1193.1 1194.6 1195.7 1196.5 1195.9 1195.5 1195.3 1195.3 1195.2 1194.6 1194.5 1194.6 1194.6 1194.4 1194.3 1195.7
" 1197.2 1198.2 1199.2 1199.6
~. 1200.5
','201.2 1202.3 1202.7 1203.8 1203.9 1204.7
-. 1204.4 1204.7 1203.4 1202.6 1200.1
'196.2 1190.2 1242.9 1289.4 1292.1 4
~,'.
11.73 8.561 8.378 8.227 8.099 7.988 7.586 7.316 7.105 5.420 5.491 5.541 5.578 5.609 3.117 3.133 3.146 3.151 3.145 3.149 3.019 2.883
. 2.781 2.669 2.575 2.495 2.414 2.263 2.122 1.986 1.817 1.616 1.424 1.332 1.204 1.043 0.8486 0.6354 0.4569 0.2410 0.00851 0.00491
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-7 ENGINEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS (Sheet 1 of 2)
System/Item Full Capacity Value Used for Peak Pressure Analyses Passive Safety Injection System (a)
Number of safety injection tanks Pressure
- setpoint, psig Volume, ft /tank 3
Active Safety Injection Systems High-pres~use safety injection a
Number of lines Number of pumps HM Flowrate, gal/min/pump Low-presser~
safety injection a
Number of lines Number of pumps HAX Flowrate, gal/min/pump Containment Spray System Number of lines Number of pumps Number of headers (c)
- Flowrate, gal/min/pump gjPN 1'0rated flow, seconds AFTCP ~
(Offsite Power Available/
Loss of coolant'accident Main steam line break accident 600 1927 1130 5000 vT IN tv'j Ir, Ti 600 1927 2
1/2(b) 1130 2
1/2 5000 8i Ri so(eq February 1985 6.2. 1-19 Amendment 14
4 e
I PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-7 ENGINEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS (Sheet 2 of 2)
System/Item Full Capacity Value Used for Peak Pressure Analyses Heat Exchangers Shutdown heat exchangers Type Number 2
Heat transfer area, ft Overall heat transfer coefficient, Btu/h-ft -'F Flowrates:
Recirculation side, gal/min Exterior side, gal/min Source of cooling water Shell and U-Tube 10,840 372 14,400 Essential cooling water Shell and U-Tube
'10,840 285 14,400 Essential cooling water a.
From CESSAR Section 6.3 b.
1
= minimum ECCS; 2
= maximum ECCS c.
- Net flow during injection phase; flow increases to gal/min following closure of miniflow bypass. valve on. recirculation actuation signal (RAS).
.. both maximum and.minimum ECCS performances were evalu-ated.
For the containment heat removal
- systems, mini-
. mum system capacity is conservative, for calculating containment peak pressures.
Passive heat sink data is provided in table 6.2.1-8.
Part A of the heat sink table is a detailed list of the geometry of each heat sink and part B describes Amendment 14 6.2.1-20 February 1985
PVNGS FSAR CONTAINMENT SYSTEMS components would be cooler than the design maximum temperature; a few regions, notably the reactor cavity walls, may be warmer.
The warmer heat sinks comprise a small area
- and, as a result, the average heat sink temperatures would initially be less than the design maximum.
Minimum heat sink areas (i.e., nominal minus tolerance given in table 6.2. 1-8 ) are used for the short-term and long-term containment pressure-temperature transient analysis of primary or secondary pipe ruptures inside containment.
Most individual structural or component heat sinks have some uncertainty associated with the exposed surface area.'he minimum area was used to provide the most conservative assumptions.
Table 6.2.1-8 part C, lists the thermophysical properties used in analyses.
Metal, concrete, and protective coating properties are typical values for the temperature range observed.
Blowdown mass and energy release rates for LOCA ~ ]gg IR d'
Jt S.
1 d
d th p
kp computer code is presented for the most severe break area at each break location in table 6.2.1-4.
Assumptions made regarding closure times of secondary system isolation valves and single active failures for MSLB analyses are discussed in section 6.2.1.4.
C.
Methods of Analysis The containment pressure analyses are performed using the Bechtel COPATTA computer program that was derived from the CONTEMPT program written for the AEC Loss-of-Fluid Test (LOFT) program.
H$4$ rrlasi larry dre 4iseussi/
I< strr<<
6.2.1-29
I
'VNGS FSAR CONTAINMENT SYSTEMS transferred between the liquid and vapor regions. by boiling, condensation, or liquid dropout.
Evaporation is not considered.
A convective heat transfer l
coefficient can be specified between the sump liquid and atmosphere vapor regions.
However, since any heat transfer in this mode is small, a conservative coefficient of zero is generally assumed.
Each region is assumed homogeneous, but a temperature difference A
can exist between regions.
Any moisture condensed in the vapor region during a time increment is assumed to fall immediately into the liquid region.
Non-condensible gases are included in the vapor region.
r
~
a Accident Identification and Results The containment pressure and temperature response and sump water temperature response versus time are given in figures 6.2.1-1 through 6.2.1-6 for the most severe LOCA breaks and the most severe MSL breaks.
It has been demonstrated in reference 1 where main steam line breaks produced a high degree of superheat that typi-cal safety-related equipment surface temperatures remained at.>>-or near the containment, 'saturation tempera-ure as a result of the short time frame at superheat conditions due to spray actuation."""
Pipe break locations, break areas, peak pressures and
\\
temperatures,"times of peak pressure, and total energy r'eleased'to containment are summarized i.n table 6.2.1-9 for each LOCA and MSLB analyzed.
Based on the results presented in these tables, the double ended discharge leg slot break, LOCA with maximum ECCS was identified a t e pipe break with the highest peak pressure
(
psig) which is below the design pressure value of 60 psig.
f'<~ ~
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v t.
~
b l'.2.1-31 4
Table 6.2.1-9
SUMMARY
OF CALCULATED CONTAINMENT PRESSURE AND TEMPERATURES Loss of Coolant Accident Results Peak pressure, psig Peak temperature, P
Time of peak pres-sure g seconds Energy released to containment up to the end of bio<<-
down i 106 Btu Main Stream Line Brcak Results Peak pressure, psig Peak temperature, P
Time of peak pres-
- sure, seconds Energy released to containment up to the end of blow-
- down, 106 Btu DEHLS 19.24 ft 45.1 273 8.6 376.00
- g. lb 102l Power 8.7 ft2 slot with Cooling Pailure M.O DESLS2 9.82 ft Max ECCS 46.3 298 87 378.10 Sl Power
.78 ft2 slot w
Cooling
- Failure, W~RR
~(5
~ le.G 351 7 DESLS2 9.82 ft Hin ECCS 46.5 300 87 378.10 Ol Power 8.78 ft2 slot Cooling Failure
~6+/,O DEDLS2 9.82 Hax ECCS
~23 I
382.90 Power 8.78 ft2 slot Cooling Failure
~ago DEDLS2 9.82 Hin ECCS 45.1 297 500 382.90 Ol Power 4.0 ft2 slot
<< th Cooling Failure
~%5 3
~LROe 5 J
34)l. 4
I C,
CONTAINMENT SYSTEMS Figures 6.2.1-7 through 6.2.1-9 are plots of the con-tainment condensing heat transfer coefficient versus time for the most severe RCS discharge and suction leg breaks and secondary coolant system breaks.
6.2.1.1.3.2 Lon -Term Containment Performance.
Long-term analyses of the worst, case pump discharge leg break, and the worst case pump suction leg break were performed to verify the ability of the containment heat removal system (CHRS) to main-tain the containment below the design conditions.
These evalu-I ations were based upon conservatively assumed performance of the engineered safety features.
The CHRS long-term operating mode is assumed to include one containment spray train.
The containment pressure-time responses for the DBA for the pump discharge leg and the pump 'suction leg cases out. to 10 seconds (11.6 days) are shown in figures 6.2.1-3 and 6
6.2.1-4 for the ESF,performance mode outlined in table 6.2.1-7.
The containment pressure-time response for the highest pressure
~
~
~
~
~
MSLB case (0% power) is shown in fig-VhiS
~
3 ure 6.2.1-6.
The-MS'nalysis shows that by 10 seconds the containment pressure is reduced to~psig which indicates that
,'. the containment pressure. will be reduced below 50% of the peak pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
No further analysis is performed
-for the MSLB since after'solation and blowdown there is no further energy input to containment.
The maximum pressure of 4//i8 42~ psig occurs at ~i seconds in the 0% power MSLB case.
The IQC'5 energy..distributions, in -the containment--versus time are shown in figures 6.2.1-10 through 6.2.1-12 for DBA LOCAs and MSLBs.
Mechanisms of energy removal from and transfer within the containment are addressed.
Included are the vapor energy (steam plus air),
sump (liquid) energy, energy contained in heat sinks, energy removed by the shutdown cooling heat exchangers, and energy transferred by sprays from the vapor to the sump.
6.2.1-33
r, r,
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 3 of 7)
C.
Worst case discharge leg break Break type:
Double-ended discharge leg slot break Maximum ECCS Time (s)
Event 0.0 14.4
- 19. 3
- 19. 3 217 Break occurs Start core flood tank injection Peak containment pressure (blowdown), 43.0 psig Start ECCS injection phase End blowdown Start spray injection End of core reflood Peak containment pressure subsequent to end 500 2I 1 7 0 ll7 of blowdown, psig 9g.5 End of steam generator energy release:
reflood
'nd of ECCS injection; start ECCS recirculation End of spray injection; start spray recirculation post
'57&o 508k) 5400 86400 10 Containment pressure at 30 psig (one-half 60 psig design value)
ECCS realigned:
50t to hot leg and 50% to cold leg Addition of RCS sensible heat completed Depressurization of containment, 2.1 psig February 1985 6.2.1-37 Amendment 14
C~
PVNGS FSAR CONTAINMKNT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 4 of 7)
D.
Norst case main steam line break (pressure)
Break type:
04 power MSL slot break Loss of cooling train Time (s)
Event 0.0 2.95
/',S5 2 ~5 4.b5.
2.95 4 k-5 3.g5 h
h Break occurs Containment pressure reaches safety injection actuation signal (SIAS) analysis setpoint of 5 psig SIAS generated Containment pressure reaches reactor trip analysis setpoint of A-psig(a)
Containment pressure reaches main steam isola-tion si nal (MSIS) analysis setpoint of 6"psig h
r High containment pressure reactor trip signal and MS IS generated~l~l Turbine admission valves closed Reactor trip breakers openCbb Containment pressure reaches containment spray actuation signal (CSAS) analysis setpoint of 10 psig
~h'EH w S.o CSAS generated a.
Value used in pipe break analysis.
b.
Events based on time (4.55 seconds) to reach reactor trip and MSIS setpoint of 6 psig used in pipe break blowdown analysis.,
Amendment 14 6.2.1-38 February 1985
r
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 5 of 7)
DE Worst case main steam line break (pressure)
Break type:
0% power MSL slot break Loss of cooling train (cont)
Time (s)
Event
+B-.M-Main steam isolation valves closed~
Main feedwater isolation valves closed (see section 1.9.2.4.10) 8v
/Bc.~
~e~
Containment spray pump at full speed Containment spray at full flow initiated inside containment buildin
. g 373.z i=
peak containment temperature of ~occurs Peak containment pressure of ~~psig occurs 4/f, c Blowdown ends February 1985 6.2. 1-38A Amendment 14
r
.1'YNGS CESAR CONTAINMENT SYSTEMS.
Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 6 of 7)
E.
Worst case main steam line break (temperature)
Break type:
102% power MSL slot break Loss of cooling train Time (s)
Event 0.0 S. O5 4, o5.--
~O9. 05 Break occurs Containment pressure reaches safety injection actuation signal (SIAS) analysis setpoint of i ~
a 5 psig SEAS generated Containment pressure reaches reactor trip analysis setpoint of 8'si P.05 d,OS
. 4.ao 4
$4.
4
~ 2.0 1
a 0 Containment. pressure reaches main steam isolation signal (MSIS) analysis setpoint of
~ ~-sD~.
High containment pressure reactor trip signal and MSIS generated~(
Turbine admission valves closed~(
Reactor trip breakers open b
~ Containment pressure reaches containment spray actuation signal (CSAS) analysis setpoint of 10 psig.
- a. 'Value used in pipe break blowdown analysis.
b.
Events based on time -(3.8 seconds) 'to reach reactor trip and MSIS setpoint of 6 psig used in pipe break blowdown analysis.
Amendment 14 6.2.1-38B February 1985
I
/ ~
PYNGS k SAR CONTAINMENT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 7 of 7)
E.
Worst case main steam line break (temperature)
Break type:
102% power MSL slot break Loss of cooling train (cont)
Time (s)
Event
~8,o
- 8. 45 is.45 CSAS generated Main steam isolation valves closed Main feedwater isolation valves closed (see section 1.9.2.4.10)
Containment spray pump at full speed Containment spray at full flow initiated inside containment building 3RS.5 F
Peak containment temperature of~~ occurs 4 I. 0 Peak containment pressure of~ psig occurs
~ %00 Blowdown ends February 1985 6 ~ 2 ~ 1-39 Amendment 14
C ~
r
'able 6.2.1-11 (a)
REACTOR CONTAINMENT BUILDING ENERGY DISTRIBUTION (Sheet 6 of 11)
C'..
Worst case-discharge leg break Break type:
Double ended discharge leg slot break, maximum ECCS Break area:
- 9. 82 ft2 Energy (10 Btu) 6 Energy Description Reactor coolant system water internal energy Safety injection tank water internhl energy Energy stored in core Energy stored in RV internals E
~
Prior to LOCh 357.992 41.846 30.299 37.386 At Peak, Pressure Prior to End of:
Slowdown 18.888 29.207.
13.923.
- 34. 959 End of B lowdown 14.243 24.869 13.741 33.948 At Peak Pressure After End of Blowdown 3~. SOi, I 659
- 6. 2/IQ
'2.l pic End of Core Re flood
- 40. 385 0.0 6.760
- 28.811 End of Post Reflood
- 31. 239 0.0
- 5. 142 23.550 24 Hours After LOCA 186400 01 26 ~ 09 0.0 0.0 0.0 Energy stored in RV walls Energy generated during shutdown from decay heat Energy stored in pres-
- suriser, primary piping,
- valves, and pumps Energy stored in steam generator tubes 88.602 124.294 32.180 88.531.
5.188 123.283 29.669 88.400 5.954 III 122.810 29.815 06-.498'Q.A1
+3++8'3
'lS.SSQ 86.937 36.621 1
4e91rkee 122. 816 10'1. 2:31 44
. 17.849 it.9~q, 84.355 "70.364 100.178 15.445 0.0 2955.30 0.0 0.0
l P
- Table 6. 2. 1-11 REACTOR CONTAINMENT BUILDING ENERGY DISTRIBUTION (Sheet 7 of 1 1 )
C.
Norst case discharge leg break Break type:
Double ended discharge leg slot break, maximum ECCS Break area:
9.82 ft2 Energy (10 Btu) 6 Energy Description Prior to LOCh At Peak Pressure Prior to End of 8 lowdown End of Blowdown At Peak Pressure After End of B lowdown End of Core Ref 1ood End of Post Reflood 24 Hours After LOCA (86400 s)
Ch lV I
I cn Energy stored in steam generator secondary walls 153.834 153.834 Secondary coolant internal energy.(in steam generators) 263.030 274.684 Energy content of RCB atmosphere(b)
Energy content of RCB(
internal structures Energy content of recirculation intake water (sump) 6.72 0.0 0.0 276.10 27 55 83 ~ 47 349.55 Energy content of RWST water 349.55 153.834 274.084 270.72 36.08 95.11 349.55
/4/. 999
/7$
J/6 /o
~ //g.oo
+7. ex) gg/, 9o 30+s~
144.963 182.987 894.37 SZ.43
>5a(
)
133.323 163 F 436 Iai. 8 029.
~9')V SS 0.0 0.0 9o 19 l/PPI QL/
0.0 Energy removed by shutdown heat exchangers Nh Nh Nh Nh Nh NA arse.
74
~ ~
Table 6.2.1-11 REACTOR CONTAINME BUILDING ENERGY DISTRIBUTION (Sheet 0 of 11)
(a)
D.
worst case sec ndary ystem brcak (containment temperature)
Break type:
102.0'4 power main stcam linc break Break area:
- 8. 70 ft4 Energy (10 Btu) 6
'nergy Desciiption Rea~:i or coolant system water internal e>>ergy Safety= injection tank water internal energy Energy" stored in core Energy stored in RV internals-Energy stored in RV metal Energy generated during shutdown from decay heat Energy stored in pressurizer, primary pipIng, valves, and pumps Energy stored in steam generator tubes Prior to MSLB 357. 992 NA 30.299 NA NA
- 0. 000 250.202 32.100 At Peak Pressure Prior to End of Blowdown 2F< P<o NA NA t
SO.870
~&75 PIZ.IFE End of Blowdown'A Ii.GG1
]J
%6 NA NA c0.5'f<
~3+8-Zl4-PS)
Table 6. 2. 1-11 REACTOR CONTAINN NT BUILDING ENERGY DiSTRIBUTION (Sheet 9 of 11)
(a)
D.
Worst case s condary syst: em break (conta'inment temperature)
Break type:
102.0% power main steam line break Break area:
8.70 ft~
Energy (10 Btu) 6
(
(
~ ~
J3 Energy Description I:>ieryy s tored in steam genera tor
- .ondas y walls
.:.:.;i.>>iIary coolant internal energy (i n
.;i*~am generators) 1((er.yy content of RCB atmosphere (b)
H( ei.>y content of RCB internal uctures(c) content of recirculation intake (sump)
I::."i(i,;~ content of RWST water Feeil<(ater to steam generator 1
Feedwater to steam generator 2
Prior to MSLB 153. 834 263. 030 6.73 0.0
- 0. 00 349. 55 0.0 0.0 At Peak Pressure Prior to End of Blowdown I~~H
/0( /nf 0
/2S /LE 3
7S-Cg u-A.
2 3'0 s. go 3'~2 End of Blowdown
/4g.~~
(gg. o sz 2g-)$
~
~
~
~
~
~
~
0 ~
~ ~
~
,c
~
~
~
~
~
~
~
~
~
~
~ t
t
3.0 REACTOR CONTAINNEN E.
Worst case s
Break type:
Break area:
Table 6.2.1-11 BUILDING ENERGY DISTRIBUTION (Sheet 11 of 11)
(a) condary system break (containment pressure) 0%
ower main steam line break
~ 0 ft2 Energy (10 Btu) 6 Energy Description Energy stored in steam generator secondary walls Secondary coolant internal energy (in steam generators)
(b)
Energy content of RCB atmosphere
~ Energy content of RCB internal structures(c)
Energy content of recirculation intake water (sump)
Energy content of RWST water Feedwat.';"
to 'team generator l Feedw".I.'.n steam generator 2
Prior to MSLB 157. 441 369. 750 6.73 0.0 0.0 349.55 0.0 0.0 At Peak Pressure Prior to End of Blowdown 13.~l laird 2'fS'b-48 2Z 8 0.0 u
Aran Fnd of Blowdown t57.574 2
>60 3s
~
4 gt-7~
89~
27.71 re~ bg 3
0.0 u 477
I
'r
- CONTAINMENT SYSTE.'".5 Table 6.2.1-25 (a)
MISCELLANEOUS AND ADDITIONAL ENERGY RELEASES Time (sec)
Energy (Btu/h) 0 500 500 86,400 86,400 20.8956 E6 20.8956 E6 0
1E6 a.
All miscellaneous energy added after blowdown ends (500 sec),
up to 1 day (86.400 sec).
4 Secondar S stem Pi e
Ru tures Inside Containment Refer to ESSAR Section 6.2.1.4...
The MSLBs.analyzed for con-C tainment pre re/temperature were the slot breaks listed in CESSAR Tables 6.2.
11 th ough 6;2.1-20.
The guillotine ruptures for the same s
e break and power levels were evaluated as -being 'less severe than th omparable slot breaks.
In addition.
PVNGS-specific mass/ene release data for the.
102'4 power MSL slot break with a 9.6 cond feedwater isola-tion valve closure time used for the in-co ainment equipment qualification peak temperature analysis are pr ded in table 6.2.1-25A.
1 Erg(s
~,y/
instrfc"'
Februa=:~
198=
6.2.1-96I
'mendment 14
INSERT "C" The MSLB methodology used for the PVNGS analysis is the same as CESSAR Section 6.2.1.4 except as follows.
The PVNGS containment free volume of 2.6 million cubic feet was used in the calculation of the time to reactor trip and MSIS on containment high pressure rather than the 3.7 million cubic feet used in CESSAR.
The PVNGS specific steam line arrangement was used for determining flow from the intact steam generator to the containment prior to isolation.
The steam line model considered choking at the MSIV throat and cross connect piping as well as at the venturi throat.
The NSIV and MFIV closure times were 4.6 and 9.6 seconds respectively.
The Palo Verde specific mass and energy release data for the most severe MSLB containment pressure and temperature cases are given in table 6.2.1-4.
II lt t
PVNGS FSAR Table 6.2.1-25A CONTAINMENT SYSTEMS MASS AND ENERGY RELEASE FOR CONTAINMENT EQUIPMENT
'UALIFICATIONPEAK TEMPERATURE ANALYSIS (Sheet 1 of 2
Break type:
Pipe ID:
Break area:
Reactor power level:
Isolation valve closure Main steam line slo 28 in.
8.78 ft2
" 102%
times:
4.6 sec MSIV
~ 9.6 sec MFIV break Ti me (s)
$ E 0 '
0.2 0.6 1.0 2.0 3.0 4 ~ 0 5.0 6.0 7.0 8 '
9.0 9 ~ 5 10.0 10.5 11.0 11.5 12
.5 13.0 15.0 20.0 Mass Release Rate (ibm/s) 9833.75 7303.17 7039
~ 74 6849.83 6524.69 6301'. 20 6127.74 4660.96 4759.56 4843.
4 492
~ 50 4
8.01 016 '7 2853.53 2864.79 2870.74 2871.49 2867
~ 25 2858 '5 2844.85 2767.35 2559. 88 Enthalpy (Btu/ibm) 1190 '$~
1190.19 119$
~ 8 7 1 42.71 194. 18 1195 '1 1195'. 99 1195.40 1195.02 1194.57 1194.13 1193.72 1193.54 1192 '2 1192.77 1192.67, 1192. 65 1192.68 1192.77 1192.91 1193.77 1196.08 Energy Release Rate (million Btu/s) 11.702518 8.695786 8.390426 8.169867 7 '91653 7 '31249 7 '28734 5.571987 5 '87764 5.786444 5.875717 5 '54272 5 '87483 3.404022 3-417024
~, 3 '23858
.424678 3.419704 3.409/35 3.393662.
3.303575 3.061811 hmonRmont 18
I
~)pig Table 6.2.1-25A CONTAINMENT SYSTEMS MASS AND ENERGY RELEASE FOR CONTAINMENT EQUIPMENT QUALIFICATION PEAK TEMPERATURE ANALYSIS (Sheet 2 o f
)
ime 25.0 30.0 X 35.0 40
~ 0 45.0 50
~ 0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 130.0 140.0 150.0 155.0 160.0 165.0 170.0 180.0 185.0 190grf) 1 4 ~ 5 Mass Release Ra te (ibm/s )
2445. 21 2374.74 2284
~ 15 2206.28 2142.79 2075 '2 1948.28 1828.22 1720.88 1610 '6 1492.90 1361.07 1249 '1/
1177.
6 1044.42 9'70
~ 36
~~903
~ 62 830.21 751.09
/
676
~ 99 512.9 380 '
261.4 Enthalpy (Btu/ibm) 1197
~ 29 1198 F 00 1198 '0 1199 '4 1200.2 1200 80 12
.84 02.70 g 1203 '4 1203.87 1204.24 1204.41 1204.36 1204.22 1203;85 1203.09 1202.46 1201
~ 57 1200.37 1198.94 1200.4 1227.3
'239 1240 '
Energy elease te (mil on Btu/s) 2.927615 2.844945 2.738457 2.646736 2.571811 2.492405 2.341530 2.198796 2.070804 1.938305 1 797816 1.639290 1.504854 1.417440
- 1. 305478 1.167428 1.086565 0.997553 0.901587 0.811673
,0.615700 0 467100 0 ~ 324000 0
tegral:
329100 ibm 395.014 million Btu
'Pah v e s a v a t 1 0 0 0 6.2.1-96K Amendment 14
(
6.2.1.6 Testin and Ins ection CONTAINMENT SYSTEMS Testing and inspection requirem'ents for the containment are discussed in section 6.2.6.
No other testing of the contain-ment structure is planned or required.
Testing and inspe'ction requirements for other engineered safety features that interface with the containment structure are discussed along with the applicable system descriptions.
6.2.1.7 Instrumentation A
lications The containment pressure is measured by independent pressure transmitters located at widely separated points outside the containment.
Refer to section 7.3 for a discussion of pres-sure as an input to the engineered, safety features actuation system (ESVAS).
Refer to section 7.5 for a discussion of the display instrumentation associated with pressure.
The containment airborne radioactivity is monitored by the P
a;irborne radioactivity monitoring system, discussed in section=ll.5.
Hydrogen concentration is monitored in the containment by the hydrogen monitoring system, discussed in section"6.2.5.
Temperature sensors are positioned at appropri-ate locations throughout the containment.
The temperature is displayed in the main control'oom along with high-temperature-alarms.
I "u f'
~
e 6.2.1.8 ualification Parameters for In-Containment Safet Related E ui ment In-containment safety-related equipment required to operate post-MSI.B. is qualified to the Main Steam line break design basis accident environment as specified in table 3E-1 (sheet 1 of 7).
This environment bounds the calculated pressure-time and temperature-time response of figure.6.2.1-20.
The transient
'conditions shown in figure 6.2.1-20 are the result of the con-2 tainment analysis for a 102% power,
.78 t
. slot-type Main Steam line break. with the loss of one con ainment spray train and no December 1980 7i /$
- 6. 2. 1-97 Amendment 3
PVNGS FSAR CONTAINMENT SYSTEMS loss of o fsite power.
Xt differs from the analysis of the Main Steam Line break of section 6.2.1.1.3.1 in that the equipment qualification analysis utilizes 8% condensate re-evaporation as allowed Iby Appendix B of NUREG-0588 A chronology of events for (5).
this main steam line break analysis is provided in table 6.2.1-27 'ombined effects of the PVNGS-specifi e ing of the 102%
MSL t break compared with reviou equipment qualification con 'nt temperature analysis based on the CESSAR mas energy se data for the same break used in won 6.2.1.1.3.1 is a 1.3 uc"ion.in peak contai nt vapor temperature (368.5 vs 369.8E) an 4.5 psig crease in peak containment pressure (45.6 vs 41.1 psig Xn-containmert safety-related equipment required'o operate post-LOCA is qualified to the LOCA design basis accident environ-ment as specified in table 3E-l,(sheet 1 of 7)
~
This environ-ment bounds the calculated pressure-time and temperature.-time response of figure 6.2.1-3 which shows the worst case LOCA transient as discussed in section 6.2.1.1.3.1.D.
~~4 11 6
2 ~ 1. <<98
~ Februar 1 '985
1
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-27 ACCIDENT CHRONOLOGY FOR CONTAINMENT EQUIPMENT QUALIFICATION PEAK TEMPERATURE ANALYSIS (Sheet 1 of 2) a Break type:
1028 power MSL slot break Loss of one cooling train Time (s)
Event 0.0 Break occurs Q riess 9> cL g>c ficM g2 7r 3 Containment pressure reaches safety injection actua-tion signal (SIAS) analysis setpoint (5 psig)
Containment pressure reaches reactor trip and main steam 'iso tion 'signal (MSIS) analysis setpoint
(~sig )
SIAS generated Reactor trip signal and MSIS generated b)
Turbine admission valves close Reactor trip breakers open~>>
~
f Containment pressure reaches containment spray actua-a tion signal (CSAS
) analysi s setpoint (10 psig )
CSAS generated Main steam isolation valves close~t Main feed>>ster isolation valves close~~~~
e a.
Value used in pipe break blowdown analysis.
b.
Events based on time (3.6 seconds) to reach reactor trip and MSIS for setpoint of 6 psig used in pipe break blowdown analysis.
February 1985 6 ~ 2 ~ 1-99 Amendment 14
a 1
V 4
PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-27 ACCIDENT CHRONOLOGY FOR CONTAINMENT EQUIPMENT QUALIFICATION pEAK TEMPERATURE ANALYSIS (Sheet 2 of 2)
Time (s)
Event
&39/
rc-s'.s
+ER-
-vesf-u-X~~enad~
CS pump loaded on essential bus CS pump at full speed CS headers filled and full containment spray flow established Peak containment vapor temperature of 368~ occurs
@'o Peak containment pressure of-45~ psig occurs Failed steam generator dryout occurs, blowdown ends 6.2.1.9 References Shoenhoff, H. M. and Braddy, R. W., "Containment and Safety-Related Equipment Transient Temperature Analysis Following a Main Steam Line Break," Bechtel Power Corporation, May 1975.
2.
BN-TOP-4, Rev. 1, Oct. 1977, Subcompartment Pressure and Temperature Transient Analysis, Bechtel Power Corporation, San Francisco, CA.
Amendment 14 6 ~ 2 ~ 1-100 February 1985
I r
VAPOR TMAX PMAX 92 S G VAPOR VAPO EMPERATURE 250 Q
L'h Iz Xz 4
20 z0U 10 SUMP 1 SPRAY TRAIN ON 887 SEC.
TMAX ~ 231 F SUMP WATER TEMPE RATURE RECIRCULATION FROM BLDG.
SUMP BEGINS C 2060 SEC.
I2 H.P. SI PUMPS)
SUMP WATER VAPOR END RCS SENSIBLE HEAT ADDITION
@ 86400 SEC.
(24 HR) 200 0
150 CC I-0 1
10 10 10 10 TIME FOLLOWING BREAK (SECONDS) 10 106 REBEL~
uo (M 6T<~~
p~D pg~~
p IQ,O~
Palo Verde Nuclear Generating Station FSAR CONTAXNMENT PRESSURE AND TEMPERATURE RESPONSE g
DOUBLE-ENDED DISCHARGE LEG SLOT BREAK, MAXIMUMECCS Figure 6.2.1-3
t I/
I l
n PMAX
- 63.OPSIA
@271.5 SEC 360 TMAX 296.2 F
@91 SEC 320 46 u.
40 C
32 SUMp po MAX*
CI 500 SEC 2BO CC 240 I
200 24 1 SP RAY T RA IN ON 691 SEC RECIRCULATION 82117 SEC Sy4 160
~"o 120 8
1 10 102 10 1O4 10 80 10 TIME FOLLOWING BREAK (SECONDS)
PROPOSED e
Pah Verde Nuckar Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE RESPONSE, DOUBLE-ENDED DISCHARGE LEG SLOT BREAK, MAXIMUMECCS Figure 6.2.1-3 B-26-85
50 40 Q
g 30 z
z>
20 Iz0 VAPOR
@ 102% PQ+E R TEMPERATURE SUMP TEMPERATURE ESSUR5 VAPOR TM 401 F
8 80 SEC SUMP T
~ 255.5 F X
8268 C
MAX
@ 154 SEC 1 SPRAY TRAIN ON@80 SECONDS 405 IL 300 C
i
(
L 200 i
I 100 0
1 10 100 TIME FOLLOWING BREAK ISECONDS) 1000 10000 R.KP&ce: wm-L Pi~<
- VRoVo~ Fit o~<
Palo Verde Nuclear Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT) NITH LOSS OF ONE CONTAINMENT COOLING TRAIN (Sheet 1 of 5)
Figure 6.2.1-6
/
0 10& POWER MAX OI 172 SEC 57 400 52 TMAx*
398.54 SEC 47 320 280 I
240 MAX
@320 SEC 42 CC.'
37 I
ER C
32 200 1 SPRAY TRAIN ON el 84 SEC 27
~
160 22 120 1
10 100 TIME FOLLOWING BREAK (SECONOSI 1000 17 IOOOO PROPOSED Pain Verde Nuclear Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 1 OF 5)
Figure 6.2.1-6 8-26-85
I
O 75% POWER VAPOR TEMPERATURE MAX O 80 SEC 400 SUMP T AX 255.3 F 4 284 EC 300 l
C 40 Q.
g 5
Ig 20 z
SUMP TEMPERATURE RESSURE PMAX 40.8 PSI G IS'l66 SEC 20v i
10G 1 SPRAY TRAIN ON IP 80 SECONDS 0
1 10 100 TIME FOLLOWING BREAK (SECONDS) 10000 RCPT Lcm
~m4 P2.O~~~
F l.Cu~
Palo Verde Nuclear Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (Sheet 2 of 5)
Figure 6.2.1-6
r
815% POWER PMAX ~ 54.1 PSIA OI 189 5 SE C 55 360 TMAx OI 84 SEC 45 320 CC
~~ 280 I
240 MAX=
CI 330 SEC cf 40 CL 35 o.
IR R
30 200 1 SPRAY TRAIN ON CI 84 SEC 25 lSO 20 120 1
10 100 TIME FOLLOWING BREAK (SECONDS) 1000 15 10000 PROPOSED Palo Vade Nuclear Generatinp Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT)
WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 2 OF 5)
Figure 6.2.1-6 8-26-85
0 60% POWER VAPOR TEMPERATURE VAPOR TMAX 398.0 F 4 80 SEC 400 40 U
g 30 g
IZ 20 z
4 Iz OU 10 SUMP TEMPERATURE PRESS E
SUMP TMAX 67 F O 312 SEC PMAX "40.6 PSIG IP 197 SEC 1 SPRAY TRAIN ON e 80 SECONDS 300 I'
CC 200 I
100 TIME FOLLOWING BREAK ISECONDS) 10000 p~Pos~
F lc ~RE Paio Verde Nudear Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (Sheet 3 of 5)
Figure 6.2.1-6
400 8 SOS POWER PMAX ~ 53.2/PSIA
@209 SEC 55 MAX 684SEC 45 320 280 CC I
240 Q
Q TMAX* 252.9F 6'50 SEC
~Z 40 35 IR a
cK 30 200 1 SPRAY TRAIN ON 6'84 SEC 25 160 20 120 1
10 100 TIME FOLLOWING BREAK (SECONDS) 1000 15 10000 PROPOSED Palo Verde Nuclear Generatinp Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 3 OF 5)
Figure 6.2.1-6 8-26-85
1 O 25% POWER VAPOR TEMPERATURE VAPOR T
~ 400.7 F 6 80 SEC 400 4Q Q
g g
30 Iz I
20 IZ0V io SUMP TEMPERATURE PRES JRE SUMP T M 257.1 F
MA IF 380 SEC MAX
@16Q SEC 1 SPRAY TRAIN ON IF 80 SECONDS SUMP I
300
(
203 I
109 100 TIME FOLLOWING BREAK ISECONDSI 10000 R~P~~
Mt~
A~Ac6 ED FlC u~
Palo Verde Nuclear Generating Station Pqj.
FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (Sheet 4 of 5)
Figure 6. 2. 1-6
0 400 I 25N POWE R MAX iP 240 SEC 55 360 MAX OI 84 SEC 45 320
~t 280 K
I 240 MAX CI 390 SEC cf 40 Cf 35 IR 4E Ct 30 C) 200 1 SPRAY TRAIN ON CI 84 SEC 25 tea 20 120 1
10 100 TIME FOLLOWING BREAK (SECONDS) 1000 15 10000 PROPOSED Palo Verde Nuclear Generatinp Station C".rj,,
FSAR CONTAINMENT PRESSURE AND TEHPERATURE
RESPONSE
MSLB (SLOT)
WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 4
OF 5)
Figure 6.2.1-6
l t
40 Q
K 30 Iz m
20 8 0% POWER VAPOR TEMPERATURE SUMP TEMPERATURE R ESSU R E VAPOR TMAX 38 F
8 80 SEC SUMP TMAX 256.6 F 8 342 SEC MAX 9 194 SEC 400 S
300 C
20" K
S 100 1 SPRAY TRAIN ON 8 80 SECONDS 10 100 TIME FOLLOWING BREAK ISECONDS) 1000 10000 R,e WcE u tM LYTE<.HED
'p~Pc st+
FlG uRE Palo Verde Nuclear Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
NSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (Sheet 5 of 5)
Figure 6.2.1-6
440 400 01'OWER TMax 084SEC ~
PMax 56 QfPSIa
@ 180.5 SEC 57 52 360 47 320 CC cr 280 CL I-240 Max*
Ia 340 SEC 42 CC 37 cL I
C 32 a
CO 200 1 SPRAY TRAIN ON 6'84 SEC 27 160 22 120 10 100 TIME FOLLOWING BREAK (SECONDS) 1000 17 10000 PROPOSED Palo Vade Nuclear Generatinp Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT)
WITH LOSS OF ONE CONTAI'NMENT COOLING TRAIN (SHEET 5 OF 5)
Figure 6.2.1-6 8-26-85
P r
h
~ 293 BTU/HR ~ FT F
@ 15 SEC.
DOUBLE END DISCHARGE LEG SLOT B AK MAXIMUM CCS 250 I-Z
'V 2OO O
~ 0
~N Z
U.
150 I
I0 I-Ill C7 L'00 OZ OU 50 TAGAMI F UNCTION TRAN ION REG N
RECIRCULATION BEGINS 8 2060 SEC.
UCHIDA FUNCTION 10 10 10 10 I
I I
I 10 10 TIME FOLLOWING BREAK (SECONDS) f~LJ c 6 bo<7H 4~~AED pta~~
Fit" oRE Palo Verde Nuclear Generating Station FSAR CONDENSING HEAT-TRANSFER COEFFICIENT, DOUBLE-ENDED DISCHARGE LEG SLOT BREAK, MAXINJM ECCS Figure 6.2.1-7
320 DOUBLE ENDED DISCHARGE LEG SLOT BREAK MAXIMUME CCS 240 R
lL 200 gg LL O
N hg R
gg 160 I,
I 120 D
a 80 TRANSITION REGION 1 SPRAY TRAIN ON
@91 SEC TAGAMI UCHIOA FUN CTION FUNCTION-
~ RECIRCULATION 5 2117 SEC 0
1 10 10 10 10 TIME FOLLOWING BREAK (SECONDSI 10 IPE PROPOSED Zb Palo Verde Nuc]ear Generating Station FSAR CONDENSING HEAT-TRANSFER COEFFICIENT, DOUBLE-ENDED DISCHARGE LEG SLOT BREAKS MAXIMUMECCS Figure 6.2.1-7
/ t 0
120 0
X 100 I
80 U
O 60 Z
K 40 I
Z Q
20 Z
QJ0Z O
102+ POWER 1 SPRAY TRAIN ON @ 80 SECONDS 10 100 TIME POLI.OWING BREAK (SECONDS) 1000 10000
~n-A It'inc PosEZ)
~iG<<E Palo Verde Nuclear Generatinl, Station
/N/P FSAR CONDENSING HEAT-TRANSFER COEFFICIENT (UCHIDA FUNCTION)
MAIN STEAM LINE SLOT BREAKS MITH LOSS OF ONE CONTAINMENT COOLING TRAIN Figure 6.2.1-9
t 4 ~
160 140 IK 120 m o 100 EA ~
Ie gg I
P 80 ill gg a
60 C) 40 102/ POWE R~
~ OYi POWER 1 SPRAY TRAIN ON 84 SEC 20 0
1 10 100 TIME FOLLOWING SREAK (SECONDS) 1000 1000.
PROPOSED Palo Verde Nuclear Generating Station FSAR CONDENSING HEAT-TRANSFER COEFFICIENT (UCHIDA FUNCTION)
MAIN STEAM LINE SLOT BREAKS WITH LOSS OF ONE CONTAINMENT COOLING TRAIN Figure 6.2.1-9 8-26-85
1
10 1
10 100 SUMP WATER RE CI RCULATIO BEGINS C~ 206
- SEC, I2 H.P. Sl PU S)
~P
~qPP
<9I" 10 10
~$
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REACTOR VESSE L CALCULATIONS INITIATEO
&500 SEC.
I' 41 QZ(
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REACTOR VESSEL WATFR 10 10 10 10 10 10 TIME FOLLOWING BREAK )SECONDS)
~ l'T~ QTT~4Q0 FlC ORE.
Palo Verde Nuclear Generating Station
- xi>,
FSAR CONTAINMENT ENERGY INVENTORYp DOUBLE-ENDED DISCHARGE LEG SLOT BREAK, MAXIMUMECCS Fi re 6.2.1-10
1011 1010 10 I
109 VAPOR SUIhP
~EP S)8 REACTOR VESSEL WATER 1O'0 1 SPRAY&
TRAIN ON
@91SEC REACTOR VESSEL CALCULATION INITIATEO CI 500 SEC RECIRCULATION 6I 2117 SEC 10 10 10 10 10 1O4 1O'O'IME FOLLOWING BREAK ISECONOS)
PROPOSED Palo Verde Nuclear Generating Station FSAR CONTAINMENT ENERGY INVENTORY, DOUBLE-ENDED DISCHARGE LEG'LOT BREAKS MAXIMUMECCS Figure 6.2.1-10
C l
1 l
aC
1O9 C: 10th POWER CONTAINMENT ATh1 OS P HE R E 10 STRUCTURAL MEAT SINKS CONTAINMENT SUMP 10 10B 10 1 SPRAV TRAIN ONC 80 SEC TIME FOLLOWING BREAK ISECONOSI I
I 1000 1000' gyros~
~i t~a~
Palo Verde Nuclear Generating Station FSAR CONTAINMENT ENERGY DISTRIBUTION MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (Sheet 1 of 2)
Figure 6.2.1-12
r
9 10'Ã POWER 1O'O'O' SPRAY TRAIN ON CI 84 SEC 1O' 10 1O'02 TIME FOLLOWING BREAK (SECONOS) 10 104 PROPOSED Pdo Verde Nuclear Generating Station FSAR CONTAINMENT ENERGY DISTRIBUTION MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 1 OF 2)
Figure 6.2.1-12
0
5 0'4 POWER CONTAINMENT ATMOSPHERE 10 STRUCTURAL HEAT SINKS CONTAINMENT SUMP 10 10 1
10 1 SPRAY TRAIN~
ON C~
80 SEC TIME FOLLOWING BREAK (SECONDS) 1000 1003:.
PktoMG~
F lc 0~
Palo Verde Fiuclear Generating Station FSAR CONTAINMENT ENERGY DISTRIBUTION MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (Sheet 2 of 2)
Figure 6.2.1-12
J r
4P OA POWER 108 1 SPRAY TRAIN ON 6 84 SEC 10 10 10 102 TIME FOLLOWING BREAK (SECONDSI 10 10 PROPOSED Palo Verde Nuclear Generating Station jr?r~".
FSAR CONTAINMENT ENERGY DISTRIBUTION MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 2 OF 2
)
Figure 6.2.1-12
'J
450 350 0
K g
300 IL 250 C
O 200 p
~Q ONTAINM ALIFICATI POST-M EOUIPM SLB IN C ENT OU ENT ON TEMPER URE t370 I
I I
TM
- 368.5oF PMAX 4
FI 5.6 PSI G 50 C
Z 40
'C 30 c
Ci 20 0
P 114 co 10 1 SPRAY TRAIN ON 8 80 SECONDS 100 TIME FOLLOWING BREAK {SECONDS) 1000 10 0
10,05b
~ AH A~Ac~Q
'pic ~s~
pil GgC Palo Verde Nuclear Generating Station FSAR EQUIPMENT QUALIFICATION PRESSURE AND TEMPERATURE RESPONSE MSLB (SLOT) AT 102%
POWER Figure 6.2.1-20 February 1985 Amendment 14
V
102Y POWER PMAX-55.5'PSIA e0 168.5 SE C 57 52 360 320
~ 280 240 2DD MAX 367.8~F I'84 SEC 1 SPRAY TRAIN ON TMAX= 254.9F
@310 SEC 1 SPRAY TRAIN ON C584 SEC cf 42 C6:
C6.
37 6R IR 32 27 160 22 120 1
10 100 TIME FOLLOWING BREAK (SECONDS) 1000 17 10DDD PROPOSED Palo Verde Nuclear Generating Station FSAR EQUIPMENT QUALIFZCATION PRESSURE AND TEMPERATURE RESPONSE MSLB (SLOT) AT 102%
POWER 0
Figure 6.2.1-20 8-26-85
l
PVNGS FSAR EMERGENCY CORE COOLING SYSTEM 7.
The containment spray system is designed and constructed in conformance with ASME III Class 2
requirements.
The material used in this system is austenitic stainless steel type 316 or 304 or other C-E approved compatible material.
System/Component Arrangement 2.
3.
The piping arrangement is such that the maximum and minimum head loss requirements presented in CESSAR Table ~~are met.
r.s.2. -5.b The piping for each safeguards train is designed such that the top of the piping junction of the pipe runs to the refueling water tank and the containment recirculation sump is located at least 16 feet below the top of the recirculation containment
- sump, which is 4 feet below the minimum water level in the containment during recircula-tion.
The containment minimum pressure below the refueling water ambient pressure does not exceed 3 psi as described in section 6.2.1; therefore, an increase in the differential height of the piping junction for containment underpressure in excess of 3 psid is not required.
Frictional losses in the safeguard pump suction piping between the containment sump and the junction with the RWT are less than 5 feet.
The location and elevation of the HPSI pumps are indicated on figure 1.2-4.
The HPSI pumps are located in the auxiliary building as close as practical to the containment.
6.3-21
PVNGS FSAR EMERGENCY CORE COOLING SYSTEM a ~
The elevation of the HPSI pumps is such that the available NPSH is at least 22 feet during the recirculation mode when the pumps take suction from the containment sump.
In determining this elevation, no credit was taken for subcooled water in the containment sump following a LOCA.
b.
The available NPSH was calculated at the pump suction that, was assumed to be a
maximum of 2.5 feet above its foundation elevation.
c.
The available NPSH was calculated taking into consideration the concurrent operation of the low pressure safety-injection and containment spray pumps tabulated in CESSAR Q,S.2.-5. a d.
Credit was not taken for water trapped above the containment floor.
The requirements on location and elevation applied to the HPSI are applied"to the location and elevation of the LPSI pumps except that the pump's suction is 14 inches below its supports.
The elevation of the LPSI pumps is such that the minimum required NPSH of 22 feet during injection at a runout flowrate of 5100 gal/min per pump is satisfied.
The design calculation assumes:
a ~
Concurrent HPSI,LPSI, and containment spray pump operation and flow in conform-ance with CESSAR Table 6.3-5a.
b.
A refueling water tank temperature of 100F.
6.3-22
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