ML19312A210
| ML19312A210 | |
| Person / Time | |
|---|---|
| Site: | 05000495 |
| Issue date: | 11/29/1978 |
| From: | NEW YORK STATE ELECTRIC & GAS CORP., STONE & WEBSTER, INC. |
| To: | |
| References | |
| NUDOCS 7909050296 | |
| Download: ML19312A210 (572) | |
Text
SWESSAR-P1 LIST OF EFFECTIVE PAGES (CONT)
Page, Table (T) , Amenciment Page, Table (T) , Amenciment op Figure (F) No. or Figure (F) No.
b-a thru d 39 6.2-20 20 6-1/ii 27 n.2-20A 30 6 -iles 25 6.2-20b 2b 6-iii/iv 25 6.2-20C thru 20G 34 6-v 34 ,A.2-20H thru L 39 n-v1,v14s,vJs 39 6.2- M 8 b-vii 34
- b.2-22, 22A 33 b-vlii,ix 39 b.2-23 25 6-X 25 6.2-24,24A z4 0.1-1 7 o c b . 2-25 thru 2b A 33 o.1-2 8 oA6.2-27/28 33 b.1-3 7 6.2-29 17 T6.1-1b2 34 6.2-30 8 T6.1-3 3F 6.2-31 22 T6 1-4 .' 6.2-32,32A 2S T6.1 -4 A (B6W) 34 b.2-33 8 To.1-5 3 b.z-34,34#1 25 T6.1-6 (W) (sheet 1) 15 6.2-3S/3b 18 Tb.1-6 (W) (sheet 2) 8 6.2-37 thru 39 24 Tb .1-6 (W-3S) (2 sheets) 17 6.2-40,41 lb To.1-b (B6W) (2 sheets) 30 b 2-42 17 T6.1-6 (C-E) (2 sheets) 10 6.2-42A 18 Tb .1-7 (sneet 1) 23 6.2-42B,42C 24 Tb .1-7 (sheet 2) 34 6.2-43/44 24 Tb .1-7 (sheet 3) 23 6.2-44A 22 T6.1-8 3 6.2-45/46 S Tb.1-9 (3 sheets) 7 6.2-47 thru 49 8 Fb .1-1 (W) 13 6.2-50 22 Fb .1 - 1 (W-33) 27 6.2-51 5 Pb .1-1 (dSW) 24 6.2-52,52A 33 Fo .1 - 1 (C-E) 28 6.2-53 17 6.2-1 thru 2A 13 n.z-54 5 6.2-3 thru 4D 28 o.2-55 8 b.2-5 thru 8 28 6.2-Sb 16 b.z-9 13 6.2-57 22 6.z-10 24 6.2-58 35 b.2-10A 13 6.2-SdA/58B 35 b.2-11 35 6.2-S9 tnru 60A 29 b.2-12 thru 12a 24 6.2-60B 34 6.z-12C thru 12E 28 6.2-61 thru 66 34 6.2-12F thrt. 12J 39 Tb .2.1-1 (W) 7 6.2-13 thru 14 34 T6.2.1-2 (W) 2 6.2-14A 34 T6.2.1-3 (W) (3 sheets) 28 b.2-15 thru 16A 10 T6.2.1-4 SS (W) 9 b.2-17 36 Tb .2.1-66 7 (W) 7 6.2-1d 20 T6.2.1-8 (W) 19 b.2-18A 21 To .2.1-9 (W) 10 6.2-19 24 T6.2.1-10 (W) 9 T6.2.1- 1 (W-3S) 17 6-a .7 3 r, aunenament 39 l J *;t bOJ 7figf7g
SWESSAR-P1 IJST OF EFFECTIVE PAGs3 (CONT)
Page, Table (T) , Amendment Page, Table (T) , Amendment or Fiqure (F) No. or Figure (F) No.
T6.2.1-2 (W-33) 20 T6.2.1-15 (W) (2 sheets) 13 To .2.1-3 (W-3S) 27 Tb .2.1-13 (W-3S) 17 (sneet 1) To .2.1- 14 614 A (W-3 S) 17 T6.2.1-3 (W-3S) (sneet 2)28 Tb .2.1-15S 15A (W-3S) 18 To .2.1-4 (W-3S) 17 T6 . 2 .1 - 16,17 ,17 A (W-3S) 28 T6.2.1-4 A (W-3S) 20 T6.2.1-16 (W-3S) 28 T6. 2.1-5 (W-3S) 18 T6.2 .1-13613A (BSW) 30 T6.2.1--6 67 (W-3S) 17 (2 sheets)
To .2.1-8 (W-3S) 29 Tb .z .1- 14614 A (dSW) 30 (2 sheets) T6.2.1-15 (BSW) 34 To . 2.1-9 (W-3S) 28 (2 sheets)
Tb.2.1-9A 28 Tb .2.1-15A (B&W) 34 (W-3S) (2 sheets) (z sneets)
T6.2.1-10 (W-3S) 17 T6.2.1-16 (B&W) 39 T6.2.1-1 (B6W) 34 T6.2.1-17S17A (8SW) 38 T6.2.1-2 (B&W) 34 T6.2.1-17b (BSW) 39 T6.t .1-3 (BSW) 3b To.z.1-18 (BSW) 35 (2 sheets) (2 sheets)
T6.2.1-4 (obW) (sheet 1) 39 T6.2.1-13 (C-E) 27 T6.2.1-4 (bSW) 36 T6.2.1-14 (C-E) 18 (sheets 2 & 3) T6 . 2.1- 14 A (C-E) 17 To . 2 .1-4 A (8bW) 36 Tb .2.1-15 (C-E) 17 (3 sneets) T6 .2.1- 15A (C-E) 17 T6.2.1-5 (dSW) 35 T6.2.1- 16 (C-E) 13 (3 sneets) T6.2.1-17 (C s) 24 To.2.1-o (d&W) 35 T6.2.1-18 (C-E) 24 T6.2.1-7 (B&W) 35 (2 sheets)
To.z.1-8 (bSW) 39 T6.2.1-19620 28 T6.2.1-9 (BSW) (4 sheets) 37 T6.2.2-1 (sheets 162) 34 T6.2.1-9A (B&W) 37 To .2.2- 1 (sheet 3) 29 ro.2.1-9B (uSW) 39 Td.2.2-2 (W) (2 sheets) 24 To .2.1-1 (C-E) (sneet 1) 15 T6. 2. 2-2 (BSW, C-E,h-3S) 24 To .2.1-1 (C-E) (sheet 2) 10 (3 sheets)
Tb 2.1-2 (C-E) 20 T6.2.2-3 32 T6.2.1-3 (C-E) (sheet 1) 26 T6.2.2-4 (sheet 1) 32 To .2.1-3 (C-E) (sheet 2) 28 T6 2.2-4 (sheet 2) 17 T6.2.1-4 (C-c) (sheet 1) 20 T6.2.3.1-1 8 To .2.1-4 (C-E) (sheet 2) 19 T6.2.3.1-2 (2 sheets) 8 Tb .2.1-5 (C-E) 20 T6.2.3.1-3 8 To .2.1-6 (C-E) 15 F6.2.3.1-4 16 To .2.1-7 (C-E) 15 Tb .2.3.1-5 (2 sheets) 20 T6.2.1-8 (C-E) 13 T6.2.3.2-1 (2 sheets) Orig To .2.1-9 (C-E) 10 T6.2.3.2-2 (W) 22 T6. 2.1-10 (C-E) 13 T6.2.3.2-2 (W-3S) 27 T6.2.1-11 36 T6.2.3.2-2 (dSW) 1 To.2.1-12 27 T6.2.3.2-2 (C-E) 26 T6.2.1-13 (W) 10 T6 . 2. 3. 2-3 thru 5 (W) 22 T6.2.1-14 (W) 12 T6.2.3.2-364 (B6W) 34 T6 .2 .1- 14 A (W) 16 Tb 2.3.2-364 (W-3S) 24 T6.2.3.2-364 (C-E) 24 b-u Amendment 39 7/14/78 O! $ [b
SWESSAR-P1 IJST OF EFFECTIVE PAGES (cot (f)
Page, Table (T) , Amendment Page, Table (T), Amendment 9 or Fiqure (F)
To .2.4-1 (4 sheets)
No. or Figure (F) No. _
35 F6.2.1-8 (uSW)
T6.2.4-2 17 (2 Sneets) 36 T6.2.4-3 (4 sheets) 35 F6.2.1-9 thru 12 (B&W) 35 To.2.5-1 13 Fb.2.1-13 (BSW) 37 T6.2.5-2 (sheets 162) 34 F6.2.1-13A (B &W) 39 Tb .z .5-2 (sheet 3) 26 Fb .2.1-1 (C-E) 19 T6.2. 5-3 (sheet 1) 26 F6.2.1 - laS 1B (C-4) 26 T6.2.5-3 (sheet 2) 28 Fo . 2.1-2 (C-E) 17 Tb.2.5-4 34 F6 .2 .1-2AS 3A (C-L) 24 T6.2.0-1 17 Fo .2.1-3 thru 5 (C-E) 15 F6.2.1-1 (W) 13 F6.2.1-6 (C-E) 27 b b .2 .1- 1 A (W) 26 F6.2.1-7 68 (C-E) 17 Fb .2.1-1d (W) 27 Fb .2.1-9 (C-E) 19 F6.2.1-2 (W) 12 F6.2.1- 10 (C-E) 17 F6.2.1-3 (W) 1b Fb .2.1- 11 (C-E) 28 F6.2.1-4 (W) 12 F6.2.1-12 (C-t) 17 Fo .2.1-5 (W) 13 F6.2.1- 13 (C-E) 13 F6.2.1 oS7 (W) 12 F6.2.1-14 13 Fb .2.1-7A (W) 13 F6.2.1-15 thru 19 (W) 7 F6.4.1-8 (W) 12 Fb.2.1-15 thru 19 (W-3S) 17 Pb .2.1-9 (W) 10 F6.2.1-15 thru 19 (B&W) 34 F o .2.1- 10 (W) 17 F6.2.1-15 thru 19 (C s) 10 F6.2.1- 11 (W) 28 Fb.2.1-20 7 Fo .2.1- 12 (W) 17 F6.2.1-21 0d) 10 r 6.2.1-13 (W) 13 F6.2.1-22 13 Fb .2 .1- 1 (W-3S) 17 F6.2.1-23 & 24 (W-33) 28 Fb .2.1- 1AS 13 (W-3S) 26 Fb .2.1-2 4 A (W-3S) 28 F6. 2.1-2 (W-3S) 18 F6.2.1-23 (BSW) 37 Fb . 2 .1-2 AS 3A (W-3S) 24 F6.2.1-23A (BSW) 39 F6.2.1-364 (W-3S) 17 Pb .2.1-2 4 6 24A (B&W) 37 F6.2.1-6 (W-3S) 17 F6.2.1-24b (BSW) 39 Fo .2.1-7 68 (W-33) 18 Fo .2.1-2 3 (C-E) 13 Fb .2.1-9 S 10 (W-3S) 17 F6.2.1-24 (C-E) 13 F6.2.1-11(W-3S) 28 Fo.2.1-25 b 26 (W-3S) 28 Fo . 2.1-12 (W-3S) 17 Fb 2.1-27 thru 40 28 F6.2.1-13 (W-3S) 28 F6.2.2-1 A (W) 13 F6.2.1- 1 (db W) 30 F6. 2. 2-1 A (W-3S) 24 Fb . 2 .1- la (BSW) 30 F6.2. 2- 1 A (B&W) 24 Fb.2.1-1B (BSW) 35 F6.2.2- 1A (C-E) 24 F6.2.1-2 (abW) 34 F6.2.2- B 20 F6.2.1-2A (dSW) 34 F6.2.2-2 thru 8 [W) 7 Fo.2.1-3 (uSW) 30 F6.2.2-2 thru 8 (W-3S) 17 Fb.2.1-3A (HSW) F6.2.2-2 thru 8 (B&W) 34 (2 sheets) 30 F6.2.2-2 thru 8(C-E) 27 F6.2.1-4 (uSW) 19 F6.2.2-9 8 F6.2.1-5 (dSW) 36 F6.2.2-10 34 Fb.2.1-6 (dSW) 35 F6.2.2-11 pi) 7 Fu 2.1-7 (bSW) 3o F6.2.2-11(W-3S) 17 Fb . 2.1 -7 A (B&W) 34 F6.2.2-11 (B&W) 34 b-c Amendment 39 7 -, ,, . 7/14/78 b0J lDO
SWESSAR-P1 LIST OF EFFECTIVE PAGES (CONT) iage, Table (T) , Amendment Page, Table (T) , Amendment or Fiqure (F) No. or Fiqure (F) No.
F6.2.2- 11 (C-d) 10 T6.3-2 18 Fo.2.2-12 22 Th . 3-3 (W) 21 Fo .2.z-13 (W-3S) 25 T6.3-3 (W-33) (2 sheets) 28 Fo .2.2-13 (n6W) 34 '
T6.3-3 (B&W) (2 sheets) 33 t o .2.2-13 (C-E) 25 To .3-4 (BSW) 30 F6.2.3.1-1 21 To.3-3 (C-E) (sheet 1) 23 F6.2.3.1-2 24 T6.3-3 (C-t) (sheet 2) 26 Fu.2.4-1 17 To.3-4 (W) (sheet 1) 10 F6.2.5-1 12 T6.3-4 (W) (sheet 2) 15 Fo.2.5-2 thru 4 (W) 27 T6.3-4 (W-3S) (2 sheets) 18 Fb .2.5-2 thru 4 (W-3S) 22 T6.3-4 (C-r.) (2 sheets) 10 Fu . 2 . 5-2 cnru 4 (BSW) 34 T6.3-5 18 F6.2.5-2 thru 4 (C-E) 26 Fb .3-1 (W) 7 Fo.2.6-1 34 Fo .3 - 1 (W-3S ) 17 6.3-1 30 F6.3-1 (C-E) (2 sheets) 21 6.3-z 35 b.4-1/2 o 6.3-2A 32 6.4-3 8 6.3-3/4 23 6.4-4 6 6.3-5 26 6.4-5 9 T6.3-1 18 T6.4-1 o 9
UOJ j;/
o-d isnendment 39 7/14/78
SWESSAR-P1 CHAPTER 6 ENGINEERED SAFETY FEATURES TABLE OF CONTENTS Section Page 6.1 GENERAL 6.1-1 6.1.1 Materials Specifications and Compatibility 6.1-3 6.1.2 Quality Assurance for ESF Systems 6.1-3 6.2 CONTAINMENT SYSTEMS 6.2-1 6.2.1 Containment Functional Design 6.2-1 6.2.1.1 Containment Analysis 6.2-2 6.2.1.1.1 Containment Analysis Analytical Model 6.2-2A 6.2.1.1.2 Containment Analysis Results 6.2-10 6.2.1.2 Subcompartment Analysis 6.2-12F 6.2.1.2.1 Subcompartment Analysis Procedure 6.2-12F 6.2.1.2.2 Break Type Definitions and Areas 6 .2 -12G 6.2.1.2.3 Subcompartment Arrangement 6.2-13 6.2.1.2.4 Obstructions to Vent Flow 6.2-13 6.2.1.2.5 Vent Loss Coefficients 6.2-13 6.2.1.2.6 Subcompartment Analytical Model 6.2-13 6.2.1.2.7 Containment Subcompartment Analysis Results 6.2-17 6.2.1.3 Mass and Energy Release Analysis for 6.2-20E I Postulated Loss-of-Coolant Accidents 6.2.1.4 Mass and Energy Release Analysis for 6.2-20F Postulated Secondary System Pipe Ruptures 6.2.1.5 Minimum Containment Pressure for ECCS 6.2-20G 27 Analysis
- 6. 2 .1. 6 Interface Requirements 6.2-20G References for Section 6.2.1 6.2-20G 6.2.2 Containment Heat Removal Systems 6.2-20J 6-i e < -- 4 rn Amendment 27
('t s ;D 6/30/76
SWESSAR-P1 TABLE OF CONTENTS (CONT)
Section Page 6.2.2.1 Design Bases 6.2-21 6.2.2.2 System Design 6.2-22 6.2.2.3 Design Evaluation 6.2-25 6.2.2.4 Testing and Inspections 6.2-29 6.2.2.5 Instrumentation Requirements 6.2-32A 6.2.2.6 Materiuls 6.2-33 6.2.2.7 Interface Requirements 6.2-34 References for Section 6.2.2 6.2-34 6.2.3 Cbntainment Air Purification and 6.2-34A Cleanup Systems 6.2.3.1 Supplementary Leak Collection and Release 6.2-35 System 6.2.3.1.1 Design Bases 6.2-35 6.2.3.1.2 System Design 6.2-36 6.2.3.1.3 Design Bvaluation 6.2-38 6.2.3.1.4 Testing and Inspections 6.2-42A 6.2.3.1.5 Instrumentation Applications 6.2-42B 6.2.3.2 Containment Spray System - Iodine Removal 6.2-43 6.2.3.2.1 Design Basis 6.2-43 6.2.3.2.2 System Design 6.2-43 6.2.3.2.3 Design Evaluation 6.2-44 6.2.3.2.4 Tests and Inspections 6.2-48 6.2.3.2.5 Instrumentation Application 6.2-48 6.2.3.2.6 Materials 6.2-48 References for Section 6.2.3.2 6.2-49 ee l ' 6: J
~ ;3) t 6-ii Amendment 27 6/30/76
SWESSAR-P1 TABLE OF COtITE!!TS (C01:T)
Section Pace 6.2.4 Containment Isolation Systems 6.2-51 6.2.4.1 Design Basen 6.2-51 6.2.4.2 System Design 6.2-52 6.2.4.3 Design Evaluation 6.2-54 t> . 2 . 4 . 4 Tests and Inspectionn 6.2-55 n-iia 7 e Amendment 25 U(J 1
/
U lVU 4/30/76
SWESSAR-P1 TABLE OF CONTENTS (CONT)
Section Page 6.2.4.5 Materials 6.2-55 6.2.5 Combustible Gas Control in Containment 6.2-55 6.2.5.1 Design Bases 6.2-56 6.2.5.2 System Design 6.2-56 6.2.5.3 Design Evaluation 6.2-57 6.2.5.4 Testing and Inspections 6.2-59 6.2.5.5 Instrumentation Applications 6.2-60A 6.2.5.6 Materials 6.2-60A 6.2.5.7 Interface Requirements 6.2-60A 6.2.6 Containment Leakage Monitoring System 6.2-60A 6.2.6.1 Design Basis 6.2-61 6.2.6.2 System Design 6.2-61 6.2.6.3 Design Evaluation 6.2-62 6.2.6.4 Testing and Inspections 6.2-64 6.2.6.5 Instrumentation Applications 6.2-66 6.2.6.6 Materials 6.2-67 References for Section 6.2.6 6.2-67 6.3 EMERGENCY CORE COOLING SYSTEM 6.3.1 6.3.1 Design Bases 6.3-1 6.3.2 System Description 6.3-1 6.3.3 Design Evaluation 6.3-1 6.3.4 Tests and Inspections 6.3-2 6.3.5 Instrumentation Requirenents 6.3-2 6.3.6 NSSS Interface Requirements 6.3-2 3 6-iii Amendment 25 4/30/76 00J l 'J $
SWESSAR-T'1 TABLE OF CONTENTS (CONT)
Section puae 25 6.3.7 Utility - Applicant Interface Requirements 6.3-5 6.4 HABITABILITY SYST1J1S 6.4-1 6.4.1 Habitability Systems Punctional Design 6.4-1 6.4.1.1 Design Bases 6.4-1
- 6. 4 .1. 2 System Design 6.4-1
- 6. 4 .1. 3 Design hvaluation 6.4-4
- 6. 4 .1. 4 Testing and Inspection 6.4-4 6 . 4 .1.5 Instrumentation Requirements 6.4-4
- 6. 4 .1. 6 Interrace Respiirements 6.4-5 0
'e ; u l_'
6-iv Amendment 25 4/30/76
SWESSAR-P1 LIST OF TABLES 9 Table 6.1-1 Cmrain=nt Design Evaluation Parameters - General Information 6.1-2 Containment Design Evaluation Parameters - Initial Conditions 4 6.1-3 Containment Design Evaluation Parameters - Design basis Accident (DBA) Results 6.1-4 Contain = nt Design Evaluation Parameters - Mass and Energy Addition Tables 6.1-4A Deleted 6.1-5 Conr ain=nt Design Evaluation Parameters - Passive Heat Sink Material Properties 6.1-6 Cont ainment Evaluation Parameters - Passive Heat Sink Inventory 6.1-7 Containent Design Evaluation Parameters - Engineered Safety Features (ESP) 6.1-8 Overall Heat Transfer Coefficients for Contairunent Atmosphere Recirculation Coolers during a DBA 6.1-9 Typical Materials Employed for Components of ESF Systems 6.2.1-1 Accident Chronology, Pump Suction DER, Normal ESF with Failure of One Containment Spray Pump 6.2.1-2 Energy Balance Table, Double-Ended Pump Suction Break, Millions of Btu 6.2.1-3 Subcompartment Design Pressure Differentia 1n 6.2.1-4 Mass and Fhergy Release to the Reactor Cavity, 150 in.2 Pump Discharge LDR 6.2.1-4A Mass and Energy Releases to the Reactor Cavity, 144 In.2 Hot Leg LDR 6.2.1-5 Mass and Energy Release to Steam Generator Cbnpartment Pump Discharge SER 6.2.1-6 Mass and Energy Release to Pressurizer Cubicle, Surge Line DER 6-v Amendment 34
- , , , , _ , 7/22/77 bO3 luj
SWESSAR-P1 LIST OF TART.RS (CONT)
Table
- 6. 2 .1 -7 Mass and Energy Release to Pressurizer Cubicle, Spray Line DER 6.2.1-8 Containment Peak Pressure Following Steam Pipe and Feedwater Pipe Break inside Containment 6.2.1-9 Mass and Energy Releases to Containment Main Steam Pipe DER 6 . 2 .1 -9 A Mass and Energy Releases to Containment Main Steam Pipe Break t> . 2 .1 -9 B Mass and Energy Releases to Containment, Main Steam 39 Pipe Break, 0.33 Ft2 MSLB, 102 Percent Power, MSIV and DG Failure 6.2.1-10 Mass and Energy Releases to Containmrent, Feedwater Pipe DER 6.2.1-11 Single Failure Analysis Results, Design Basis Break location 6.2.1-12 Main Feedwater Isolation Times O 6.2.1-13 Summary of Reactor Cavity Subcompartment Vent Loss Coefficient 6.2.1-13A Summary of Reactor Cavity Subcompartment Vent Loss Coef ficients and Ef f ective L/As for RELAP4 MOD 3 6.2.1-14 Vent Areas, K-factors, and Vent Flow Models Used in the Pressurizer Cubicle, Pressurizer Support Skirt, and Pressurizer Relief Tank Compartment Analyses 6.2.1-14A Mode Length-to-Area Ratios for the Pressurizer Relief Tank Compartment Analysis with RELAP4 6.2.1-15 Vent Areas, K-f actors, and Vent Flow Models Used in the Steam Generator Compartment and Shield Wall Analyses 6.2.1-15A Vent Areas , Ef fective L/As, and K-Factors Used in the Steam Generator Compartment and Shield Wall Analyses bbb ll*
6-v1 Amendment 39 7/14/78
SWESSAR-P1 LIST OF TABLES (CONT)
Taole 6.2.1-16 Peak Cuatainment Structure temperatures, Worst Steam Pipe Break Accident, 102 percent Power, MSIV and DG 3g Failure ,
6.2.1-17 Accident Chronology, Steam Pipe rareak 6.2.1-17A Accident Chronology, Steam Pipe Break 6.2.1-17B Accide7t Chronology Steam Line Break, 0.33 Ft2 SLB, 102 Percent Power, MSIV and DG Failure 39 b.2.1-18 Breaks Considered in the Subcompartment Analysis 6.2.2-1 Containment Spray System Component Data 6.2.2-2 Failure Analysis for Containment Heat Renoval Systems 6.2.2-2A Failure Analysis for Containment Cleanup Systen (W-3S, BSW, C-E) 6.2.2-3 Net Positive Suction Head (NPSH) for Engineered Satety Features Pumps 6.2.2-4 Parameters Used for Radiological Consequences of Recirculation Liquid Leakage 6.2.3.1-1 Supplementary Leak Collection and Release System Component Design and Performance Characteristics 6.2.3.1-2 Containment Leak hate Test Data 6.2.3.1-3 Containment Types BSC Test Data 6.2.3.1-4 Design Parameters for Areas Served by SLCRS 6.2.3.1-5 Potential Leakage That Can Bypass Areas Served by SLCRS 6.2.3.2-1 Nomenclature Used for Equations in Section 6.2.3.2 6.2.3.2-2 Iodine Removal Coefficients for Containment Spray o.2.3.2-3 Containment Spray System Flow Rates During an Accident 6.2.3.2-4 Chemical Composition of the Containment Spray and Sump Solution 6-viA Amendment 39
}} 7/14/78
SWESSAR-P1 LIST OF TABLES (CONT) Table 6.2.3.2-5 Chemical Composition of the Containment Spray and Sump Solution 6.2.4-1 Piping Penetrations Per General Design Criterion No. 54 6.2.4-2 Instrumentation Penetrations Per Regulatory Guide 1.11 6.2.4-3 Required Testing for Fluid Lines Penetrating Containment Structure 6.2.5-1 Combustible Gas Control System Design Parameters 6.2.5-2 Parameters Used in Calculating liydrogen Sources 6 . 2 . 5 -3 Parameters Used for Post-DBA Hydrogen Purge Analysic 6.2.5-4 Nuclear Steam Supply System Zirconium and Aluminun Inventory 6.2.6-1 Containment Leakage Monitoring System, Number of Temperature / Pressure Detectors b.3-1 initial Containment Conditions for ECCS Analysis 6.3-2 lieat Removal System Performance Parameters Used in t.CCS Analysis 6.3-3 Emergency Core Cooling and Boration Systems Intertace Requirements 6.3-4 Heat Sinks for ECCS Evaluation 6.3-5 Overall Heat Transfer Coefficients for Containment Atmosphere Recirculation Coolers During a LOCA for ECCS Analysis b . 4 -- 1 Control Room Dose Resulting From a DBA O 6-viB Amendment 39 7/14/78 (.' 1.n9 ,J ,n i;g _)
SWESSAR-P1 LIST OF FIGURES Figure 6.1-1 Engineered Safety Features 6.2.1-1 Parameter 3 for Analysis of Reactor Cavity Pressure Differentials 6.2.1-1A heactor Cavity Nodalization, Plan View through Centerline of Nozzles 6.2.1-1B Reactor Cavity Nodalization, Section View through Centerline of Nozzles 6.2.1-2 Parameters for Analysis of Steam Generator Compart-IV .at Pressure Differentials 6.2.1-2A Steam Generator Cubicle Arrangement (2 Sheets) 6.z.1-2A Steam Generator Cubicle Arrangement (4 sheets) 34 (Bf,W only) 6.2.1-3 Parameters for Analysis of Pressurizer Cubicle Pressure Differentials 6.2.1-3A Pressurizer Cubicle Nodalization 6.2.1-3A Pressure Cubicle Arrangement (Sheet 2) (B&W only) 6.2.1-4 Pressure Response, Pressuri7.er Cubicle, Spray Line DER 6.2.1-5 Pressurizer Cubicle and Support Skirt Pressure Response, Surge Line DER in Support T-irt 6.2.1-6 Pressure Response, Pressurizer Relief Tank Compartment, Surge Line DER 6.2.1-7 Pressure Response, Steam Generator Co:npartment, Pump Discharge SER 6.2.1-7A Deleted 6.2.1-8 Pressure Response, Steam Generator Compartment, Hot Leg SES (2 Sheets) 6.2.1-9 Reactor Cavity Pressure Response 6.2.1 ',0 Pressure Response, Reactor Cavity, 144 in.2 Hot Leg LDR 6'-vii ,, ,,,; Amendment 34 hOJ lO/ 7/22/77
SWESSAR-Pl LIST OF FIGURES (CONT) Figure 6.2.1-11 Calculated Pressure Dif f erential vs Angle from Break (Break at 00) 150 in.2 Pump Discharge LDR 6.2.1-12 Calculated Pressure Dif f erential vs Angle from break (Break at 00) 144 in.2 Hot Leg LDR 6.2.1-13 Containment Atmosphere Steam and Feedwater Pipe Breaks Pressure 6.2.1-13A Containment Atmosphere Pressure, 0.33 Sq Ft Main 39 Steam Lune Break 6.2.1-14 Containment Atnosphere Recirculation Cooler Heat Transfer Coefficients 6.2.1-15 Se am Condensing Coefficient (Tagami) , Injection Phase, Pump Suction DER 6.2.1-16 Steam Comdensing Coefficient (Tagami) , Injection Phase, Pump Suction 0.6 DER 6.2.1-17 Steam Condensing Coefficient (Tagami) , Injection Phase, Pump Suction 3 Sq Ft Split 6.2.1-18 Steam Condensing Coefficient (Ta gami) , Injection Phase, Pump Discharge DER 6.2.1-19 Steam Condensing Coefficient (Tagami) , Injection Phase, Hot Leg DER 6.2.1-20 Conceptual Flow Chart of the THREED Computer Code 6.2.1-21 Reactor Cavity Nodalization Study 6.2.1-22 Cooler Heat Duty (1 Cooler) 6.2.1-23 Steam Condensing Coefficient (Uchida) , Steam Pipe Break 6.2.1-23A Steam Condensing Coef ficient (Uchida) , 0.33 Sq Ft 39 Main Steam Line Break 6.2.1-24 Containment Atmosphere, Liner and Concrete Tempera-tures, Steam Pipe Break 6-viii Amendment 39 6 7/14/78 (:;,; lg]
SWESSAR-P1 LIST OF FIGURES (CONT) Figure 6.2.1-24A Containment Atmosphere, Liner and Concrete Tempera-ture, Steam Pipe DER 6.2.1-24B Containment Atmosphere, Liner, and Concrete Tempera- 3g tures, 0.33 Sq Ft Main Steam Line Break 6.2.1-25 Containment Pressure Comparison, Main Steam Line Break 6.2.1-26 Containment Pressure Comparison, Main Steam Line Break 6.2.2-1A Containment Spray System 6.2.2-1B Containment Spray System 6.2.2-2 7 Temperature Transients, Minimum ESF, Pump Suction DER 6.2.2-3 'Ibtal Fan Cooler Duty, Minimum ESF, Pump Suction DER 6 . 2. 2 -4 Heat Absorbed by Sinks, Minimum ESF, Pump Suction DER 6.2.2-5 Total RHR Heat Exchanger Duty, Minimum ESF, Pump Suction DER 6.2.2-6 Containment Pressure, Pump Suction DER, Normal and Minimum ESF 6.2.2-7 Containment Pressu e Transients, Various Break Incations , Injection Phase 6.2.2-8 Containment Pressure Break Spectrum at Pump Suction, Injection Phase 6-ix Amendment 39 7/14/78 663 lb9
SWESSAR-P1 LIST OF FIGURES (CONT) Figure 6.2.2-9 ESF Sump Screen Arrangement 6.2.2-10 Spray Nozzle Orientation 6.2.2-11 Component Cooling Water Temperature Transients, Minimum ESF, Pump Suction DER 6.2.2-12 Droplet Size Data 25 6.2.2-13 Spray and Sump pH vs Recirculation Time 6.2.3.1-1 Supplementdry Leak Collection and Release System 6.2.3.1-2 SLCRS Pulldown Time
- 6. 2. 4 -1 Containment Isolation System 6.2.5-1 Combustible Gas Control System 6.2.5-2 Recombiner Requirements Post-LOCA Hydrogen Generation 6.2.5-3 Purge Requirements Post-LOCA Hydrogen Generation O 6.2.5-4 Hydrogen Generation Sources 6.2.6-1 Containment Leakage Monitoring System 6.3-1 Emergency Core Cooling System Modification
(! ? i , ,, UUJ i/U 6-x Amendment 25 4/30/75
l n t
SWESSAR-P1 @ CHAPTER 6 ENGINEERED SAFETY FEATURES 6.1 GENERAL Engineered safety features (ESF) are those systems which are provided to mitigate the consecuences of postulated serious but unlikely accidents and to protect the public by preventing or minimizing the associated release of radioactive fission products to the outside atmosphere. Summarized below are those systems that are classified as engineered safety features. A composite flow diagram of the containment heat removal system portion of the engineered safety features is shown in Fig. 6.1-1. The containment design 7 evaluation parameters are listed in Tables 6.1-1 through 8.
- 1. Containment - The containment structure is a cylindrical, carbon steel lined, reinforced concrete structure which encloses the components and major pipina within the reactor coolant pressure boundary.
The structure is designed to contain the radioactive fluids and fission products which may result from postulated accidents inside containment. The containment is described in detail in Sections 3.8.1, 3.8.3, and 6.2.1.
- 2. Containment Heat Removal Systems - The removal of heat trom the containment structure is ensured by the follow-ing systems:
- a. Containment atmosphere recirculation system (Fig . 9.4. 5.1-1)
- b. Containment spray system (Fig. 6 . 2 . 2 -1. )
- c. Residual heat removal / low head safety injection system (NSSS Vendor's scope) .
The containment spray system, in conjunction with the containment atmosphere recirculation system and the residual heat removal low head safety injection system can depressurize the containment to a low pressure within one day following the design basis accident (DBA) and maintain this low pressure inside the containment indefinitely. The containment heat removal systems are described in Section 6.2.2.
- 3. Containment Atmosphere Purification and Cleanup Systems The containment atmosphere purification and cleanup systems consist of the:
- a. Containment Spray System (Fig. 6. 2. 2-1)
- b. Supplementary leak collection and release 6.1-1
' h ~ Amendment 2/28/75 7
SWESSAR-P1 a. b. Containment Spray System (Fig . 6.2.2-1) Supplementary leak collection and release ll system (Fig. 6. 2 . 3 .1 - 1)
- c. Combustible gas control system (Fig. 6. 2 . 5 -1)
Caustic (sodium hydroxide, NaOH) is added to the containment spray to reduce the concatration of radioactive iodine in the containment indefinitely following a DBA. Following a LOCA, the supplementary leak collection and release system filters the air of the annulus building prior to release, thus collecting fission products which may have leaked, either through the containment penetrations or from components in the annulus building containing radioactive materials. The combustible gas control system maintains the concentration of hydrogen in the containment atmosphere below the flammable limit following a major loss of coolant accident (LOCA) . This system is started manually after an accident as required to maintain the concentration of hydrogen below 4 volume perc ent ds specified in Regulatory Guide 1.7. The combustible gas control system is described in Section 6.2.5.
- 4. Emergency Core Cooling System (ECCS) - The ECCS (NSSS Vendor's scope) provides an emergency borated cooling water supply to the reactor core for the entire spectrum of reactor coolant system (RCS) break sizes to limit the core temperature, maintain core integrity, and provide negative reactivity insertion for additional shutdown capability. For the Westinghouse design only, the ECCS includes the emergency boration system for n egative reactivity addition. The interface requirements for the ECCS is discussed in Section 6.3.6.
- 5. Containment Isolation System - To ensure containment integrity following an accident, fluid system piping which penetrates the containment is provided with containment isolation valves. The containment isolation valves are located both inside and outside the containment. The containment isolation system consists of check valves, normally closed manual valves, and valvas which close automatically on receipt of a containment isolation phase A (CIA) signal or containment isolation phase B (CIB) signal.
Section 6.2.4 describes the containment isolation system and Section 7.3 describes the actuation of the isolation valves. The ESF systems are designed to be in accordance with all 8 applicable requirements of the single failure criterion as discussed in Section 3.1.65. . , 6.1-2 bb) ' Amen dment 8 3/28/75
SWESSAF-P1 The enoineered safety features systems and their auxilia ry supportinc systems which include the reactor plant component cooling water system (safety related portions) and the reactor pla n t service water system are powered from the ereraency buses (Section 8. 3.1) . 6.1.1 f:uterluls Specifications und Compatibilitv T3pical naterials employed ior components of the ESF, includina tlone exposed to the containment spray and core coolina water, ure listed in Tuble 6.1-9 with the corresponding specification numbers as applicable. The selection of materials is based upon corrosion resistance to tl e boric acid anc sodium hydroxide solutions. General corrosion, Jntercranular Corrosion, Caustic, and Chloride stress corrosion have been considered. The material properties are 7 raintainec durine instdllation consistent with the requirements 01 Reculatory Guide 1.71 as described in Section 3A.1-1.71. The reauirements of Feculatory Guides 1.31 anc 1.44, as discussed in Ecctions 3A .1- 1. 31 and 3A.1-1.44, are not for austenitic stainless steel. The recuirements of Feaulatory Guide 1.50 as discussed in Section 3A.1-1.50 are ret for low alloy steels. Contuminatien und cleanliness control are provided consistent with rer uirements of Peculctory Guides 1.37, 1.38, und 1.39 un discussed in Sections 3A.1-1.37, 3A.1-1.38, and 3A.1-1.39. 6.1.2 ouality Assurance for FSF Systens To ensure a ligh uuality level in the FSF components and systems, a cuality assurance program is implemented durina the desian, manufacture, installation, and testina of the ESF systems. The ouullty ussurunce procram is described in C1. apter 17 of this SAR for Stone S Webster supplied equipment, and in Chapter 17 of the USSS Vendor's SAF tor the NSSS Vendor supplied equionent. v a
\
v 6.1-3 Arendrent 7 2/28/75
SWESSAR-P1 TA13 LE 6.1-1 LbNTAlfE.14T DwlGti FNAI UATION PA1AMFTERS - GENadu INFUkMATION B&W C-E W-41 W-3S A. Design pr essur e, psag 48 48 48 48
- a. Desi g n t e-sy mr .s t u r e , t 280 280 280 280 C. t'r ee visl ume , 106tt8 3.44 3.42 3.42 3.06 D. Design (:nax al lowa ble) leak r ate 0.2 0.2 0.2 v.2 41.4 43.5 39.8 M (wt 1/(ts y ) .s t. mix calculataxi pressure 38.7 (PSig)
E. E.xt ermal <lesign pr essur e, psig 4 .0 4.0 4.0 4.0 o g&, C 3.a c:D3 e]< '; 3 (.~;
-',1, .r> &
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g c. j Vias cs c L: m m/ 1 of I p.me s t w ri' 34 1/2 ' ff !
SWESSAR-P1 TAliLE b .1-2 CONTAINMEffT DESIGN EVA.LUATION PARAMETERS - INITIAL CONDITIONS A. Peactor coolant sys tesn engineered satety 1eatures ratings B&W C-E W-41 w-3S
- 1. Reactor; core p>wer level, Mwt 3,876 4,100 4,100 3,636
- 2. Coolant tm> era tures , F Inlet 568 564.5 559.5 555.9 14 Outlet 627 621.2 623.8 616.4 p~ 3. Ksss of reactor coolant liquid plus steam, lixn 576,000 586,111 579,570 504,640 L- 4. Liquid plus steam energy, 106 ut u 350.4 350 351 302 B. Co nt.ainment 1
__ _ ' 1. IT e ssure, psia 15.7 15.2 16.2 16.2 L 34 _,,,, 2. Inside tenqerature, F 105 105 105 105 A
- 3. Outside tenperature, r 100 100 100 100
[' ,-]
- 4. Relative humidity (inside) , 1 40 40 40 40 LJ 5. Service water temperature, F 95 95 95 95 rm 120 120 120 120 7 mJ e,1 (')
- 6. Refueling water temperature, F
- 7. Initial component cooling
$"h)
W =, water tengwarature, F 105 105 105 105 EPe. .s c D 1% 1 of 1 A:nendment 34 7/22/77
3 LIII3 MCIS ACCIDCf! (DM). F_E__L'_L7_5 W-41(2) w.33(5) N Ptrp IW Hot leg tisetArce Per Suetien Het Lei Discharse Prp Suet ier. Prr Euction Split (') D9 On DD Da Lpp split Eplit b) re litN DD CD LOP 1r lit 8.25 10.5 6.29 3.0 9 17 e.25 10.5 6.29 3.c 2 9.92 7.5t 5.99 9 17 41.9 40,o 3g .k 43 5 42.9 42.2 36.9 35.5 33 9 39 0 3? .1 41.9 41." M 3.0 2t2.9 F 3.0 258.1 255.k 2' 3.* Ptr.7 261.6 252.9 25C.5 257 9 25/.4 255 0 15 15.3 151.0 155.0 1p g,c 209 0 214.0 214.0 17.0 19.4 159 0 161.0 173.0 17 .
- _
505.9 k97.3 3 2.1. - 313.6 43e .0 kg.2 % . c. 505.7 507.0 Sc5,4 3T C 375.0 513 1
%,4 19.3 00 91.1 S7.8 f9.4 21.0 18.9 83.5 62.7 52.2 79.7 Po.5 33 A:endrent 35 1 of 1 10 / 6
P[jh0;dC.'.il . >
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n.r T_AC EEADtTC OEION D'Al?"ATI:N FAFAC W Dr; Break location Not leg (l) Pqil) Discharge Pz; Sudion Het les Discharge Break Type EEF Q Q Q Q Q Q Srlit Crlit Break area, ft 2 15.75 8.55 4.27 2.136 11.17 5 585 2.793 19.2 9.52 Peak pressure, psig 38.2 35.2 38.3 36.E 36 .9 33.7 36.6 37.4 b0.0 Feak tarperature, F 256.1 256.1 256 .3 253.8 254.3 256 .9 253.5 255-9 FC .1 Time of peak pressure, see Ik.0 195.0 174.0 2C5 0 104.0 15?.0 172.0 11.3 259.0 Energy released to contairasent at time of peaa pressure,10' Btu 3e9.8 3to.B 354.4 355.b 358.7 355.3 351.2 355.3 512.9 Energy absorbed by passive sinks at time of peak pressure, 106 Btu 17.41 05.3 79 2 79 0 64.6 60.4 7a.0 7.17 St .2 Results shown on figure 6.2.2-8 6.2.2-7 6.2.2-7 The contairement atmosphere and sep water teperature transients are shown in Fig. 6.2.2-2. The conttiment pressure transients are shown in Figs. t .2.2-6 through 6.2.2-6. (1) Failure of one diesel gener. tor is the most severe failure for the B&'d ESS. (2khe second contalment peak pressure follovira a cold leg break is calculated asseing a single active failure of one coctairnent spray FM for tr e Westinghouse W-41 ESS. I3) Failure of one diesel generator is the most severe str41e active failure for the 0-E ESS. (b)The pressure transients for treaks at various locations and trese for the epectre of breais at the pep suction considers failure of one cmponent coolird water pump (failure of both air coolers on ene train to receive cocling water) for the C*E ESS. (5)The second contairsnent peak pressure fol.lovira a cold leg breal is calculated asselr4 a single active failure of one caponeLs coolir4 water p'. ssp (failure of both air cociers on one train to receive coeltr4 water). 4 ~ f f (J \6 ( i U ./
SWESSAR-P1 TAnLE 6.1-4 CONTAINMENT DESIGN EVALUATION PARAMETERS - MASS AND ENERGY ADDITION TABLES
'Ihe mass and energy addition rates are supplied in the NSSS Vendor's SAR. The followina table lists the specific sources used in containment analyses.
NSSS Vendor SAR Tables Table Issue Date Babcock & Wilcox B-SAR-205 6.2-4a May 20, 1977 thru 6.2-5b, 6.2-6a thru 34 10b, 6.2-11a thru 12b, 6.2-13a thru 15b Combustion Engineering CESSAR 6.2.1 ~, thru Original 6.2.1-12 6.2.1-16 thru 6.2.1-19 Westinghouse RESAR 41 6.2-3 thru 6.2-7 April 1974 6.2-9 thru 6.2-14 6.2-17, -18 Westinghouse RESAR 3S Note 1 July 1975 Note 1: Refer to Shepard, R.M., Massie, H.W., Mark, R.H., and Doherty, P.J., " Westinghouse Mass and Energy Release Data for Containment Design," WCAP-8624-P-A, June 1975 (proprietary) and WCAP-8312-A, Revision 1, June 1975 (non proprietary) . A ref?rence will be provided for the mass and energy release rates used in the steam pipe @ break analysis when it becomes available. 1n () lj "1 lI / 1 of 1 Amendment 34 07/22/77
SWESSAR-P1 Table 6.1-4A is deleted 34 l l 7/22/77
SWESSAR-P1 TAB' E 6.1 -5 CO!rl' AIM:11!T DI; SIGN LVALUATION I' APAMETER S - F/J;SIVE IILAT SINK MI,TLF I AL PPOPERTIE S Sp>CifIC Trierm l Heat Con <f uct ivit y Capscity Density pit er i.e 1 Eufb r/t t /F) f Bt uL1 trn/F) (1 tan /f t 8) 3 Patnt 0.125 *
- Concrete 0.8 0.1b 145 Stainless steel 10 0.11 490 Ca r ix>rt steel 2h 0.11 490
*Patut ir, a s swn**1 to absor b ran heat , but to merely present a resist ance to heat transfer. With .in assumi t hickness of 0.006 in., the g.a irs t 111m preserats a mnductance of 250 Btu /hr/ft2/F.
O O L.' J C3 1 of 1 Amenttment 3 10/15f14
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- e. g,cn t g strc e alai s ui l ro e>ma t
nirm oe l ncx,ti rl l i rta tj o a t rg sr an li r t e ssa r grter ug u c on,f u s orr tn d tirl mn rk mto e ci ctee uo ac acpen um g af ve cr lu eapna ra n eiot co or teuir tr i Rlcs Ab Pt Srslc Sf R S 5 6 7 8 3 9 0 - 1 1 1 1 1 2 W
SWESSAR-P1 CD c-TABLE 6.1-6 CONTAI!NENT EVALUATION PARAMETIES - c7 PASSIVE REAT SINK INVENTORY - y Slab Slab Surface Area Material Exposure Exposure No. Description Each Face, ft2 Material Thickness Face A Face B 1 (bntainment 35,200 Painted Steel O.5 in. Containment Outside Dome Concrete 2.5 ft Atmosphere Atmosphere 2 Containment 75,600 Painted Steel 0.375 in. Containment Outside Side Walls Concrete 5.25 f t Atmosphere Atmosphe,re 3 Containment 3,760 Painted Steel 0.375 in. Water Outside Sidewalls Concrete 5.25 ft Atmosphere 4 Containment 13,600 Painted Concrete 2.0 ft Water Insulated Floor Steel 0.25 in. mncret e 10.0 ft 5 Refueling 8,230 Stainless Steel 0.35 in. Containment Containment Cavity Walls (bnerete 4.0 ft Atmosphere Atmosphere 6 Interior 18,870 Painted Concrete 2.0 ft Containment Containment 3e Concrete Atmosphere Atmosphere Walls 7 Interior 5,020 Painted concrete 2.0 ft Water Insulated Concrete Walls and Floors l 8 Interior 2,050 Painted Concrete 2.0 ft water Water Concrete Walls 9 Interior 6,330 Painted concrete 3.0 ft Containment Containment Concrete Atmosphere Atmosphere Walls and Floors 10 Interior 3,020 Painted Cbnerete 3.0 ft Water Insulated Concrete Walls 11 Interior 31,535 Painted concrete 4.0 ft containment containment Concrete Atmosphere Atmosphere Walls and Floors 12 Ductwork 85,700 Painted Steel 0.04 in. Containment Insulated
& Cable Trays Atmosphere BT,W 1 of 2 Amendment 30 1/28/77
_I D hC , Nde H t t t t t t 07 ne ne ne ne ne ne er 37 er d er er er er / e me e me me me me me tB r nh t nh nh nh nh nh n2 uB ip a ip ip ip ip ip e/ s as l as as as as r as m1 oe t o u to to to to e to d pc xa nm s n nm ot nm ot ot nm nm ot t a nm ot n e ot CA cA cA CA W CA e EF CA I T t t t t t t t ne ne ne ne ne ne ne er er er er er er er e r me me me nh nh me me nh nh me nh me nh nh ip ip ip ip ip ip uA ip as r as s as as as as as to e to oe to to to to to pc xa nm ot nm nm ot ot nm nm ot ot nm ot t a nm ot EP CA CA CA CA CA CA W CA s . ls ae . . n . . . . n in n n i n n n n i rk i i i i i i ec 0 2 5 0 5 5 0 ti 0 0 ah 3 MT 2 2 0 0 1 1 1
)
T N O 1 C ( P l l l
- l l l l l R 6 e e e e e e e e 2 A - e e e e e e e e t t t f S 1 t t t t t S S o S S S S S S S E 6 l 2
W a d is d d d d d d S E i e e e e e e e e L r t t t t t t t t U e n n n n n n n n A t i i i i i i i i T a a a a a a a a a M P P P P P r P P 2 t af e r, 0 0 0 0 0 O 0 0 Ae 8 5 0 0 0 ? 0 0 c 1, 7, 0, 2, 2, 4, 5 2, ea 7 9 1 cP 7 1 5 8 a 3 2 2 fh rc ua SE ,
, s s s t s t r k t t n et sr e
,ro n r e ee l r o
r o , a i a n rep re dr a i p p o T A n nu ,x r,p p lmr,l i evp dou,izGis ,& a aP iatsP t g r Cr,ng so osu u aesg p rCSs rt n gss C tD ttSs Ss Rteet i i neur i nrr eier r ar n n SPLr r Gh e m gs o d iee r r gl goee mroem em e o bc cnudsp eo tlt a adddpnd aepnu nu nnepp as rtaliep ro aol l li ni p a r enpal al ainmp l e SD oarorru BBCCBPs bl CF roi o GCP P oraruri teuro ro raiuu PBl BSCG SGSCC CC CMLPS l b w
- a. 3 1
4 1 5 1 6 1 7 1 8 1 9 1 0 2 c lSNo~ B
pO4
- 7. 3 ,
cC g 05
, . 17 s . . n /
ls . n . . n i . . t5 ae n . it nt n i . . n . . n n1 in t t t t it nt F tiF it 5 n n i n n i e/ rk F F F F F iF 5 F F 0 7 i i i i m5 ec 5 75 50 5 3 8 5 0 d ti 5 0 0 0 20 55 32 20 0 1 5 0 2 5 0 n ah 5 2.6 2 e MT 1 2 3 4 06 02 05 101 04 0 0 0 1 0 1 2 1 m A t e et e e t lt e e t lt e e e et e ee e ee r c cr rc rce tcl e l e r c em t e el l l e l e l l e e e le l n n n n sn e o t e n se e et e e e et e t e Y R o c co c co sc ss o t s o c sc s s t t s t s t s s s s t O l e e e s T a d d d d ed dt dt d t ed d d d d d d d d N E i r e e e e t l e ee ee e e l e e e e e e et e e t t t nt tr tr tl r nt t t t t t t t V N e n n n n in i nc ne nec in n n n n n n n n I t a ia ai aia ita ai imina an i ei att w ai i i t a c a i a i a i i i i a a a a t M P P P P SP PO PO PSO SP P P P P P P P P e P I S e T c A a D f* I rt uP E S 0 0 0 0 0 0 0 0 V , 0 0 0 0 0 0 0 0 0 0 0 5 0 0 0 0 I S ea ce 9 3 0, 2 6 2, 6, 5, 3 0, 0, 9, 3 7, 5 1 S ar 0, 3, 5 9, 1, 5 7 3 6, 6 5 3 3, 3 0 8 1, 2, A FA 2 8 1 7 2 3 7 1 9 8 2 3 4 8 3 5 1 1 6 P P - 2
- 1 -
R f A 6 S f o S S E R E os e 1 E L T .c w B E oa S A M Nr 2 2 2 2 2 1 1 1 2 1 2 1 1 1 1 1 2 T A R A P N e O r d l I e n Va T h a Rr A r p u U o s , ,t L A o n u e sc l t du N I s f t J a a g ur tt r & s g ss g T o y t gn n N o s a n ni ed i s I l l r essri m dn m k H f l t u l a i a a n P I d l a a w e re il er uf u g, f r t a A n n r l ao f k T n a a e l o y b to el Vc l l n N o c n n l o t a nc re Ea e i o O i s s s s a a l i c o e ,r e a i C t p l l l l l l l l r e d w f v a d cno ss,t se y w t s r t c r i a a a a f t t t c n di r r gi e e e r c w w w w an n e n e n e g a nt el eam aa va vra a a n a l n j n s r r r r a m m n gu cu or nor <tf r u r c i r u e m k l D e o o o o r i i i i t i n i n i n i l r nc s y c o, t t h r e e r e r e r a a a e . ir hc h l c r t r l t t t u ti cu cde u a e a a t t t t e n n n f ac trattue r l S I n nI n I n Iu F C o C o C o R e w re D Gr Bs Hsa S P S att t l f o a oP l b
- a. E lo o 1 2 3 4 5 6 7 -
SN 1 2 3 4 5 6 7 8 9 t 1 1 1 1 1 1 1 C
SWLSSAR-P 1 C.O TABLE 6. '-6 (COffr) e-slab No. of Face Surface Material No. Slab Descriptiora Faces Area, Pt a Material Thickners iG
'D 18 Steam generator and reactor coolant 1 7,170 Painted steel 3.0 in.
ptanp supports, min stearn line g restraints, RV stud tensioners, RV head lif t rig spreader, upper guide structure, upper guide st r . lift rig, and corm wpgx>r t barrel lift rig 19 Ring duct 1 11,200 Painted steel 0.060 in. C-E 2 of 2 Amendment 10 5/15/75
3 3 2 2 { l 36 27
/
d t1 e a 3 r s te/ me e ueyr l m3 mnt u a r A iet e ngt a 0 r i nae b Ar b MECt 4 > t 1 1 1 1 1 1 h d 3 3 r s e e e e 1 eyr t t t u nt u o Jo riet s N ! r i f a 0 onae 6 NESF 4 6 2 2 2 2 2 2
)
F d S e L ( r s me ueyr e S mnt u 0 E iiet 0 R ni vf a 0 U 1 i r s. e 5 2, A A T. T 4 MFSF 3 6 2 2 3 N N 2 2 E W F d Y e r s T L e e l eyr 1 ant u 0 A uiet 0 S t rgf a 0 onae 5 8, A A D 1 4 ESF 3 6 3 3 4 N N 3 3 T J F E d T e ner se T I C ueyr N nnt u 0 E iiet 0 njf a 0 1 E ihae MESF 4 0 6 4 1 1 A A N N 4 1 S - R C 7 L d 1 - T e F 1 L r a h f MA l eyr e e 3 A R ant u 0 S E A mict 0 S L P rgia 0 f 2, o E W s? N onn e NESF 4 0 6 A A S I
' O 4 2 2 N N 4 2 1 I
T d A e U. r s e t I d e m eyr V i t nnt u E iiet ngf a 0 0 N i nar 0 G 0 A A I W MESP 2 6 4 1 7 N N 2 1 E S T I E d D e T, r s e e l eyr h t ant ti 0 e f miet rgf a 0 0 l onae 0 4, A A A NESF 2 6 4 2 1 N N 2 2 T. m s s s sp s sp c - e p ) - e ) e c - - - n m 8 s n m 8 n m j e t eg e c n i u p ( e i u ( - i u i l ) r l p ) n l p n si j ,l pn ,l i I s n e f f ea o f f ea f f y s r ep r I r u o o tt ao ei tt o o tt ao r e o o t o re u, y s r r e e rt( ac r r e rt( u r r e t st t s s ie e s e e t m a b. sn em e b b n w dj b b w s b b ae St l u rN r t ei ro at f e rn po u m z t l onp mi en m um uom en l p m ro m u u s m Pp Ss i N N Fg r N N Fg pi N N ey u y ht ee t vS c c eS v gc t r nu wc i sn
. . ie . .
c c
. oe . .
G A a b i n i lj a b I s a b Lj a b so to ai ci Pt . At . . . 1 1 2 3 A B
' J V O
SWFSSAR-P1 TABLE 6.1-7 (COffr) B&W C-E W-41 W-3S Nor:nal Minimum Normal Minin:um Normsl Mirtiurum Nor:M1 tiininuu:n rhjineereu Engineered Engineered Engineered hjineered Engineered Enginwred Engineered Satety Sa f ety Safety Safety Satety Safety Saf ety Satety Fe.s tu r e s Features Features Features Features Fea t ure s Features Features
- c. Flow rate, gin (total) c a 10,000 5,000 10,000 5,000 8,700 5,800 6,000 3,000 C. Cantairuent Spray System
- a. Ntsabe_r of lines 2 1 2 1 3 2 2 1
- b. Number of ptzsps 2 1 2 1 3 2 2 1
- c. Number of headers 4 2 4 2 6 4 4 2
- d. Flow rate, g pa (total) 7,000 3,500 7,000 3,500 10,500 7,000 7,000 3,500 D. Fan Lbolers
- a. Ntsaber of units 4 2 4 2 3 2 4 2
- b. Air side ilow/
fan, cfm 67,000 67,000 67,000 67,000 67,000 67,000 (7,000 67,000
- c. Component cooling water floes / fan, glum ( 2 ) 2,000 2,000 2,000 2,000 2,000 2,000 2,000 2,000
- d. Heat transfer coefficient See Table 6.1-8
- e. Heat transfer area / tan, fte 1,315 1,315 1,300 1,300 1,300 1,300 1,300 1,300 E. Heat Exchangers O 1. Residual heat riinoval
& Parallel-counteuflow Paralle l-teterf low Parallel-counterflow ,j a. Type Paralle l-counter f low
- b. Ntsaber 2 1 2 1 3 2 2 1
- c. UA, 106 utu/ 14 4.00 2.00 5.50 2.75 7.929 5.286 4.6 2.3
-D hr/F
~
2 o1 3 Amendsmnt. 34 7/24/77
SWESSAR-P1 TABLE 6.1-7 (CONT) 35W C-E h-41 W-3S florTaa l th > r ma l M i n i mtL"1 No mal Minimum Norwl Minimum flinimtun Engineered Engineered Engineered Engineered Engineered Engineered Enqineered Engineered Saf et y Safety Saf et y Satety Safety Saf ety safety Satet y Features Feat ur es Feattires Fea t " r e s Features Features Features Fe.it u re s
- d. Flow rates (total (1) Fe cir cula -
tion side, 8,600 5,700 5,800 2,900 l23 gpm(*) 10,000 5,000 7,000 3,500 (2) Component cooling water 7,000 7,'20 0 22,000 11,000 14,700 9,000 15,200 [3 flow, gpn(2) 15,000 ( * )During the blowhawn, reflood, and post-reflood periods f ollowing a LOCA, the ettect of t i.e safety injection r;ystem is considered in the NSS$ Vendor's mass and energy release rate analysis. %e rates shown are the maximum values used by LOCTIC to calculate the long term contsinment response. (r>The temperature of the compone nt cooling water is shown in Fig. ti . 2. 2- 11. ( 2)!DCTIC does not consider these pumps in the long term containnent analysis as t hey take s' ct ion f rom the residual heat rnoval pumps. C' Q v:
~l tQ 3 of 3 Amendment 23 3/31/76
SWESSAR-P1 TABLE 6.1-8 OVERALL HEAT TRANSFER COEFFICIENTS FOR CONTAINMENT ATMO5PHERE RECIRCULATION COOLERS DURING A DBA Heat remaoval capacity per fan vs containment atmosphere mole traction steam Mole Fraction Steam U, htu/hr/ft2/F 0.0 90 0.035 130 3 0.1 205 0.2 330 0.3 450 0.35 495 0.4 525 0.5 578 0.6 620 1.0 730 The containment atmosphere temperature vs time curve is shown in Fig. 6.2.2-2. The heat removal rate vs time curve is shown in Fig. 6.2.2-3. 1 of 1 Amendment 3 10/15/74 (, (, ,') 1n?
\ iJ
f*TSSAF-P1 T' ' ' E 6.1-9 TYPICT I. f'1,TF117.IE Et'.PLOVr.D FOR C0"POr~ TS OF F SF SYST1 ?T Connonerat ar.d Vaterial f4a t eria l Creci t i ca t ion
- li'Ine
- 1. Stainless Steel St.- 312TP 3 0 0 , C7 -3 7 0'"r 3 0 4 ,
SA-3 5 8 TP 3 0 4 CL .1, SA - 3 76 T'316
- 2. Carbon Steel SA-106Gh.B Fittinas, Connect ior.n , and Flanges
- 1. Stainless Steel SA-1921304,SA-403MP304, S7 -4 0 %'P 30 4W , C/.- IR 2F316
- 2. Carbon Steel SA-105, FA-234hPI.
Gaske F]exitallic, St yle CG TP30uSS with 7.nbestor 7 Filler, Asbestos imt eren-nuted Srh Boltine
- 1. Studs SA-143 GP.B6,SA-193 GP.B7
- 2. t.uts SA-194 GR . 6 , S A- 1'3 4 Gl< . 2H Valven
- 1. Stainless Steel
- a. Bonne t ?'uts S7-194 GF . 6
- b. Valve St er. n S7-182 GP.F6,SA-187F304 "onol 400 or K-500
- c. Pody Cart inos EI.-351CF8,S7 -3 51CF8?'
- d. Focy 1orcinos S A- 18 2 F 304 , Sli- 18 2F 316
- e. Packino Grat hi t+- Impreonated 7.nbeston with Inconel Insert O
, 1 n .i 1 of 3 (, (~, ,'; i / ' tame ro' men t 7
./28/75
SWITS A F -P 1 TABLF 6.1-9 (CGI:T) Cortson en t ..nd fauteriul l'at e ria l Simc i f i ca t i o_n
- 2. Carbon Ste el n
- e. Ponnet !uts S7.-14h GR.2P
- b. Bonnet Studr. F7-143 GF.B7
- c. Valve Ster.n SA-182 GR.F6 c' . boty Ct.ntanca SI -216hCI
- e. Cody Forcincs SA-105
+
. ?>ackinc Graphite !"Tr ecn u ted 7nbestos, with Inconel Innert
* t n % alr are Fool Rinun 7
- 1. ?'et a l:. Austenitic Stainless Steel
- 2. Plartics Polyet' ylene , t'ylon
' . Ll ant orie rn PON2-1 ,Viton,haturi;l I t;bber Con tainrient Snray ?!ozzles luntenitic Stainless Steel Con t it 3 n"'e nt ::prav Purn
- 1. Suction Casina SA-182F300,ST-312'.<304
- 2. Dischurce Column S7-312TP30u
- 3. eisch<: rue teat ST-312W70u,S1.-1R2F30u 4 " 1: .r' a t m . Flano" C7-1E2F30u
- 5. Isoltina St-19318
- n. ' ~oc! a ni ccl 'eal Tuncnten Carbide, Carbon ESF Furpr
- 1. Plate SA-2u0
- 2. '"ra sh I< a ck s Galvanized Ftainless Steel i
2 of 3 Amendment 7 2/28/75 i
SWT SSI.It-P 1 TAPLE 6.1-4 (CO!T) Cmtonent a n t' t 'a t or ia l fatorial Speci t ica tion
- 3. I;creen Stainless Steel tonia 2_nmont 7 tmosph e re Recirculation Chol ers
- 1. Cooline toils Co/Cu-t:i (40-10) alloy SL-111 Islloy 706 (0.004 Kall)
- 2. Four.inc Galvanicon Stainless Steel h el d in e. Iaterials
- 1. Iorritic St
e SuuP O ' rw I R17 m - rH g *, i I^w ( i y 1 F l i O -+-K: >0-+C .
-T x l
FIG. 6.1-1 1 4h @ ENGINEERED SAFETY FEATURES T a/ D '
'A 1
O' ku = f -
~
PER REFERENCE PLANT SAFETT ANALTSIS REPORT ( -) SIESSAR-P1 3 n y . b .,) i AMEN 0 MENT 13 6/30n5
I INS CCN SIR ACCUMULATOR "o; w e :c n : N W , 4_ 3 I N . sz q , HEAT EXCHANGER (TTi TO . COLO LEG /vs/vs AAAA 1800 SPRAf 1800 SPRAT HEADER HEADER A dd
%^ 'c $4 =
' O
~
f rA DG I 4- i r
'r
-> 5[ REACTOR PLAkT
[ 1r C0uPONENT COOLING IATER (TTP. OF 3) FIG. 9.2.2-1 3630 SPRAT HEADERS A/sAA a :c x O
'G-DC
@ O I
REACTOR UN PLANT
~
/
1 C0uPONENT CCOLING IATER FIG. 9.2.2-1
- k<
4 (TTP.) b[ If CONTAINNENT ATu0 SPHERE RECIRCULATION COOLER FIG 9.2.0-1 (TTP. OF 3) ; 7\7\7\ 7\ 3333 1800 SPRAT 1800 SPRAT NEADER HEADER (; ti J i e V
T REFUELING
~ 2 p RATER SAFETY INJECT 10N REFUELING I WATER
" CHErtCAL N
T ADDIT 10N 4 TANM I
, r, I W ,
BPPPN INJECTION g, w T o t,". M Ou
--=0 3 3 0.
\('
REACTOR b MC w
\
@b L.D.
PLANT C0"PONENT O C00 Lit,G W ATER(TYP ) gp Mggg -_ /1 fMFIG9.2.21 -E J ru.g euveS e g/,oEXCHANGER en1 ,
-
- 44
__ C& A-1 5,R A, { PUYP T7
~
-, 2 h-t.0. l' t.0.T t _T + -CD- 43~~3 6
mumme ,
. l. ENT SFRAY ense "r
LH$t.0.&g a
=
c-4 : 2 L _I FIG. 6.1-1
=
y lc ENGINEE.EC C SAFETY FEATURES PWR REFERENCE PLANT SAFETY ANALYSIS REPORT SWESSAR-PI R-3S f, 7 i O ) u) 1// AMENOMENT 27 6/3046
INSIDE INSIDE CPNTAIN4ENT --e.+ ANNULUS STRUCTURE Bull 0 LNG
/ ~
ALCUMUIATOR l (TYP) TO SOT
& vA LEGS
- 2 l 1 TO HDT LEGS l
4
}-
t - r, A e i L i , d i Qu IO CPLO M I M 9 a rq es.J LEGS q
- -e-
, M l i 6 b , ,
, )
i, u
- (h.
c
\ -
/ -
/\ /\ 'l 3E0 0 SPPAY e s , \ is HEADERS
/ ' '
u ESF - Stju p i r
- le k _
,~ r, -
L.0 q TO COLD LEC (M P) 360 0 SFPAV
'\'\',
; 'c I
, HEADERS
, l REACTOR A , f\ ,-
, PLANT i ,
COMPONENT , g C00LINC EST - M eATE9 StjuP r' FIG.9.2.2-1 , , . CONTAINMENT ATNOSPHERE t1 b X RECIRCULATICN I
/ HEA' COOLER t -
' EDM FIG.9.2.8-1 (TYP. OF 4)
INSIDE INSIDE ( ) I CONTAIN"ENT D ANNULUS STRUCTLRE BUIL0 LNG -, , D ? .) L_ u O
O REFUELING WATER STORAGE TANK CHEMICAL A00lil0N
- TANK 7
i , M l0 3
.C. l o , ,
1 ,
- M AKE UP/H I GH 25 PRESSURE INJECTION :
PUNP (1YP. ) , L.D. h
/ 1, G _ _ L.O.X XL0
'k I DECAY HE AT REMOVAL ,
1 LOW PRESSURE
-t INJEC!iDN PUWP (TYP) l x
v, ._~ ~&' W
"-S'\
Qj 1 r H I X7 U ir T 4,L.D. CONTAINVENT k WI
.a SPRAY PUWP (TYP. )
6N
) ,
'L.D.
i r gr ~
=
1 r mr sk , r ' X IC s
's DECAY HE AT REMOVAL / 9 COOLER (TYP.) I
. i h, b[~ O p i f /4
' /
t.0. l l ll : L.h. 2 Yi eli. v L.D. FIG.6 1-1 t.0. , g ENGINEERED SAFETY FEATURES i - N PWR REFERENLE PLANT i SAFETY ANALYSIS REPORT SNESSAR-PI B&W c_ ] AMENCWENT 24 4,'2376
lNSIDE INSIDE CON T A I NVE NT
-**,4-ANNULUS STRUCTURE Bull 0 LNG TO COLD -.- 7 _
LEGS q __ ,, _ A (TYP.) L.O. L.O. CORE FLOOD TANK (TYF.) L . D . L. O.'
* - REACTOR /
m :: ,4
, r -
d ' PLANT M CONPONENT >$ E FIG.9.2.2 ' W b CCNTAINWENT ATWOSFHERE
~
g I j RECIRCULATION COOLER F1 G.9 2.B-1 ( TYP. ) TO CORE - j FLOODING N0ZZLE ESF ' l
< le
( TYP. ) SUWP I I l l __n
, /vVN/se /v~v v N/ ,, ,,
360 SPRAY HEADER _ (TYP.) : ._',, m r
= xl 4 k i f I
, r d b L.D. L.O.
O
- rI , ,
4 L.0. L.O. E {
, .M o M
hl
;c -__ c , x- -
4 T g
!hP 1)
, /\/vV v ' /vVV W te -
- Le m ._ _vs
_, P ..* h +,s
1 REFUELING WATER REFUELING
' "q ,a
'N '
STORAGE WATER L.O. TANK CHEMICAL HIGH PRESSURE IN!ECTION PUMP __. X 'p ADDITION TANK t a r, i L.O. M LOW PRESSURE INJECTION PUMP q 'A A - l4 L . O. .M XL.O. CONTAINMENT SPRAY PUPP , _ L.0. y' ~' .
- - - ~
! L.O. h r l : _
, r 1 r I
HIGH PRESSURE INJECTION PUNP L.O.
--N :^
LOW PRESSURE INJECTION PUMP L.O. N
--N :' X
=
CONTAINMENT L.O. , g SPRAY PUMP M i L.0. l 1/ l IN g FIG. 8.1-1 ENGINEERED SAFETY FEATURES P6R REFERENCE PLANT SAFETY ANALYSl3 REPDF.T SIESSAR-P1 nq7 (, (;. , L. U J CE
4 1 SAFETY ,, 1 - ECil0N ,y _ ,3 , (TYP OF 4) l ) - C- '( ^T, ;
< r l 1
l 1 -
&D s
A f -
/MYM M T lc h b f ._ '
1
~. aun r 3600 SPRAY g ~T hy y l3 - - -
, HEADER (TVP OF 4) _L ("J v 9 ESF _ - i , i- = = = 7
\ SUMP
*a- -
~ - -
; ry 1
m A _ i_ T
/ l I
l '
--> i r I
ii) s I tA S
,~ _ < ;
I A a . I
'A 4---- T0/FROM REACTOR PLANT
'" T I
COMPONENT l _ _ , COOLING IATER , , I CONTAINMENT , 1, ATMOSPHERE , - RFalRCULATION j , /\fgfg fy, _t COOLER , (TYP) T I
/ AA/' /\
1D@ 0 1_gN ESF T SUMP w, INSIDE CONTAINMENT ANNULUS / r ,, g STRUCTURE BUILDING [: C J c U 'h
P E
,Q U J ,d w u
SWESSAR-P1 6.2 CONTAIMENT SYSTEMS 6.2.1 Containment Functional Design The containment structure is designed to withstand internal pressurization due to high energy pipe breaks .tnside the containment and external pressurization due to inadvertent actuation of the containment heat removal systems. Internal Design Pressure The containment structure is designed to withstand the peak calculated pressure following a loss-of-coolant accident (LOCA) with an adequate margin. The LOCA which results in the highest calculated containment pressure is the design basis accident (DBA) for the containment structure. The percent margin is defined as: % margin = 100X [ design pressure (psig) - peak DBA pressure (psig) l peak DBA pressure (psig) The method of calculating the peak DBA pressure is described in Section 6.2.1.1. 13 External Design Pressure Inadvertent operation of the containment heat removal systems may cause a decrease in the pressure inside the containment, thereby imposing an external pressure differential on the containment structure. The maximum external pressure differential is calculated by determining the minimum attainable pressuce inside the containment and subtracting this value froni the barometric pressure. The containment is of the atanospheric design and thus has a normal operating pressure equal to or greater than the barometric pressure. A minimum inside pressure calculation is performed which considers the initial relative humidity to be 100 percent, an initial containment pressure of 14.7 psia, and the initial containment temperature to be at its maximum value of 105 F. The above assumptions are chosen to minianize the mass of air inside the containment. The calculation also assumes that containment heat removal sy stems initiation results in a 32 F heat sink for cooling the containment atmosphere. This heat sink is assumed to instantaneously condense all water vapor from the atmosphere resulting in a pure air atmosphere at 32 F. The maximum external differential pressure calculated trom the above assumptions is 2.9 psig. The design external pressure is listed in Table 6.1-1. 6.2-1 Amendment 13
,,, ,, n . 6/30/75 beJ 'uO
SWESSAR-P1 The pressure dif f erential of 2.9 psi is conservatively high since the spray water cannot be at a temperature which is less than the 65 F minimum temperature held inside the refueling water storage and chemical addition tank area. Also, the minimum 13 chilled water temperature for the fan coolers is 44 F. 6.2.1.1. Contauunent Analysis There can be two peaks in the containment pressure transient following a LOCA depending on where the LOCA occurs. There is only one peak following a reactor coolant system (RCS) hot leg rupture; there are two following an RCS cold leg rupture. 13l A pressure peak occurs near the end of the initial blowdown of the reactor coolant system after a double ended rupture (DER) of either a hot or a cold leg. Its magnitude is a function of the following parameters:
- 1. the containment free volume
- 2. the mass of air inside the containment structure (a function of initial pressure and temperature)
- 3. the amount of energy flow out the break during the initial blowdown of the RCS
- 4. the surface area of passive heat sinks within the containment structure The highest first peak pressure occurs after a DER of a hot leg.
This event releases the most energy to the containment atmosphere during the initial blowdown since the hot leg pipe size is larger than that of a reactor coolant system pump discharge, and there is no resistance to flow due to an RCS pump as is the case with a pump suction DER. The magnitude of this peak pressure is independent of the active engineered safety features (" minimum" or " normal") because they do not beccxne effective until af ter the first peak is reached. However, the accumulators do have a small effect on the first peak. Following a cold leg break (but not af ter a hot leg break) , a second pressure peak may occur after the end of the core reflooding period because the core effluent during and after reflooding passes through the steam generators to reach the break. The steixa generator secondary water contains suf ficient energy and the steam generator tubes have sufficient surface area to boil the liquid portion of the core effluent and to superheat all the etfluent to a temperature approaching that of the steam generator secondary water. The magnitude of the second peak is a function of the following parameters: b[ h , 6.2-2 Amendment 13 6/30/75
SWESSAR-P1
- 1. the containment tree volume
- 2. the mass of air inside the containment structure
- 3. the surface area of the passive heat sinks within the containment structure
- 4. the rate of heat removal frexn the containment atlaospher e by the cantainment heat removal systems
- 5. the retlood velocity in the core accomplished by the ECCS
- 6. the rate at which core effluent is vaporized and superhea ted
- 7. the amount of heat stored on the secondary side of the steam generators
- 8. the length of the retlooding period The initial containment atmosphere condition selected for use in the analysis is that combination of normal operating pressure, temperature, and humioity which yields the highest calculated peak pressure following the postuluted DBA. The values used are given in Table 6 .1-2 .
6.2.1.1.1 Containment Analysis Analytical Model The LOCTIC computer program wbich is used to model the containment system, the RCS, the heat sources and sinks, and the containment heat removal systems was developed by Stone & Webster. A topical report (1) describes in detail the assumptions used and the mathematical formulations employed. Comparisons made with the standard CONTEMPT (2) code show excellent agreement. LOCTIC has been under development for 10 years and was used in the design of two operating stations: Surry Power Station Units 1 and 2(a) and the Maine Yankee Atomic Power Station ( * ). Other station designs that have utilized the LOCTIC code are the Beaver Valley Power Station Units 1(sa) and 2,(IS) the North Anna Power Station Units 1, 2, 3, and 4(5)(6), Surry Power Station Units 3 und 4,(zo) und Millstone Nuclear Power Station Unit 3.(17) Tables 6.1-1 through 8 list the input data used in LOCTIC, which performs a digital integration of the effects of mass and energy releases to the containment atmosphere. i: , n r. o f'i U J cdo 6.2-2A Amendment 13 6/30/75
SWESSAR-P1 6.2.1.1.1.1 Mass and Energy Releases to Containment 28 The mass and energy releases to containment for the Reference huclear Power Plant are supplied by the NSSS Vendor. The NSSS Vendor's SAR should be consulted used. for the mathematical models Loss of Coolant Accident The following calculations which are bypassed during the blowdown, reflood, and available in LOCTIC are post-reflood periods because the effects are considered in the NSSS Vendor's analysis:
- 1. Mass and energy releases to containment
- 2. Decay heat addition to the reactor coolant 3.
Power coastdown heat addition to the reactor coolant 4. Core sensible heat addition to the reactor coolant 5. RCS hot metal heat addition to the reactor coolant
- 6. Steam generator secondary and (fluid metal) energy addition to the break effluent 7.
Accumulator flow to the reactor vessel
- 8. Pumped safety injection flow to the reactor vessel After the post-reflood period, LOCTIC computes the mass and energy release (boil-o f f) to the containment atmosphere, which is governed by decay heat.
The rates of mass and energy release to the containment during the blowdown, reflood, and post-reflood periods are supplied in the NSSS Vendor's SAR. listed in Table 6.1-4. The applicable SAR tables and figures are The LOCTIC interface wit? NSSS Appendix 6B. supplied data is discussed in Steam Pipe Break Accident Mass and energy release rates for this accident are discussed in Section 6.2.1.4. 6.2.1.1.1.2 Thermodynamic State 28 LOCTIC calculates the temperature and pressure of the containment atmosphere break as a function of time following a high energy pipe accident. This calculation assumes the containment atmosphere to be a homogeneous mixture of steam and air. 6.2-3 , Amendment 28 087 8/6/76
SWESSAR-P1 When the atmosphere is determined to be saturated with water vapor, the break effluent is added to the atmosphere and is assumed to a ttain thermal equilibrium with it (temperature flash) . When the atmosphere is superheated, a break effluent quality is calculated, and the steam portion is added to the atmosphere. The liquid portion is added to the containment floor without mixing with the containment atmosphere at the saturation temperature corresponding to the total pressure which exists during the time interval (pressure flash) . 28 6.2.1.1.1.3 Condensation Heat and Mass Transfer The equations for heat and mass rates of steam removed by condensation by the fan coolers is presented in the discussion of the containment atmosphere recirculation system. Condensation on the containment heat sinks is discussed below. CESSAR NSSS 28 The condensation heat transf er rate, q, to the passive heat sinks is obtained from: I q=h{ A. 1 (T -T W ) (1) Sat i=1 i where: h = Uchida condensing heat transfer coef ficient, cond Btu /hr-ftz-p A = Surf ace area of the i-th heat sink, f t2 1 T = Saturation temperature corresponding to the partial Sat pressure of the vapor (dew point) , F T y, = Surface temperature of the i-th heat sink, F 1 I = Number of heat sinks The specific internal energy change of the condensed steam, U" , becomes f9 U" =U (T ) -U (T ) +U (T ) + 3/8 C (T -T ) (2) fg sh c g Sat) f9 Sat p Sat w where: U sh (Tc ) = Specific internal energy of the superheated vapor at the containment atmosphere temperature, Btu /lbm U (T ) = Specific internal energy of saturated vapor at 9 Sat T Sat, Btu / lum s U (T ) = Dif f erence between specific internal energies of f" Sat 6.2-4 Amendment 26 8/6/76 e
- t. J L)
SWESSAR-P1 the saturated vapor and liquid states at T , Btu /lbm Sat C p = Specific heat of water, Btu /lbm-F T, = Average heat sink surface temperature, F Equation 2 describes the specific internal energy change of the condensed steam from the superheated state to a subcooled liquid state. The subcooled liquid state is based on temperature gradient through the film and subcooling a to linear the average liquid temperature of the condensate film (Ref 24) . The average heat sink surface temperature is obtained from: I I T = { A T A w 1 =1 i w i[= 1 , 1 (3) i The mass rate of steam removed by condensation on passive heat sinks, m cond , is calculated from m = q cond U" fg and theisenergy carried to the floor by the condensed steam, q , cond q =m U (T ) - 3/8 C
~
(T -T) cond cond_f Sat p Sat w (5) RESAR-3S NSSS The condensation heat transfer rate, q
,to the passive heat sinks is obtained frcm: cond 28 I
q (6) cond =h{ . A. 1 (T Sat
-T W.
)
i=1 1 h = cond Uchida condensing heat transfer coefficient, Btu /hr-fta_p A 1 = Surface area of the i~th heat sink, ft2 T Sat = Eaturation temperature corresponding to the partial pressure of the vapor (i .e . , the dew point temperature), F T
- Surf ace temperature of the i-th heat sink, F 1
6.2-4A Amendment 28
,,; m <
1 8/6/76 bbJ Li i
SWESSAR-P 1 i = Number of heat sinks The mass rate of steam removed by condensation on the i-th heat sink, m cond ' 18 1 e m = il / (H sh -H ) cond. cond. sub i (7) 1 1 where H = specific enthalpy of the vapor, Btu /lbm H = specific enthalpy of subcooled liquid film on sub the i-th heat sink
= H7 -3/8 C (T Sat
~T W }
i Cp = specific heut of water (1 Btu /lbm - F) The denominator of equation (7) describes the specific enthalpy change of the condensed steam from the vapor state to a subcooled liquid state. The subcooled liquid state is based on subcooling to the average liquid temperature of the condensate film (ref erence 24) . 28 The rate of energy transfer to the floor by the condensed steam from the i-th heat sink is:
- U 9 sump condi sub t ( 8)
The total rates of mass and energy transferred to the floor are calculated by summing equations ( 7) and ( 8) over all heat sink slabs. 6.2.1.1.1.4 ECCS Spillage During the blowdown period of the accident, the accumulator spillage mass and energy rates are included in the Vendor's mass and energy release data which is added to the containment atmosphere. During the reflood period, the ECCS spillage rates are not included in the Vendor's break effluent data but are supplied by the Vendor as a separate tabulation. Since this spillage is colder than the containment atmosphere, LOCTIC adds the ECCS spillage during the reflood period directly to the containment floor. During the post-reflood period, the ECCS spillage rates are calculated as the difference between the pumped ECCS rates to the reactor coolant system and the break effluent rates. The program checks the spillage enthalpy for saturation at the calculated containment pressure. If two phases are calculated from the
<~,
c 6.2-4B Amendment 28 8/6/76
SWESSAR-P1 Vendor's data, the steam portion is added to the containment atmosphere. Table 6 .1 -4 lists the applicable Vendor SAR tables. d.2.1.1.1.5 Condensing Heat Transfer Coefficient 78 No distinction is made between the condonning coef11cients for wat er vapor on steel and painted concrete surfaces. Although Reference 7 states that the experimentally measured heat trunnter met r icient on concrete surfaces was found to be 40 percent of the value measured on steel surfaces, the same reference also states that "a painted concrete surface can allow dropwise mndensation and, therefore, have a great transter coetficient comparable to the value for steel." All concrete surfaces which are exposed to the containment atmosphere are painted. A thermal conductance of 250 Btu /ft2-hr-F, based upon an assumed luint film thickness of 0.006 in., is also used in the heat transfer model. The thermal resistance of the paint tilm is in series with that of the condensing film. loss of Coolant Accident The Tagami heat transfer correlation is used in the containment LOCA analysis. The use of the Tagami coefficient as described below differs slightly from Reference 7 in that the peak condensina coefficient is taken to occur at the time of the first peak pressure rather than at the end of blowdown. This is logical since the steam / air ratio is greater at the tim of peak pressure than at the end of blowdown. The maximum heat transfer coef ficient, which occurs at the first containment peak pressure, is given by: [E \0.6-max Vt UUJ c. ; J 6.2-4C Amendment 28 4/23/70
SWESSAR-P1 h = Tagami heat transfer coefficient, max Btu /hr-ftz-p E = Energy released to the contaimnent E atmosphere at the time of the first peak pressure, Btu V = Containment free volume, ft3 tg = Time of first peak pressure, sec 75 = empirical constant (see Ref erence 7) Before the first peak pressure is reached, the heat transfer coefficient is calculated as: h=h (t/t ) Where t is the time in seconds after the accident. After the first peak pressure is reached, the heat transfer coef ficient equation takes the following form (Reference 7): h=h + (h -h )e -0.05(t-tp) h Stag max Stag where: h = 2 + 50 x Stag and x = steam / air mass ratio When the temperature of the containment atmosphere is less than or equal to the heat sink surface tamperature, a natural convection coefficient of 1.8 Btu /hr-f t2-F is used. The program checks the surf ace temperature of each heat sink slab and uses tbu.) 6.2-4D Amendment 28 8/6/76
SWESSAR-P1 the condensing coefficient or natural convection coefficient as applicable. The coefficients used in the containment analysis are shown in Fig. 6.2.1-15 through 19. Steam Pipe Break Accident The Uchida heat transfer correlation (Ref 25) is used to calculate the heat transfer from the containment atmosphere to the passive heat sinks following a steam pipe break. The heat transfer coefficient is calculated with the following Expression: h = AP s
/ (3.15Pt) it 0.015Ps /Pt 5 0.19 or
-3.5 h = Ae (1-Ps /Pt )
if P s /Pt> 0.19 Where: A = heat transfe coefficient for pure steam (300 Btu /hr - f tz - y) Pg = partial pressure of the steam, psia P = total pressure of containment atmosphere, psia. Fig. 6.2.1-23 presents the Uchida heat transfer coefficients for the most severe steam line break accidents as a function of time. I 6.2.1.1.1.6 Passive Heat Sinks .0 The passive heat sinks include the containment s tru cture , internal concrete, and miscellaneous metal equipment within the containment. The containment structure and internal concrete are divided into up to 20 slabs for the purposes of this analysis. The slab thicknesses and surface areas of the heat sinks used in the containment analysis are given in Table 6.1-6. Each slab is assumed to be exposed to the containment atmosphere on one or both sides, as appropriate. Heat conduction through the containment structure but is of minor significance. to the ambient air is also considered Resistance to heat transf er at the liner-concrete interface is considered in the containment analysis by use of a conservatively low value of thermal contact conductance of 100 Btu /hr/ftafy 6.2-5 y.g ) 3 Amendment 28 8/6/76
SWESSAR-P1 (Ref 21, 22, 2 3) . Since the steel liner is used as a form for pouring of the concrete, and since the concrete mix is very wet, the liner, in effect, becomes " glued" to the concrete. The bond is so intimate that in an initial attempt to determine the shear strength of the concrete anchor studs the load applied to the edge of the steel plate exceeded the stud shear strength by about a f actor of 4 before the concrete finally failed in shear. In order to obtain a configuration with which the stud shear could be successfully measured, it became necessary to oil the steel plate to prevent bond development when the concrete was poured. 2 The steel plates used in the test were 3/8 in. thick by ft x 3 ft : the concrete blocks were 1-1/2 ft thick by 2 ft x 3 ft; and the anchor studs were 5/8 in. diameter and 6-9/16 in. long. The model considers transient heat conduction to the containment structure paint through the composite thermal resistance made up of the film on the steel liner, the liner itself, the liner-concrete interface, and the concrete. The LOCTIC computer program uses the explicit method of Dusinberre(23) for calculating heat absorbed by the static heat sinks. Since the program allows input of the time increment to be used in the calculation, the maximum heat sink slice thickness consistent with this method is calculated as follows: t/2 DX = (AMat) where: DX = slice thickness A = thermal dif f usivity,k/p c 9 = material density k = material thermal conductivity c = material heat capacity at = time increment M = dimensionless modulus defined by above equation In order that the method be numerically stable, M must be equal to or greater than 2. A value of 4 is used. Sufficient mesh points are used to include initially the entire steel liner as well as 5-1/2 to 6 in. of the concrete which is in contact with the liner or the first 5-1/2 to 6 in. of concrete where no liner exists. As the temperature change penetrates the heat sink, the mesh points are extended further into the material. Approximately 100 points are established at all times, and 6 in. of material are added at a time (if possible) . Upon 6.2-6 Amendment 28 Lu o
', u/6/76 8
SWESSAR-P1 each addition, the mesh point spacing is reestablished always within the requirements of the previously stated method. The mesh spacing used in the containment analysis is 0.1 in. or less into the first 5-1/2 to 6 in. of concrete. 6.2.1.1.1.7 Containment Atmosphere Recirculation System 28 Under accident conditions, continued circulation of the containment atmosphere through the containment atmosphere recirculation coolers removes latent and sensible heat from the atmosphere by condensing steam and cooling the steam-air raixture, thus aiding in depressurization of the containment af ter a LOCA. The effect of the atmosphere recirculation coolers on containment depressurization is determined by performing heat balances across ta.e cooler and on its gas and water streams. The heat balance on the containment atmosphere gas side is given by: q=m [H -H - (S -S ) H ] air m1 m2 1 2 lig Where: 9 = rate of heat transfer from steam-air mixture to the cooling water for one cooler, Btu /hr m = mass flow rate of dry air for one cooler, air lbm/hr H = enthalpy of the steam-air mixture at the cooler ml inlet, Btu /lbn dry air H = enthalpy of the steam-air mixture at the cooler m2 outlet, Btu /lbm dry air S = absolute humidity at the cooler inlet, lbm steam / 1 lbm dry air S = absolute humidity at the cooler outlet, lbm steam / 2 lbm dry air H = average enthalpy of the liquid condensed from the lig steam-air mixture, Btu /lbm The heat balance on the cooling water side is given by: q =m C (t c -te) 6.2-7 g ,; --
; Amendment 28 UU) L i /
8/6/76
SWESSAR-P1 Where: q = rate of heat transfer to the cooling water for one cooler, Btu /hr m = cooling water flow rate for one cooler, lbm/hr C = heat capacity of cooling water, Btu /lb - F pc tc = cooling water inlet temperature, F tc = cooling water outlet temperature, F 2 The containment atmosphere recirculation coolers are cooled by the reactor plant component cooling water system. The component cooling water temperature versus time curve is shown in Fig. 6.2.2-11 for the DBA. The heat transfer rate for a counterflow cooler is calculated by:
- t t \-
Saj-t Sat - t e ) q = U. A. [ Sat - 2 t 1
)
En Sat 1 c) 2 Where: q = rate of heat transfer to the cooling water for one cooler, Btu /hr U = overall heat transfer coefficient based on A , i Btu /hr-ft2-p i A = inside tube surface area for one cooler, fta i t Sat
=
inlet saturation temperature (dew point) of steam-air mixture, F 1 t Sat
= outlet. saturation temperature (dew point) of steam-air mixture, F 2
t and t are as defined above. c c 1 2 663 2:3 6.2-8 Amendment 28 8/6/76
SWESSAR-P1 When the inlet steam-oir mixture is saturated, the exit stream is assumed S2, is to be saturated, and the absolute humidity at the outlet, calculated value of dry bulb temperature for a saturated t t steam-air mixture at a trial m2 : Sat 2 , 13 When the entering steam-air mixture is superheated, on absol ut o humidity (S coil ) is calculated for a saturated steam-air mixtuw at the average cooling temperature of the last row of coils over which the mixture passes before leaving the cooler. The absolute humidity at the exit (S ) is then 2 calculated by linearly interpolating between the entering superheated absolute humidity (S1) and the absolute humidity at the average temperature (Scoil ) or at the saturation value ot the last coil absolute humidity at t m2 , whichever is less. S = min of S 2 Sat m2 or - g m2 coll _ coil + S -S m 3 - t coil t . 1 coil where t ml = inlet dry bulb temperature of steam air mixture, F t m2 = outlet dry bulb temperature of steam air mixture, F in which the trial value or temperature t g is used. By a method or iteration, these four relationships are used to determine the four unknown quantities: the heat transter rate (q) , the atmosphere exit temperature (tm2) (S } , and the cooling water exit temperature nd absolute humidity transfer rate is then multiplied by the magnitude (tc2). The heat of the time increment and the number of coolers assumed to be operating to yield the amount of heat removed by the containment atmosphere recirculation system curing the time increment. This simplified calculation has shown good agreement with calculations which have been performed by the air cooler vendors. The heat transfer rate as a function of containment atmosphere saturation temperature is given in Fig. 6.2.1-22. The rate at which steam is condensed and removed from the containment atmosphere by one cooler is given by:
, fn = [n (S -S )
cond air 1 2 The of rate of mass condensation is then multiplied by the magnitude yield time increment and the number of the coolers operating to the sump atthe an unount averageoftemperature condensateof: which is added to the containment i3 t = 1/4 (t +t +t +t ) cond Sat Sat c1 c2 1 2 This temperature corresponds to the average liquid enthalpy (11 ) defined above. lig bUJ -a
'i/
6.2-9 Amendment 13 6/30/75
SWESSAR-P1 Method for Determining U;: The heat transfer capability of the cooling coils for normal plant operation is verified by the cooler vendor. In addition to cooling coils are normal cooling requirements, however, the expected to remove heat from the containment atmosphere following a LOCA. The heat removal capability of the cooling coils has been determined experimentally (1*) under accident conditions. These experimental coefficients are used in the heat transfer calculations f or the containment atmosphere recirculation system. The experiments show the heat transfer coefficient for the coils increases with increasing containment pressure under accident conditions. This is due to the increased heat transfer by vapor condensation on the coils with increased moisture content of the atmosphere. The mole fraction of steam or, equivalently, the pressure fraction of steam in the atmosphere is, therefore, the most indicative independent variable against which to correlate the experimental heat transfer coefficients. This heat transfer experiment was performed under somewhat ideal conditions using clean cooling water and clean tubes. It is desirable, however, to use values for the heat transfer capability expected in a situation in which some fouling will likely occur on the inside of the tube, thus reducing the heat transfer coefficient. The degree of fouling depends on the condition of the cooling water. The containment atmosphere recirculation coolers are cooled with reactor plant component cooling water. For component cooling water, a fouling factor of 0.0005 hr-ft2-F/ Btu is used. The heat trans f er coe f ficients used in the analysis are listed in Table 6.1-8. These values are compared to the experimental data in Fig. 6.2.1-14. 6.2.1.1.2 Containment Analysis Results 24 6.2.1.1.2.1 Loss-of-Coolant Accident The containment pressure transient analysis is performed with the LOCPIC computer code. A summary of the peak containment conditions following a LOCA is given in Table 6.1-3. The design basis accident (DBA) break location f or containment design is that break location which yields the highest containment peak pressure. This " worst break location" is determined from the containment pressure transients shown in Fig. 6.2.2-7 for each NSSS Vendor. The containment pressure transients for a spectrum of break sizes at the " worst break location" are given in Fig. 6.2.2-8 for each NSSS Vendor. NcJ -l3 6.2-10 Amendment 24 4/23/76
SWESSAR-P1 The cesign basis single active f ailure for containment design is that single failure which yields the highest calculated containment peak pressure. The results of this analysis are given in Table 6.2.1-11. The contairuaent pressure tollowing a pipe rupture at the " worst cold leg break location" is shown in Fig. 6.2.2-6 tor a period or one day following the accident. The containment pressure is shown f or nonnal und minimum t.SF operation. (See Section 6.2.2.1 for definitions of normal and minimum ESF operation.) These
< 7 on1 bOJ LL i 6.2-10A Amendment 13 6/30/75
SWESS AR -P 1 pressure transients include the addition of steam generator secondary side energy to the break effluent. The containmant atmosphere and sump water temperature transients for the " worst cold . break location" are shown in Fig. 6.2.2-2 for the minimum ESF as ,amption. The chronology of events for the " worst break location" is given in Table 6.2.1-1. The distribution of energy in the containment before and during the DBA is shown in Table 6.2.1-2. Also shown in the ta ble is the heat generated by the various sources and the heat removed from the containment by the various heat removal systems.
' 35 RESAR-41 NSSS The above conclusions are based on analyses performed with an initial containment pressure of 16.2 psia. To demonstrate that the containment design incorporates adequate margin at the Preliminary Design Approval stage, a reanalysis was performed with an initial containment pressure of 15.2 psia. This pressure is consistent with the initial pressure used for the CESSAR NSSS analysis, and includes an uncertainty for the pressure monitoring instrumentation.
For the revised initial pressure, the containment peak pressure following a DBA (i.e., pump suction DER with normal safeguards except failure of one spray pump) is calculated to be 42.1 psig (5 6.8 psia) . The revised initial pressure thus yields a design margin of 14 percent between the peak calculated pressure and the containment design pressure. This analysis is based on the mas' and energy releases as presented in RESAR-41. These were developed using a reference ba ck pressure of 52 psia. Sensitivity studies have shown that under certain conditions the mass and energy releases increase as the hick pressure increases . An evaluation has been made which indicates that the effect would be an increase of approximately 1 psi for the range 52 to 62.7 psia. 6.2-11 //' 7 Amendment 35 10/6/77
SWESSAR-P1 The Westinghouse reflood code results for a " typical" 4-loop plant (Ref . 2 6) indicates, as shown in Fig. II-1-10, an approximate 1 percent difference in the energy discharged during the reflood when the calculation is made with a back pressure of 52 psia (interpolated) and compared with a calculation made with a back pressure of 61.7 psia (higher than the SWESSAR-P1 calculated pressure of 56.8 psia) . This comparison is valid for the RESAR--41 plant because the same reflood code is used and both containments are of the dry, positive pressure design. There are various parameters for which values have been assumed, and the following initial assumptions currently utilized in the analysis will be verified at the Final Design Approval stage:
- a. Heat sinks
- b. Pump flows
- c. Containment volume
- d. Spray water temperature
- e. Other tactors having a lesser effect.
The margin between the calculated peak pressure (4 2.1 psig) and the design pressure (48 psig) is thus considered adequate for a Preliminary Design Approval. Results of an analysis performed using mass and energy releases developed using the SWESSAR-P1 containment design pressure as the reference back pressure will be provided in the first construction permit application referencing SWESSAR-P 1/RESAR--41. A 10 percent margin between the calculated and design pressure will be maintained f or those construction permit applications referencing SWESSAR-P1/RESAR-41 Preliminary Design Approval. 24 6.2.1.1.2.2 Main Steam Pipe Break Inside Containmer,t Structure The pressure responses resulting from a main steam pipe break accident and a main f eedwoter pipe break accident are calculeted with the containment analysis computer program LOCTIC. The program is used to calculate the thermod ynamic state of the containment due to the mass and energy addition to the containment atmosphere. The mass and energy addition rates are input to LOCTIC in tabular form and are discussed in Section 6.2.1.4. CESSAR The main steam pipe break analysis is performed with the mass and energy release rates provided in CESSAR for this accident. In generating these release rates, Combustion-Engineering has assumed that auxiliary f eedwater to the affected steam generator 24 is terminated automaticully. The auxiliary feedwater system provided in SWESSAR-P1 therefore includes this capability. The balance of plant requirement for automatically terminating duXiliary feedWater Will be evaluated in the Utility-Applicant's ( i; _ mm
/ L .)
6.2-12 Amendment 24 4/23/76
SWESSAR-P1 application for a Construction Permit. This feature will be deleted if the analysea show that it is not required. The y evaluation will also be provided in the application for Final Design Approval of SWESSAR-P1. Main steam pipe break analyses are performed at various power levels and break areas in order to identify the worst event. A failure to close the main steam isolation valve (MSIV) in the broken main steam line is the worst single active failure for this postulated event. The MSIVs in the intact main steam lines are assumed to receive an actuation signal when the reactor is scrammed and are assumed to be completely closed 5 seconds later. Analyses performed at three power levels and three break sizes at each power level with the MSIV failure identify the no load, 25 percent break area case as the worst (see Table 6.2.1-8) . The MSIV failure is compared with the failure of a diesel generator to start. Failure of a diesel is less severe, as shown in Table 6.2.1-8. Failure to isolate main feedwater flow from the affected steam genera'.or is not considered because each main feedwater line has a containment isolation valve and a feedwater isolation valve, both outside containment and both safety class valves. There are no check (non-return) valves in the main steam lines. However, the MSIV is capable of tight closure against flow in either direction. The pressure transient for the worst case is shown in Fig. 6.2.1-13, and the mass and energy b1cwdown rates used to calculata it are given in Table 6.2.1-9. The Uchida condensing heat transf er s ~'icient(25) is used in the main steam pipe break analysis. The : asing ocefficient transient is shown in Fig. 6.2.1-23. The containment atmosphere, liner, and concrete temperatures for this accident are shown in Fig. 6.2.1-24. As shown in this tigure, the containment liner and concrete temperatures do not exceed the containment design temperature of 280 F while the atmosphere temperature reaches 416 F. The peak calculated containment liner and concrete temperatures are listed in Table 6. 2.1-16 for the worst main steam pipe break accidenc. The liner temperature is taken at the inside surface while the concrete temperature is taken at the liner-concrete interf ace. The internal conct , ad steel temperatures are Also listed in Table 6.2.1-16. shown in the table, the containment calculated temperatu. .s are within the envelopes allowed by the g codes referenced in Sections 3.8.1.2 and 3.8.3.2 respectively. 6.2-12A //" b Amendment 24 7 4/23/76
SWESSAR-P1 The operation of safety related equipment inside the containment is not expected to be degraded during the main steam pipe break accident due to the short duration of temperatures in excess of the containment design temperature of 280 F. A chronology of events for the main steam pipe break with failure of the MSIV is given in Table 6.2.1-17. RESAR-41 A signiticant amount of liquid carryover is expected to occur during the blowdown of a Westinghouse RESAR-41 Model E steam generator af ter rupture of a main steam line. However, mass and energy release data incorporating liquid carryover for this accident are not yet available from the Vendor. Because the steam generator and the assumed maximum main steam pipe break areas are largest for the C-E plant, the mass and energy release rates to the containment are expected to be highest for the C-E NSSS. Therefore, the C-E release rates are used in the containment pressure analysis for the main steam pipe break accident. A comparison of pertinent data for the Westinghouse RESAR-41 and C-E steam generators is presented in Table 6.2.1-9 (W) . While the Westinghouse Model E steam generator secondary is 26 in. taller than that of the C-E design, the diameter is 68 in. less. This difference in dimensions is expected to result in higher two phase level rise rates for the Westinghouse RESAR-41 steam generator. Thus, liquid carryover is expected to occur earlier for the Westinghouse plant for a given break area. The auxiliary feedwater system for C-E has the capability to automatically isolate the auxiliary feedwater to the faulted steam generator. Tb ensure that the mass and energy releases computed for the C-E NSSS are conservative for the W-41 NSSS, the 24f W-41 auxiliary feedwater systen design incorporates this same automatic isolation capability, as shown in Fig. 7.8.2-15. The main steam pipe break analysis for the W-41 NSSS will be incorporated into SWESSAR-P1 prior to application for Final Design Approval. In addition, the analysis will be submitted in support of the construction permit application for the first Utility-Applicant referencing SWESSAR-P1/RESAR-41. When this analysis is completed, the requirement for the automatic auxiliary feedwater isolation will be reevaluated. This capability will be deleted at that time if the analysis supports that conclusion. (, t _. c - -) 6.2-12B Amendment 24 4/23/76
SWESSAR-P1 RESAR-3S NSSS Following a steam pipe break, a steam-water mixture ot varying quality is released to the containment from the ruptured pipe. Energy is transferred to the steam generator secondary coolant from the primary coolant. After a short time period, energy is transferred into the reactor coolant system as a result of reverse heat transfer in the intact steam generators. The sources of heat transferred to the broken loop steam generator secondary coolant are reactor coolant system sensible heat (i . e . , me tal , coolant, and core sensible heat) , fission product und heavy element decay heat, f2ssion energy generated in the core because of the reactivity increase due to reactor coolunt cooldown, and the het steam generator metal sensible heat. Analyses are performed for four different core power levels based on available mass and enargy release rates computed by Westinghouse with the MARVEL computer cot 2 descri' 4 in KCAP-7909. Three different break areas are analyzed at 102 <na O percent power levels, including DERs. Full DERs ure not analyzed at the 70 and 30 percent power levels. The analyses performed at 0 and 102 percent power, however, envelop these power levels because a conservatively high steam enthalpy of 1,205 Btu /lbm is used. Thus, the stecm for the DERs considered 20 enters the containment superheated by approximately 15 Btu /lbm. The results are tabulated in Table 6.2.1-8. For this unalysis the main feedwater is isolated by low steam line pressure, low pressurizer pressure, or the high containment pressure setpoint of 5 psig (see Fig. 7.8.1-3 and 5) . The MSIVs are isolated by high steam pressure decay rate, low steauline pressure, or the high-2 containment pressure setpoint of 10 psig (see Fig. 7. 8.1-3 ) . All analyses assume a 2.5 second maximum signal processing time. The main feedwater and main steam isolation valves close within 5 seconds after receipt of their respective signals. Therefore, both valves are tully closed within 7.5 seconds after rea ching their respective isolation se tpo ints . No credit is taken for valve closure until 7.5 seconds after the respective setpoint is reached at which time they are closed instantaneously. Auxiliary feedwater flow to the affected steam generator is terminated by operator action at 1,800 seconds after the accident. The conteinment spray system is actuated at the containment pressure setpoint of 40 psia. The time required to start the pumps and fill the headers is 90 seconds. Although the fan coolers are normally operating, for this analysis it is considered that they switch to their emergency mode of operation at the contaanment pressure setpoint of 5 psig plus 30 acconds 2 and are assumed to not remove any heat from the containment while in the normal cooling mode. bb 6.2-12C Amendment 28 8/6/76
SWESSAR-P1 Failures of an MSIV, a containment spray pump, a component cooling water pump, and a diesel generator are considered in the analysis. A failure to isolate main feedwater is not analyzed since each main feedwater line contains two isolation valves with a maximum closure time of 5 seconds. The main feedwater isolution scheme is, therefore, single failure proof and main feedwater flow is terminated when isolation is required. Loss of offsite power results in loss of power to the reactor coolant pumps. The subsequent loss of one diesel generator results in the loss of one motor driven auxiliary feedwater pump. According to the NSSS Vendor 's mass and energy release analysis, loss of the reactor coolant pumps greatly reduces heat transfer in the steum generators that results in lower mass and energy release rates to the containment. The auxiliary feedwater flow rate to the affected generator is also reduced tollowing a diesel generator failure further decreasing the mass of water available for boiloff into the containment. Thus, loss of offsite power following a steam pipe breuk accident is less severe than the case where offsite power is not lost. All combinations of tailures, power levels, and break sizes need 28 not be analyzed. The temperature and pressure are analyzed for the entire spectrum of power levels and break sizes in conjunction with an MSIV failure (see Table 6.2.1-8) . An MSIV f ailure increases the amount of steam released to the containment but both active heat removal mechanisms, fan coolers und sprays, are operable. The two highest computed pressures occurred for the 1.4 ft2 at 102 percent power and 0.4 ft2 at 30 percent power cases. The two highest computed temperatures occurred for, again, 1.4 ft2 and 102 percent power and 1.4 ft2 at 0 percent power. Spray pump failures in conjunction with these three cases are presented on Table 6.2.1-8. The failure of a spray pump results in higher pressures and the same as or lower temperatures as the comparable MSIV failure. Note that the pea % temperatures and pressures do not necessarily occur at the same time. The fan coolers are active in removing energy from the aantainment atmosphere before spray initiation since the fan coolers are started much earlier in the transient. Thus, a f ailure of a train of fan coolers (caused by the failure of a component cooling water pump) can lead to a higher peak tempera ture . Once the full complement of sprays is initiated, however, the latter part of the transient is more effectively controlled than is the case with a spray pump failure. This is illustrated by an analysis of the case that leads to the highest peak temperature for both the MSIV and spray pump failure. Thus a component cooling water pump failure leads to the highest peak temperature when c7mpared with the MSIV and spray pump , failures ,_, 0e; e =. / 6.2-12D Amendment 28 8/6/76
SWESSAR-P1 for the 1.4 fta at 102 percent power break. As shown in ' Table 6.2.1-8, the calculated temperature is slightly higher than the spray pump failure case, and the pressure slightly lower. A failure of a component cooling water pump in conjunction with a 1.4 itz break at 0 percent power is not analyzed since the spray pump failure with this break does not result in the limiting atmosphere temperature or pressure and compor: ant cooling water pump failure, as discussed above, leads to only minor changes in peak temperature and pressure when compared to a spray pump failure. The peak containment atmosphere temperature, therefore, results from a full DER of one main steam pipe at 102 percent power with failure of one component cooling water pump. The other failures considered with the full DER at 102 percent power result in lower temperatures. The full DER break at zero percent power results in a lower temperature than the same failure occurring at 102 percent power. The peak containment pressure results from a steam pipe LDR of 0.40 fta at 30 percent of core full power. This break yields the highest containment pressure because the initiation of tne g containment sprays is delayed by the initial low pressurization rate of the containment. Before initiating the sprays, the containment heat sinks have absorbed sufficient heat to render the heat sinks less effective than for the 1.4 ft2 DER case. In addition, the smaller break results in a slower depressurization of the steam generator secon dary. This in turn results in a longer dryout time for the. smaller break. The net result of the smaller break area and higher secondary side pressure is a higher blowdown rate to the containment after sprays are initiated for the 0.4 ft2 LDR case. This results in a high containment pressurization rate until the auxiliary feedwater to the affected steam generator is terminated manually at 30 min. The LDR of 0.40 ft2 at 30 percent power break results in a higher pressure than would the same break occurring at zero percent power. At zero percent power, breaks larger than 0.2 tta LDR result in moisture carryover. The moisture carryover would cause the containment atmosphere to remain saturated, and mass release would end much earlier so that the sprays, when initiated depressurize the containment more rapidly than for this break at 30 percent power. Also, steam generator dryout would occur earlier. Thus, the containment atmosphere temperature and pressure for this size break ut zero percent power would be less than those computed at 30 percent power. The pressure transients for these limiting pressure and temperature cases are shown in Fig. 6.2.1-13. Comparisons of the computed containment atmosphere temperatures and pressures with the corresponding mass release rates are shown on Fig. 6.2.1-25 t:, J L 0 6.2-12E Amendment 28 8/6/76
SWESSAR-P1 ana 26. The chronology or events for these two limiting cases are given in Tables 6.2.1-17 and 6.2.1-17A. The containment atmosphere, liner, and concrete temperatures are snown in Fig. o.2.1-24 and 6.2.1-24A. As shown in these figures, the temperatures do not exceed the containment aesign temperature of 280 F. The peak calculated containment liner ana concrete temperatures are listed in Taale 6.2.1-16. The liner temperature is taxen at the inside surface while the concrete temperature is taxen at the liner--concrete interf ace. The internal concrete and steel temperatures are also listed in this table. As shown in the table, the containment calculated temperatures are within the envelopes allowed by tne codes ref erenced in Sections 3.8.1.2 and 3.d.3.4. The mass and energy release rates ror the peak temperature and pressure calculations are taoulated in Tables o.2.1-9 ana 9A, respectively. The calculation or these rates is described in Section b.z.1.4.
'1h e Uchida ccadensing heat transter coetf1cient(25) is used in the main steam plee areak analysis. The condensing coefficient translents are shown in Fig. 6.2.1-23.
b-SAR 205 Following a steam pipe break, steam is released to t_he containment from the ruptured pipe. Energy is transferred to the steam generator secondary coolant from the primary coolant. Arter a short time period, energy is transferred into the reactor coolant system as a result ot reverse heat transfer in the intact steam generator. The sources of heat transferred to the broken 39f loop steam generator secondary coolant are reactor coolant system sensinle neat (i.e., metal, coolant, and core senslole heat), rission product and heavy element decay heat, fission energy generatea in the core because of the reactivity increase due 't o reactor coolant cooldown , anc the not steam generator metal sensible heat. Analyses are presented tor three Dreak sizes (i.e., DER, 2.0 It2, 39 and u..$3 It2 oreaxs) at 102 percent power. Since cry superheated st eam (i .e . , no moisture carryover) is released to the containment rollowing rupture ot a main steam line, analysis or otner power levels is unnecess ary . The areak size spectrum considers the delay in isolation or auxiliary feedwater to the afrected steam generator tnat results trom steam line ruptures smaller than a DER. Auxillary teedwater to the a ff ected steam generator is isolated automatically by FOGG (flow only to good generator) logic. For the 0.33 tt2 breax, FOGG logic is assumed 39 to actuate when the steam generator pressure reaches 600 psia resulting rrom the reactor coolant system temperature decreasing to the saturation value for 600 psia. This occurs several minutes after the start of the accident. See Section b.2.1.4 tor LeJ mm / 6.2-12F Amendment 39 7/14/78
SWESSAR-P1 a more detailed discussion or the calculauon or FOGG actuauon time Ior the small creak. For this analysis the main teedwater and main steam are isolated ay low steam generator pressure, low reactor cooiant pressure, or tne high containment pressure setpoint of 4 psig (see Fig.
- 7. d .1-3) . All analyses assume a 2.5 second maximum signal processing time. The main steam and main reedwater isolation valves close within 5 and 15 seconds after receipt or tnelr signals, respectively. Therefore, tne valves are f ully closed witnin 7.5 and 17.5 seconds arter reaching their rese ecuve isolauon setpo11ts. No credit is taken for steam line isolation valve closure unul 7.5 seconds af ter the respective setpoint is reacned; at whicn time they are closed instantaneously. The main 39 reedwater isolation valves are assumed to close in linear rasnion.
Tae containment spray system is actuated at tre containaent pressure setpoint ot 40 psia. The time required to sta rt. tae pumps and till the headers is 90 seconds. Altnough the tan coolers are normaily operating, for this analysis it is consicereo that tney switch to their omorgency mode of operanon at the containment pressure setpoint ot 4 psig plus 30 seconds and are assumed to not remove any heat trom tne containment walle in the normal cooliny mode. Failures or an MS1V to close and the emergency Dus to oecome energized are assumed in the analysis. Since the rallures are assumed to occur simultaneously, tne results ootained for eacn areak size envelop both f ailures. A rallule to isolate main teeawater is not analyzed since each main closure reedwater line contains two isolation valves with a maximum cine ot 15 seconds. The main feedwater isolation seneme 13, therefore, single tallure proof and main feeuwater Ilow is terminated when isolation is requirea. as snown in Taole 6.z.1-8, the limiting pressure results trom a steam line DER. This Dreax results in the fastest uncontrolled ulowaown of tne atfected steam generator to the conta inment. The limiting temperature results from a 2.0 It2 steam line rupture. This Dreax yielos a slightly nigner temperature tnan the b .R wnen an initial pressure of 14.7 psia Ls used . This indicates tnat tne peax containment tempera ture is insensitive to large variations in break area. Auxiliary reeawater is not initiated to the arrected generator tor the rull DER or 2.0 It2 break. The J.33 It2 break is not limiting with respect to pressure or temperature because the mass ano energy rates are lower than 39 those for the Dr.R and 2.0 tte break. p _ su o.2-12G Amendment 39 7/14/78
SWESSAR-P1 Containment pressure transients are shown on Figs. 6.2.1-13 and -13A. 39 The chronology or events for these two limiting cases are given in Tables 6.2.1-17, -17A, and -17B. The containment atmospnere, liner, and concrete temperatures are 39l shown in Figs. 6.2.1-24, -24A, and -248. As shown in these I figures, the liner and concrete temperatures do not exceed the containment aesign temperature of 280 P. The peak calculated containment liner and concrete temperatures are listed in Table 6.2.1-16. The liner temperature is taken at the inside surface while the concrete temperature is taken at the liner-concrete interface. The internal concrete ana steel temperatures are also listen in this table. As shown in the tanle, tne containment calculated temperatures are wirnin the envelopes allowed by the codes referenced in Sections 3.8.1.2 and 3.8.3.2. 39l ss and energy Mease ram W m anQsis ne ddad I in Tables 6.2.1-9, -9A, and -93. The calculation or these ri.tes is described in Section 6.2.1.4. The Uchida condensing heat transfer coef ficient( 2O is used in the main steam pipe break analysis. The condensing coefficient transients are shown in Fig. 6.2.1-23 and -23A. 6.2.1.1.2.3 Feedwater Pipe Break Inside Containment Structure Tne reedwater pipe break is not as severe as the main steam pipe nreak since the break effluent is at a lower specific entnalpy. The reedwater pipe break analysis is perrormed at the hot standby condition since the steam generator secondary mass inventory and pressure are a maximum at tnis power level. A DER of one main ieedwater pipe is assumed in the analysis. An MS1V tallure is assumed in the reedwater pipe break analysis. The main feedwater isolation t2mes are given in Table 6.2.1-12. The calculated peak pressure following a feedwater pipe DER accident is given in Table 6.2.1-8. The resultant containment pressure transient is shown in Fig. 6.2.1-13. The main teedwater pipe break analysis for the CESSAR NSSS envelops this accident for the B-SAR-205 NSSS. The CESSAR NSSS utilizes steam generators much larger than those of the 6-SAR-205 NSSS with main feeawater steam generator nozzles of the same size. Theretore, the total mass and energy released to the containment for this accident is greater for the CESSAR NSSS, and the calculated containment pressure is higher than that for the b-SAR-zuS NSSS. Thus, a feedwater pipe break analysis is not provided ror the B-SAh-205 NSSS. 4 9 6 / ,t 6.2-12H Amerrient 39 7/14/78
SWESSAR-P1 b.2.1.2 Suncompartment Analysis 6 . 2 .1. 2 .1 Subcompartment Analysis Procedure Two types of pressurization conditions are considered in the subcompart mnt analysis. The rirst is the development or a uniform pressure over the entire wall. The second is tne development of a pressure profile over the wall. Tne pressure profile will result in pressurization or a portion of the wall to a higher pressure than the uniform pressure. 'Ihe remaining portion would then be pressurized to a pressure walch is less tnan the uniform pressure. This is consistent with the laws or conservation or mass and energy. The uniform pressure model is used to calculate the maximum uniform pressure on the subcompartment walls. A one-noce model is used to represent the volume enclosed by the walls since tnis yields the unirorm pressure to which the walls must be designed. In some subcompartments, obstructions to tlow (e.g. , grating in tne steam generator compartment or the hot and cola legs in tne reactor cavity) may resu3 t in a local pressure which is higher than the uniform pressure. For these cases, the volume a broken into as many nodes as there are significant Ilow resistances. The pressure prot 11e tor this calculation consists of the pressures in each node enclosed by the walls at the time of calculated maximum local pressure in the volume. in some cases a cubicle other than the cubicle containing the ruptured pipe must De analyzed ror pressure bu11 dup. In these cases a conservatively high vent flow model (full trictionless Moody flow) is used to predict the flow into the cubicle to be analyzed. The subcompartment analyses are pertormed with the THREED computer code and the RELAP4 computer code. THB an accounts for two phase, two component (steam-water-air) flow through the vents, but does not consider the inertial ef f ects whicn may be sign 1ricant at low and intermediate pressures. The REIAP4 code consloers the inertial erfects but does not include the ef f ect of air mixing with the flowing steam-water mixture. The higner of the two results (THREED or RELAP4) is presented in Table b .2.1-3 with identification of the code used. 6.2.1.2.2 break Type Definitions and Areas Two types of breaks are used to analyze containment subcompartments. The first type is a guillotine break. A guillotine break which results in a break Ilow area of two pipe cross sections is called a double ended rupture (DER) . In some suncompartments, pipe restraints 11mit the displacement of the two broken ends of the pipe so that the break flow area is less than two pipe cross sectional areas. This type break is called a limited displacement rupture (LDR) . The special case of an LDR 6.2-121 Amendment 39 by) [}] 7/14/78
SWF.SSAR-P 1 or one pipe cross sectional area is called a single ended rupture (SER) . Ite second type or break is a longitudinal split which is equivalent to a hole in the wall of tne pipe. A split which results in a break flow area or one pipe cross section 1s called a single ended split (SES) . Tae breax type (s) and area (s) to which the subcompartment walls must. De designed are listed in Table 6.2.1-3. All nreaxs analyzed witnin a particular subcompartment are descrlhed in Section t> .2.1.2.7. Pipe restraints are provided to limit the Dreax areas to those analyzea. Tne derinition or the limiting breax type (s) , location (s) , and area (s) which must ce considered in the sunccxnpart men t analyses is within the ILSSS Vendor's scope, as defined in Table 1.8-1 and descrined in Section 3.6.2.1. The NSS Vendors nave provided break type (s) , location (s) , area (s) , and blowdown cata in timir saks rereienced in Section 1.6 and/or the topical report.s reterenced in Table 3.t>-3. In general, the breaks discussed in the SaRs are not mechanistically determined. The topical reports descrine criteria and analytical methods used to establish the discrete Dreak locations, orientations, types, and opening times postulated for the primary coolant system piping. This methoaology is used in con 3 unction with the piping stiffness and local mechanical restraints to define tne breaks to be considered in an applicant's SAR. In particular, the nreaks derined ror the reactor cavity and the steam generator subcomparment analyses are mechanistically determined considering the pipe restraints. The breaks considered in the SWESSAR-P1 subcompartment analysis are in accordance with the breaxs specified by the NSSS Vendor. A Du is ut211 zed in the analysis for the pressurizer cubicle. nreaxs with less than two cross-sectional Ilow areas are used in tne analyses for the reactor cavity and steam generator suncompartments. A summa ry or these breaks is given in Tab:e 6.2.1-18. i e.cs
' O a) 6.2-12J Amendment 39 7/14/78
SWESSAR-P1 6.2.1.2.3 Subcompartment Arrangement Detailed plan and section drawings of the containment subcompartmonts are provided in Fig. 1.2-3. The figure shows the arrangement of structures and components within the containment. 6.2.1.2.4 Obstructions to Vent Flow No credit is taken for movable obstructions to vent flow in the subcompartment analyses. All insulation is assumed to renain in place in the calculation of subcompartment volumes and vent areas. 34 The vent areas provided in the subcompartments are sufficiently large to preclude any significant blockage of flow from the subecmpartments following a high energy pipe break. Pipe and component insulation (except insulation near the break location) will remain intact since it is secured. 6.2.1.2.5 Vent loss Coef ficients 'Ihe vent loss coefficients used in the subcompartment analysis depend on the geometry of the particular vent. The basis for the coefficients is the Handbook of Hydraulic Resistance.(s) 6.2.1.2.6 Subcompartment Analytical MMel Description of the THREEn Code The THREED computer code is used to calculate pressure and temperature transients in various nuclear power plant subcompartments following a postulated high energy pipe break. THREED allows the user to subdivide all cubicles into nodes in order to take into account all known major flow obstructions. The output is used mainly for design purposes in establishing the peak pressure differentials across the subcompartment walls. Assumptions The derivation ot the analytical model is based on the laws of conservation of mass and energy, the equation of state of air, steam, and water, and the principles of two phase flow. In order to approximate this problem numerically, the following simplifying assumptions are made:
- a. Adiabatic Process The system is defined as the compartment atmosphere at any given time. This includes any air, steam, and water droplets present, but not the walls, equipnent, or internal structure of the compartment itself. Heat
,77 n ' is g ;) j L J 't 6.2-13 Amendment 34 7/22/77
SWESSAR-P1 sinks which may exist within the compartment are neglected.
- b. Quasi - Steady State Mass and energy flows are calculated on the basis of the node thermodynami c state, as determined at the end of the previous time interval. The thermodynamic state is determined from mass and energy flows during the time interval based on flow rates evaluated at the beginning of the time interval.
C. Complete Mixing in the Node The atmosphere in the node mixes instantaneously and
!xxnogeneously. At each point in time the atmosphere is in a state of thermodynamic equilibrium.
- d. Independent Inflow The mass and energy inflows are independent of the compartment back pressure. This assum 3 ion allows the mass and energy inflows to be specif.i ed as input to the 34 Program, and it is accurate for thon" time periods when the flow through the pipe break is sonic. For subsonic break flow, it is conservative with respect to pressure buildup in the compartment. 'Ihe inflow may be divided amongst as many nodes as is appropriate to the configuration under investigation.
Computational Methods THREED numerically solves finite difference equations which account for mass and energy flows into and out of a node. The computational approach used in THREED is summarized in the conceptual flow chart shown in Fig. 6.2.1-20. Equations
- a. Mass and Ener(7y Release Rates from a High Energy Pipe Break The rates of mass and mergy release from a high energy pipe break are supplied as input to THREED. These blowdown rates are obtained from the NSSS Vendor. The blowdown model for subccxnpartment design is discussed in the NSSS Vendor's SAR.
bb 6.2-14 Amendment 34 7/22/77
SWESSAR-P1
- b. Calculation of the 'Ihermodynamic State of a Node At the end of each time interval of the numerical calculatica, the stagnation temperature and pressure in 34 the node are determined based on new inventories of mass and internal energy.
6.2-14A , , , ,,, Amendment 34 OOJ LJU ]/22/))
SWESSAR-P1
- c. Internadal Flow Rates The TW2EED computer code inclu :cs three two phase vent flow options. The first two options consider the vent flows to be homogeneous. For vents with a cont ra ction at the inlet, an isentropic entrance effect is included in the Homogeneous Vent Flow Model Nunber 1 (HVFM-1).
When there is no contraction the isentropic entrance effect is not appropriate and the correct option is Homogeneous Vent Flow Model Number 2 (HVFM-2). The third flow model is the frictionless Eoody flow model. This model is used for design of compartments which do not contain high energy lines, but which are adjacent to compartments which do. In this case, flow to the compartment under consideration from the source nodes is considered to be frictionless Foody flow with a multipler of 1. The approach used in the derivation of the homogeneous vent flow models (HVFM) is a modification and extension of that presented in Reference 16. The major assuruptions used in deriving the models are as follows:
- 1. The flow is quasi-steady state.
- 2. The flow is one-dimensional,
- 3. The flow is homogeneous (no slip between phases) .
- 4. The flow is adiabatic (no heat transf er between the vent and the fluid or between the fluid phases) .
- 5. The quality is constant over the length of the vent.
- 6. The air-steam mixture is considered as a perfect gas.
- 7. The sonic velocity of the mixture is equal to the sonic velocity of the gaseous phase only. The basis for this assumption is experimental data which show that, for the high void fractions (2. 9 5) which are expected to occur in subcompartment analys es, this is true (10, 11, 15).
- 8. Pres su re changes within the vent due to gravity effects are negligible.
- 9. The flow is accelerated isentropically from the source node to the vent inlet (HVFM-1 only) .
6.2-15 Amendment 10
., , - ,. 5/15/75 beJ m J/
SWESSAR-P1
- 10. The flow rate becomes critical when the exit Mach 101 number of the gaseous phase becomes unity.
- 11. The vent pressure loss coefficient (K) is constant during the time interval.
- 12. The vent flow area is constant during a time interval.
Solving the equations of state, continuity, energy, and mcanentum consistent with the above assumptions leads to the following ecuations: 2 - 2 M 2 -M M X 3 Y+1 2 2 3 2 2Y y,Y 2 g z2 1
- M ,2 M _ M 2 2X 3
~
1/2 P o2 1 2 + Y-1 X M 2 M 1 1 2 p (HVFM-1) 01
\y + y_1 2 2 M -
1 + Y-1 2x M 2 X 2 1)Y-1 _ j _ P O 02 = 2 + "1 x Mf Mf (HVFM-2) 01 2 + 7 -1 M 2 g2 where: P01 = stagnation pressure in the source node, psia. P 02 = stagnation pressure in the sink node, psia. X = quality of the mixture. 7 = air-steam mixture specific heat ratio Mj = Mach number of the gaseous phase at the inlet to the vent. M = Mach number of the gaseous phase at the exit of the vent. 2 K = Vent pressure loss coefficient. The K-factor used in THREED includes the contraction, bend, and skin friction losses which are encountered in flow through the ven t . The vent exit loss is implicit in the calculation scheme 10 of THREED. In the calculation of the thermodynamic state, the flow velocity within each node is assumed to be zero. Thus, one full velocity head is lost at the exit of the vent. This is equivalent to a X-factor of 1 at the exit of the vent. t "n s ) .:D 6.2-16 Amendment 10 5/15/75
SWESSAR-P1 This system of nonlinear equations is solved for M j and M2 using a Newton-Raphson interation technique. If M 2 21, the flow is assumed to be choked. When M2 >1, quation (1) is solved for M; with M 2 set equal to 1. Once M2 has been determined, the mass flow rate per unit area is determined by the following equation. where: 12M Yp g pg g v2 G= X 1 + Y-1 M 2 1
/ Y-1 3 + M i NU 2X 1 2X 1 3
12M - - 1/ 2 G= 7A P g (HVFM-2) x 01 ol c G = maas flow rate per unit area, lbm/fta sec Pg = dencity of Vapor in the source node, lbm/f t3 gc = 32.1739 lbm - ft/lb -sec2 Isentropic ExponentY The liquid effects are considered in the momentum ecuation for calculating acceleration and local irreversible losses (16 ) . The specific heat ratio is based on an air-steam mixture and flashing is not allowed in the vents, since the residence time in the vent is asually too short to allow flashing. Thus, the composition in each phase remains constant and only the vapor phase increases in volume. The value of 7 ranges from 1.1 to 1.4. Le , .> ) 6.2-16A Amendment 10 5/15/75
SWESSAR-P1 6.2.1.2.7 Containment Subcompartment Analysis Results The design pressures are 1.4 times the peak calculated pressures and are listed in Table 6.2.1-3 for each subcompartment. Pressurizer Cubicle The pressurizer cubicle is analyzed according to the nodalization schematic of Fig. 6.2.1-3. A spray line DER at the top of the cubicle is considered in the pressurizer cubicle analysis. In addition, a surge line DER is analyzed for the W-41, C-E, and BSW NSSS. The surge line break occurs inside the pressurizer support skirt for the W-41 and C-E NSSS and inside the surge line compartment for the BSW NSSS. The W-3S and B&W pressurizers do not utilize the support skirt design. The W-3S pressurizer is supported directly on the floor. Therefore, there is no venting to the cubicle for the W-3S NSSS. The B&W pressurizer is supported from the cubicle walls, thus allowing surge line venting to the cubicle through the annulus space between the pressurizer and floor. The mass and energy release rates for these breaks are given in Tables 6.2.1-6 and 6.2.1-7. The vent areas, K-factors, and flow models for the THREED analysis are listed in Table 6.2.1-14. Data for RELAP4/ MOD 3 analysis are listed in Table 6.2.1-14A. In order to analyze the surge line DER inside the pressurizer skirt for the W-41 and C-E NSSS and the surge line compartment for the B&W NSSS, the flow to the cubicle is conservatively assumed to be frictionless Moody flow. The pressure response for the pressurizer cubicle is shown in Fig. 6.2.1-4 for the W-41 and C-E NSSS spray line DER, and in Fig. 6.2.1-5 for the W-41 surge line DER. 36 The peak calculated and design differential pressures for the pressurizer cubicle are given in Table 6.2.1-3. The time of peak differential pressure can be read from Fig. 6.2.1-4 and 6.2.1-5. n n [vj L1d 6.2-17 Amendment 36 12/21/77
SWESSAR-P1 Pressurizer Support Skirt (W-41 and C-E Only) The pressurizer support skirt is also analyzed according to the nodalization schematic shown in Fig. 6.2.1-3 for the W-41 and C-E NSSS. As described above, Stone & Webster does not supply a support skirt for the W-3S and BSW NSSS. A surge line DER inside the pressurizer support skirt is considered in the analysis. The vent areas, K-factors, and flow models for the THREED analysis are listed in Table 6.2.1-14. The resulting pressure response for the skirt is shown in Fig. 6.2.1-5. The peak calculated and design pressure differentials are given in Table 6. 2.1-3. The times of peak differential pressure can be read from Fig. 6.2.1-5. Pressurizer Relief Tank Compartment (W-41, C-E , W-3S) and Surge Line Compartment (B&W) This compartment is analyzed according to the nodalization schematic of Fig. 6.2.1-3. A surge line DER is assumed to occur inside the ccrnpartment in the analysis. The vent areas, K-factors, al l flow models for the THREED analysis are listed in Table 6.2.1-14. The node length-to-area ratios for the RELAP4 analysis are listed in Table 6.2.1-14A. The pressure response for the compartment is shown in Fig. 6.2.1-
- 6. The peak calculated and design differential pressures are shown in Table 6.2.1-3. The time of peak differential pressure can be read from Fig. 6.2.1-6.
The madel used to compute the pressure in this compartment for RES AR-41 is currently undergoing a generic review by the NRC. If, in resolution of this item the S&W calculated pressure increases over that reported in Table 6.2.1-3, the subcompartment design for the RESAR-41 NSSS will accommodate the resultant design pressure. Stean Generator Comnartment Below Operating Floor The nodalization schematic used in the steam generator com-partment is shown in Fig. 6.2.1-2. The NSSS breaks analyzed in the compartment are in the scope of the NSSS Vendor. Pipe movement, in some cases, is restricted by mechanical pipe restraints. The maximum break areas are given in Table 6.2.1-3 and in the respective figure titles for the pressure re..,mnse inside the compartment. 6.2-18 Amendment 20 1/23/76 (U O .] L 'Y
SWESSAR-P1 The vent areas, K-factors, and flow models for the THREED analysis are listed in Table 6.2.1-15. W e node flow areas and length-to-flow area ratios for the RELAP4 analysis are listed in Table 6.2.1-15A. We pressure response for the " worst cold leg break" and " worst hot leg break" are shown in Fig. 6.2.1-7 and 6.2.1-8, respectively. The mass and energy release rates for the most severe break are given in Table 6.2.1-5. Steam lines are not routed through any portion of the compartment and are not considered in the analysis. A feedwater line DER results in a lower pressure than that for M e RCS break presented in Table 6.2.1-3 due to the lower init 'l temperature and pressure in the feedwater line. The peak calculated and design pressure differentials are given in Table 6.2.1-3. The time of peak differential pressure can be read frcxn the pressure response curves of Fig. 6.2.1-7 and 6 . 2 .1 --8 . , CESSAR and RESAR-3S NSSS
%e peak for the cubicle walls is considered to be the highest g calculated pressure differential in the steam generator compartment below the operating floor. Thus, a separate uliform pressure analysis for the campartment is not performed.
Steam Generator Shield Wall Above Operating Floor his silield wall is analyzed for pressure buildup according to the nodalization schematic of Fig. 6.2.1-2. We vent areas, K-factors, and flow models for the THREED analysis are listed in Table 6.2.1-15. The pressure responses for the shield wall due to the worst cold leg break and worst hot leg break are shown in Fig. 6.2.1-7 and 6.2.1-8, respectively. The peak calculated and design pressure differentials are given in Table 6.2.1-3. The time of peak presmire differential can be read from Fig. 6.2.1-7 and 6.2.1-8. RESAR-81 NSSS There are no high energy lines within this compartment. However, the shield wall may become pressurized due to a high energy line break below the operating floor. f7 (D b j l 'r L 6.2-18A Amendment 21 2/20/76
SWESSAR-P1 Flow into the shield wall region above the operating floor i rom the at eam generator compartment below the operating iloor in assumed to be frictionless Moody flow. B&W NSSS A hot leq DER is postulated in the steam generator shield wall reqion above the operating floor for the B&W NSSS. CESSAR and RESAR-3S NSSS The peak pressure for the cubicle walls is considered to be the highest calculated pressure differential in the steam generator 2" compartmen+ below the operating floor. Thus, a separate analysis for this compartment is not performed.
<r nr D0J c 't J 6.2-19 Amendment 24 4/23/76
SWESSAR-P1 Volume Below Steam Generator Support Pedestal RESAR-41 and BSW NSSS The nodalization schematic for this analysis is the same as that used in the steam generator compartment and shield wall analysis (Fig. 6. 2.1-2) . This volume may become pressurized following a high energy line break above the steam generator support pedestal. The design pressure for the volume is based on full frictionless Moody flow into the volume. The vent areas, K-factors, and flow models for the THREED analysis are listed in Table 6.2.1-15. The pressure response for the worst break (see Table 6.2.1-3) is shown in Fig. 6.2.1-7A. The peak calculated and design pressure differentials are listed in Table 6.2.1-3. The time of peak differential pressure can be read from Fig. 6.2.1-7A. CESSAR and RESAR-3S NSSS The peak pressure for the cubicle walls is considered to be the highest calculated pressure differential in the steam generator compartment below the operating floor. Thus c a separate analysis for this compartment is not performed. Volume Above In-core Instrumentation Drive Room (RESAR-41 Only) This volume may become pressurized following a high energy line break in the steam generator compartment. The nodal schematic for the analysis is shown in Fig. 6.2.1-2. Flow into this volume is assumed to be full frictionless Moody flow. The vent areas, K-factors, and flow models for the THREED analysis are listed in Table 6.2.1-15. The pressure response for the worst break is shown in Fig. 6.2.1-7A. The peak calculated and design pressure differentials are listed in Table 6.2.1-3. The time of peak pressure can be read from Fig. 6.2.1-7A. Reactor Cavity The design of the rea ctor vessel support shield tank and the reactor vessel insulation prevent venting downward below the top of the shield tank. This arrangement is shown in Fig. 5.5.14-1. i,~ i 'l 6.2-20 Amendment 20 1/23/76
SWESSAR-P1 A detailed view of the insulation installation is shown in Fig. 5.5.14-1. Thus, the reactor cavity analysis considers pressurization of the upper cavity only. For guillotine breaks, the primary shield wall pipe penetration through which the broken pipe passes is assumed to be blocked by the broken pipe. The upper reactor cavity is analyzed according to the nodalization schematic shown in Fig. 6.2.1-1. The minimum number of nodes required to predict the peak local pressure is determined by performing a nodalization sensitivity study. The results of the study are shown graphically in Fig. 6.2.1-21 (W) . The number of nodes plotted is the total number of nodes within the reactor cavity. The design nodal configuration for the reactor cavity enploys a vertical plane through each reactor vessel nozzle centerline. Thus, the number of circumferential nodes is equal to the number of reactor vessel nozzles. A horizontal plane is also passed through all reactor vessel nozzle centerlines. This plane defines the node boundaries in the axial direction. Thus, the total number of nodes in. aide the reactor cavity is two times the number of reactor vessel nozzles. With this configuration, all node boundaries are placed at the minimum flow area available for internodal flow. Analyses have shown this to be the most conservative configuration for reactor cavity pressurization calculations. Since the geometry is very shnilar, this nodal model is used for all four of the NSSS upper reactor cavity analyses. The ver.t flow models, K-f a ctors , and vent areas used for the THREED analysis are listed in Table 6.2.1-13. Data for analysis with RELAP4/ MOD 3 for the B&W NSSS are listed in Table 6.2.1-13A. 30 The breaks used in the design of the reactor cavity and the resultant peak calculated and design wall differential pressures are given in Table 6.2.1-3. The mass and energy releases for the most severe break are given in Table 6.2.1-4. The mass and energy releases for the hot leg LDR areas are gisen in Table
- 6. 2 .1 -4 A pi--3S and BSW only) . 30 The pressure responses of the reactor cavity are ...own in Fig.
6.2.1-9 for a pump discharge break and in Fig. 6.2.1-10 for a hot leg break. The pressure profile around the reactor vessel at the time of peak asymmetric loading of the vessel is shown in Fig. 6.2.1-11 and 6.2.1-12 for a pump discharge break and a hot leg break, respective ly. The azimuthal position of the peak differential presssure within the reactor cavity is not critical. The biological shield wall design allows the design pressure profile (presented in Table 6.2.1-3) to be rotated to any angle.v nrr @ buJ LtJ 6.2-20A Amendment 30 1/28/77
SWESSAR-P1 26 Nodalization Sensitivity Study - RESAR-41 NSSS Reactor Cavity A total of seven inside the upper reactor cavity. nodal configurations are analyzed
- different The different configurations consist ot 1, 4, 8, nodels. The bulk containment and 48 is nodes and three different 16-node represented by an additional node in these analyses. A 15 0 in . 2 pump discharge LDR is postulated in the nodalization study.
The one-node model considers the upper pressurization to be uniform. This model is us ed reactor to cavity the unitorm pressure differential calculate on the primary shield wall above the reactor vessel support shield tank. The four-node model considers the upper reactor cavity to be circumf erentially divided with node boundaries placed vertically at alternate reactor vessel nozzle centerlines. The resultant pressure is significantly higher than the one-node model value (see Fig. 6. 2.1-21) . The eight-node model is an extension of the tour-node model with the node boundaries placed vertically at each reactor vessel nozzle centerline. The resultant pressure again increases. Three 16-node models are analyzed. nodalization is increased to 16 nodes When the circumferential with an additional 8 vertical boundaries placed midway between those for the 8-node case, the resultant pressure decreases value. from the 8-node .nodel The second 16-node model is similar to the 8-node model described above with the reactor the addition or a horizontal plane which contains all vessel nozzle centerlines. Thus, there are 8 circumferential the plane nodes below this plane and an additional 8 above This model results in a significant increase in the calculated peak pressure from that of the 8-node model. This is the design model for the reactor cavity. The third 16-node is similar to the preceding except that tne horizontal model plane passes through the reactor cavity horizontally RV nozzle centerline. The resultant pressure decreases below 45 in, above the that calculatec for the 8-node model. The 48-node model consists or 16 vertical plane boundaries placed circumferentially as the first addition, horizontal plane boundaries 16-node model described above. In are defined by a plane which through contains the all RV nozzle centerlines and a plane which passes reactor cavity 45 in. above the RV nozzle centerlines. The re;ultant peak pressure predicted in the design case. decreases from that O f: > ,,0 6.2-20B Amendment 26 6/2/76
SWESSAR-P1 Nodalization Sensitivity - Subcompartments Other Than the RESaR-41 Reactor Cavity a nodalization sensitivity study was performed for the RESAR-41 reactor cavity as described above. Fig. 6. 2.1-21 (W) shows a large sensitivir.y of peak pressure differential to the number or nodes inn ide the RdSAR--41 reactor cavity until the number of nodes is increased to the number of physical restrictions to vent flow between noces represented by the reactor coolant legs. As can be seen from the tigure, an increase in the number of nodes beyond 16 ( the numoer of phy sical restrictions in the cavity model derived from the 8 reactor coolant pipes and nozzles) does not result in an increase in the calculated peak pressure. A generalization drawn I r o:n this study is the basis for the nodalization ror all other subcompartment analyses in SWESSAR-P1. That is, all significant restrictions to vent flow within a subcompartment have been included as nodal boundaries. Theretore, nocalization sensitivity studies in subcompartments other than the RESAR-41 reactor cavity are not expected to signiricantly alter the reported results. Asyrnetric Pressure loads on Maior Component Supports Asymmetric presture loads which may exist following rupture of a high energy pipe inside containment will be developed from ap- 34 propriate nodalization models. These will be provided to the NSSS Vendor for support design as stated in Table 5.1-1. Blowdown to Reactor Cavity RdSAR-841, hES,m-3S and n-SAR-205 NSSS Tne blowdown trom a guillotine type break inside the reactor cavity is assumed to divide equally between the four surrounding noces. CESSAR NSSS The blowdown from a guillotine type break inside the reactor cavity is assumed to civide equally between the four surrounding nodes. The CESSAR NSSS utilizes a biological shield wall equipped with guard pipes in the pipe annull to protect the reactor cavity 6.2-20C ' 7 7 '7 Amendment 34 b'. u) -r/ 7/22/77
SWESSAR-P1 walls from high local "ressures that may develop from a postulated split type Dreak. The guard pipe is discussed in Section 3.8.3.1. 'Ihe split is an ellipse, with a major axis equal to twice the inside dian.eter or the pipe, placed along the pipe axis. The guard pipes extend tully into the reactor cavity, completely covering that portion or the break within the reactor cavity and are designed to withstand the maximum reactor coolant system pressure during normal plant operation. The areak area tor the hot leg split is distributed as rollows: 16 percent or the area is located in the steam generator coa.partment and 84 percent is within the guard pipe. One-halt the blowdown into the guard pipe is assumed to flow toward the reactor cavity. The other half is assumed to flow to the bulk containment. 'Ih us , 42 percent of the blowdown riows to the reactor cavity and 56 percent flows to the bulk containment. The blowdown to the reactor cavity is assumed to flow unirormly trom the guard pipe to each node. Thus, 10.5 percent or the blowdown rlows into each node around the guard pipe. The break area tur the cold leg split is completely covered by the guard pipe. As for the hot leg, one-half the blowduwn into the penetration and guard pipe is assumed to flow toward the reactor cavity and one-half is assumed to flow to the bulk containment. Again, the blowdown from the guard pipe to the reactor cavity is assumed to flow uniformly to each node around the guard pipe. Thus, 12.5 percent of the blowdown is addeo to each node around the pipe. The RESAR-41 reactor cavity study was based on a pump alscharge leg LDR. The blowdown flow from this break is evenly distributed to the four nodes that surround the pipe. The limiting break for the CdSSAR NSSS is a split in the hot leg. However, since guard pipes which completely cover the portion of the split within the reactor cavity are used, the blowdown flow is equally distributed to the tour : modes that surround the pipe, as discussed above. Theretore, the rlow distribution to the reactor cavity is similar to that used in the RESAR-41 sensitivity study and thus the generalization made under "Nodalization Sensitivity", i.e., that nodal boundaries placed at the obstructions to flow result in the peak calculated pressures, is valid for the CESSAR NSSS reactor cavity design. Thus, a 12 node model is used for the CESSAR NSSS reactor cavity analysis. As noted on Table o.2 1-18, the analyses for those breaks will be updated to the then current criteria in the application for a Construction permit ny a Utility-Applicant referencing SWESSAR prior to Final Design approval, and on the SWESSAR application for Final Design Approval. g
- ~ ,, W beJ cyj 6.2-20D Amondment 34 7/22/77
SWESSAR-P1 Primary Shield Wall Pipe Penetrations RESAR-41, RESAR-3S, and B-SAR-205 NSSS There are no breaks postulated in the primary shield wall pipe penetrations. The penetrations are conservatively designed to withstand the maximum design pressure within the upper reactor cavity . CESSAR NSSS The pipe penetrations are fitted with guard pipes that are designed to withstand the maximum RCS pressure. 6.2.1.3 Mass and r.nergy Release Analysis for Postulated Loss -ot-Coolant Accidents Containment Analysis 'Ihe LOCA mass and energy release data for the containment analyses are supplied by the NSSS Vendor. Section 6.2.1.1.1 and appendix 6B describe the interface between the NSSS Vendor data and the LOCTIC computer code. Table 6.1-4 lists the references for the data used in the containment analysis for each NSSS. Subcompartment Analysis The IDCA mass and energy release rates for subcompartment analyses are also supplied by the NSSS Vendor. The computer codes used by the various Vendors to calculate these release rates are as follows: Vendor Computer Code Babcock S Wilcox CRAFT Combustion Engineering CEFLASH-4A Westinghouse SATAN V, SATAN VI 'Ihe mass and energy releases for the design of each subcompartment are given in tabular form in Section 6.2.1.2.7. 663 249 6.2-20E Amendment 34 7/22/77
SWESSAR-P1 6.2.1.4 Mass and Energy Release Analysis for Postulated So m dary System Pipe Ruptures Main Steam Pipe Break Accident CESSAR We mass and energy release rates from the steam generator secondaries are in the scope of the NSSS Vendor. The applicable tables in the NSSS Vendor's SAR are listed in Table 6.1-4. Mass and energy release from the main steam piping volume is also included. The main steam piping volume is treated as a high pressure tank in the analysis. This assumption neglects the Ilow resistance represented by the main steam piping. Flow from the main steam piping volume is calculated with the frictionless Moody flow model. The flow is assumed to be constant at its initial value until the piping inventory is exhausted. The mass of water in the unisolated portion of the feedwater line of the affected loop is completely exhausted to the containment following depressurization of the affected spam generator. Mass and energy relea.ce rates and integrated steam pi.pe break analysis are tabulated in Table 6.2.1-9. RESAR-3S ne steam generator mass and energy release rates utilized in the steam pipe break analysis are in the scope of the NSSS Vendor. A significant amount or liquid carryover is expected to occur during the blowdown of a Westingbouse RESAR-3S steam generator dfter rupture of a main steam line. Analyses have shown that the inclusion of moisture carryover results in approximately a 50 F reduction in the calculated peak containment atmosphere temperature. The NSSS Vendor, however, has not yet subnitted to the NRC tor its review unss and energy release rates incorporating moisture carryover and so this analysis is based on dry steam only. This main steam pipe break analysis will be updated to include moisture carryover when the NRC has accepted the NSSS Vendor 's metnoa and data. The dry steam rele.ases used in this analysis are generated with the MARVEL code described in Rererence 29. The topical report to be suhaitted by the Vendor will detail the calculations perf ormed to tailor a standard set of blowdown tables to a particular balance of plant such as SWESSAR-P1. a brief description of the method, as specified by the Vendor, follows. Ebr guillotine breaks, the steam flow from the affected steam generator side of the break (forward flow) is considered separately from the turbine plant and intact steam generator side of the break (reverse flow) . The sum of the forward and reverse flow rates is the total flow rate to the containment. q f au; _su 6.2-20F Amendment 34 7/22/77
SWESSAR-P1 For steam pipe DERs, the main steam and turbine plant piping steam inventory enters the containment at the Moody mass flow rate for saturated steam based on the initial secondary system pressure. This mass rate is held constant until all inventory upstream of the turbine stop valves (TSV) and ux>isture separator reheater isolation valves (MSRIV) back to the steam generator nozzles is exhausted (total piping volume = 11,940 f t3) . After the turbine plant 11.ventory is exhausted to the containment, reverse Ilow to the containment begins immediately from the intact steam generators and continues until closure of the main steam isolation valves (MSlV) at which time it is instantaneously terminated as specified by the Vendor. For steam pipe LDRs, failure of the MSIV in the broken loop to close results in the release of the broken main steam pipe and turbine plant steam inventory following isolation of the Ilow from the irtact steam generators. The initial rate of turbine piping blowdown is equal to the total flow rate in all intact loops just before isolation (assumed to be instantaneous) . The mass rate is linearly ramped to zero in a time interval sufficient to exhaust the entire inventory at the average rate (total piping volume = 10,497 f t3) . For split type breaks, the main steam and turbine plant inventory is added to the steam generator secondary inventory and has the ettect of increasing the steam generator dryout time. The dryout time is the time at wnich the broken loop forward flow rate f alls celow the auxiliary feedwater flow rate or the time at which the steam generator boils dry, whichever comes first. For break stzes other taan DERs where the MSIV in the broken loop does not tall, the intact main steam pipes and turbine plant piping inventory is success f ully isolated and does not enter the containment. The steam inventory in the unisolate portion of tne broken steam pipe is exhausted to the containment following MSIV closure (total piping volume = 481 ft3). Additional mass and energy flow from the affected steam generator to containment results from liquid flashing from the unisolated portion of the main feeowater pipe, pumped main feedwater, and auxillary feeowater flow before isolation. The main f eedwater flow rate to the af fected steam generator is conservatively assumed to be at system runout. All main teedwater rlow is directed to the broken loop. No credit is taken for reduced reedwater flow due to reduced power levels. The analysis is tnerefore conservative as a considerable amount of water mass is anded to the affected steam generator by the main reedwater pumps before isolation (e.g., 30,000 to 40,000 lba). The time or main reedwater isolation depends on the particular nreak conditions as shown in Table 6.2.1-8. The auxiliary feedwater Ilow to the affected steam generator is assumed to be at the system runout flow of 190 ltrn/sec. Tne 6.2-20G Amendment 34 6 (, f ta'~1, 7/22/77
SWESSAR-P1 auxiliary flow to the broken loop is isolated manually by the operator at 30 minutes. The initial inventory of the affected steam generator is increased to account for the liquid in the unisolated portion of its main reedwater pipe. The enthalpy for the two DEks analyzed is taken as 1,205 Btu /11xn. The enthalpy for the remaining breaks is that calculated by MARVEL. The resultant mass and energy release rates for the steam pipe break the accident are tabulated in Tables 6.2.1-9 and 6.4.1-9A for limiting containment atmosphere temperature and pressure cases, respectively. B-SAR 205 The mass and energy release rates to containment are based on B-SAR-265 methodology. The data provided in 3--SAR-20 5 Taales 15.1.14-6 and 15.1.14-7, Amendment 15, are Ltilizeo to determine the time that the containment pressure reaches '5 psig for a 0.22 and 2.0 ft2 steam line break, respectively. The mass and energy release rates are then recalculated by Babcock & Wilcox with the efrects of secondary system isolation and reactor trip included. The full DER is analyzed with the mass and energy release rates provided in B-SAR-205 Table 15.1.14-2, Amendment 15. The mass and energy release rates for a full DI.R (pressure limiting case) are listed in Table 6.2.1-9. 'Ihis is conservative since the total mass release f rom the B-SAR-205 data is greater than the mass release from the SWESSAR design. This is mostly due to the addition of 40,000 11rn from the main feedwater line in the b-SAR-205 analysis as opposed to 10,000 lbm which would actually De stored in the design. The results of main feedwater line in the SWESSAR/B-SAR-205 the recalculated mass and energy release rates for the 2.0 ft2 break (the temperature limiting case) are listed in Table 6.2.1-9A. Since the auxiliary f eedwater system runout flow to the affected steam generator is greater for SWESSAR-P1 than that assumed in B-SAR-205 (2,850 gpm vs 1,840 gpn), the small break area to be analyzed is recalculated for the SWESSAR-P1 design. As in the case for the B-SAR-205 0.22 ft2 breax, the SWESSAR-P1 0.33 ft2 39 break flux at is based on mass flux out the break equal to the Moody mass 630 psia. 'Ih e mass and energy release rates are calculated with B-SAR-205 methodology until the mass rate out the break equals the auxiliary feedwater flow rate to the af fected steam generator. From this point in time to the time auxiliary feedwater flow is isolated by the FOGG logic, enough energy is removed from the NSSS and other steam generator so that the NSSS temperature is uniform at the saturation value for 600 psia
,e q r
6.2-20H Amendment 39 7/14/78
SWESSAR-P1 (4 8boF) . After the flow into and out of the affected steam generator becomes equal, the following energies are renoved from the NSSS before the auxiliary feedwater flow to the affected steam generator is isolated. Source Energy, 106 Btu 39 Reactor coolant 17.22 Fuel 1.36 Primary metal 11.35 Steam generator fluid 3.86 Decay heat 38.25 RC pung heat 6.92 Main Feedwater Pipe Break Accident T'ne flow rate out of the steam generator side of tne broken reedwater pipe Ls assumed to be frictionless Moody flow at the initial secondary pressure. The cecondary system pressure is assumed to be constant until the steam generator secondary water inventory is exhausted. The flow out ot the broken feedwater pipe volume between the nreak and the f eedwater control valve is also assumed to De predicted by the frictionless Moody Ilow m0<101. The initial maximum flow is assumed until the inventory in the feedwater pipe is exhausted. Since the marimum auxiliary feedwater temperature is 120 F, this flow is not considered in the reedwater pipe break analysis. The mass and energy release rates and the integrated releases for the feedwater pipe break analysis are given in Table 6.2.1-10. 6.2.1.5 Minimum Containment Pressure 1ar ECCS Analysis The containment backpressure used in the ECCS analysis is included in the NSSS Vendor's ECCS evaluation model as discussed in Section 6.3. o . 2.1.^ Interface Requirements The Utility-Applicant shall provide the following in his application for a Construction Permit:
- 1) A verification as discussed in Sect 1on 6.2.1.1.2 that the containment pressure analysis is applicable (RESAR-41 only) .
- 2) A verification as discussed in Section 6.2.1.1.2 that the main steam line break analysis is applicable
{RESAR--41 only) . References for Section 6.2.1
- 1. "LOCTIC -
A Computer Code to Determine the Pressure and Temperature Response of Dry Containments to a Loss-of-Coolant Accident," SWND-1, Stone & Webster Engineering Corp., 6.2-20I Amendment 39 h!)j []} 7/14/78
SWESSAR-P1 September 1971, Letter from W.J.L. Kennedy to P.A. Morris, et al.
- 2. " CONTEMPT - A Computer Program for Predicting the Containment Pressure-Temperature Response to a Loss-of-Coolant Accident,"
IDO-17220, June 1967.
- 3. Safety Evaluation by the Division of Redctor Licensing, USAEC, in the Matter of Virginia Electric and Power Company, Surry Power Station Units 1 and 2, FeDruary 23, 1972, Docket Numbers 50-280 and 50-281.
- 4. Safety Evaluation by the Division of Reactor Licensing, U S A 1.C , in the Matter ot Maine Yankee Atomic Power Company, Maine Yankee Atomic Power Station, February 25, 1972, Docket Numner 50-309.
- 5. Safety Evaluation by the Division of Reactor Licensing, USAEC, in the Matter of Virginia Electric and Power Company, Nortn Anna Power Stations Units 1 and 2, October 14, 1970, Docket Numbers 50-338 and 50-339.
- 6. Safety Evaluation ny the Directorate of Licensing, USAEC, in the Matter of Virginia Electric and Power Company, North Anna Power Station Units 3 and 4, Decemoer 29, 1972, Docket Numbers 50--404 and 50-405.
- 7. Slaughterbeck, D. C., "A Review of Heat Transfer Coerficients for Condensing Steam in a Containment Building Following a Loss-of-Coolant Accident," Interim Task Report, Subtask 4.2.2.1, Idaho Nuclear Corp., January 1970.
- 8. Idel'Chik, l.E., Handbook of Hydraulic Resistance, AEC-TR-6630, 1960.
- 9. " Protective Coatings for Light Water Nuclear Reactor Containment Facilities," ANSI N101.2-1972, Section 4,
" Procedures for Testing Coatings (Paints) at Simulated DBA Conditions."
- 10. denry, R.6., Grolmes, M.A., and Fausr.e, H.K., " Propagation Velocity et Pressure Waves in Gas-Liquid Mixtures," Presented at the International Symposium on Research in Concurrent -
Liquid Flow, September 18-19, Waterloo, Ontario, Canada.
- 11. Moody, F.J., "A Pressure Pulse Model for Two Phase Critical Flow and Sonic Velocity," Journal of Heat Transfer, Trans.
ASME, Series C, Vol. 91, No. 3, August 1969.
- 12. Moody, F.J., " Maximum Flow Rate of a Single Component, Two-phase Mixture," APE.D-437 8, General Electric Company, October 25, 1963. ., , ,
1 [V.) s 6.2-20J Amendment 39 7/14/78
SWESSAR-P1
- 13. McAdams, W. H., Heat Transmission, Third Edition, 1954, page 44.
- 14. Trinkle, J., " Experimental Verification ot the Analytical Procedure Used to Determine the Haat Removal Capability of the Containment Cooling Units," CYAP-104, 1967.
- 15. Quandt, E.. " Analysis of Gas-Liquid Flow Patterns," A.1.Ch.E.
Preprinti No. 47, presented at Sixth National Heat Transfer Conference, Boston, Mass., August 11-14, 1963.
- 16. bilanin, W.J., "The General Electric Mark III Pressure Suppression Containment System Analytical Model," NEDO-20533, General Electric Company, June 1974.
- 17. Supplement No. 1 to the Satety Evaluation Report ny the Di-rectorate of Licensing, USAEC, in the matter of the Millstone Point Company, et al, Millstone Nuclear Power Station Unit 3, Docket No. 50-423, May 16, 1974.
- 18. Safety Evaluation Report by the Directorate of Licensing, USAEC, in tne Matter of Duquesne Light Company, Toledo Edison Company, Pennsylvania Power Company, Beaver Valley Power Station Unit 1, Docket 50-334, Octoner 11, 1974.
- 19. Supplement No. 2 to the Safety Evaluation by the Directorate or Licensing, USAEC, in the Matter of Cleveland Electric Illuminating Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania Power Company, and Toledo Edison Company, Beaver Valley Power Station Unit 2, Docket nu. 50-412, March 20, 1974.
- 20. Safety Evaluation by the Directorate of Licensing, USAEC, in the Matter of Virginia Electric Powu. Company, Surry Power Station Units 3 and 4, Docket No. 50-434/50-435, May 23, 1974.
- 21. barzelay, M.E. and Holloway, G.F., Effect of an Interface on Transient Temperature Distribution Composite Aircratt Joints, National Advisory Committee for Aeronautics, National Advisory enmmittee for Aeronautics, Technical Note 3824, Washington, D.C., April 1957.
- 22. Barzelay, M.E. and Holl oway, G.F., Interface Thermal Conductar3ce of Twenty-Seven Riveted Aircraft Joints, National Advisory Ccumnittee for aeronautics, Technical Note 3991, Washingto : D.C., July 1957.
- 23. Barzelay, M.E., Range or Interface Thermal, Conductance for Aircraf t Joints, National Aeronautics and Space Administration, Tecnnical Note D-426, May 1960.
n 1 m" 6.2-20K UUJ tsJ Amendment 39 7/14/78
SWESSAR-P 1
- 24. Rohsenow and Choi, Heat, Mass, and Momentum Transter, Prentice-Hall, Englewood Clif f s, N.J. 1961
- 25. Uchida, H., Oyama, A., and Togo, Y., " Evaluation of Post-Incident Cooling Systems of Light-Water Power Reactors,"
Proceedings of the Third International Conference on the Peacetul Uses of_ Atomic Energy Held in Geneva, 31 August-9 September 1964, Vol. 13, New York: United Nations, 93-104 (A/ CONF 28/P/43 f.)
- 26. Shepard, R.M., et aA, "Westingnouse Mass and Energy Release Data for Containment Design," WCAP-8312-A, Rev 1, Westinghouse Electric Corporation, June 1975.
- 27. Schmitt, R.C., et al, " Simulated Design Basis Accident Test of the Carolinas Virginia Tube Reactor Containment - Final Report," UC-80, Idaho Nuclear Corporation (Dec. 1970).
- 28. Norberg, J.A., et al, " Simulated Design Basis Accident Tests of the Carolinas Virginia Tube Reactor Containment -
Preliminary Results," IN-13 24, Idaho Nuclear Corporation (Oct . 1969).
- 29. Hargrove, H.G., " MARVEL - A Digital Computer Code for Transient Analysis of a dult11oop PWR System," WCAP 7909, Westingbouse Electric Corporation, October 1972.
6.2.2 Containment Heat Removal Systems The containment heat removal systems consist of:
- 1. The containment spray system
- 2. The containment atmosphere recirculation system (Section 9.4.5.1)
- 3. The low head safety injection / residual heat removal system (NSSS Vendor's scope)
The containment neat removal systems return the containment pressure to a low value following a break in either the primary or secondary system piping inside the containment. Heat is transrerred f rom the containment atmosphere to the spray water and the reactor plant component cooling water by the containment spray system and tne containment atmosphere recirculation system.
-c
/,
JuJ f_ a v 9 6.2-2 0L Amendment 39 7/14/78
SWESSAR-P1 respectively (refer to Fig. 6.2. 2-1 and 9.4. 5.1-1) . In addition, heat is transferred from the water which has fallen to the containment floor to the reactor plant component coolina water system in the residual heat removal (RHk) heat exchangers (NSSS Vendor's scope) af ter the refueling water storage tank (RWST) has emptied (recirculation phase) . The containment spray system is shown in Fig. 6.2.2-1 and the system component da ta in Table 6.2.2-1. The iodine reraval capability of the containment spray system is discussed in Section 6.2.3.2. 6.2.2.1 Desian Bases The design bases for the containment heat removal systems are:
- 1. The containment peak pressure, following the DBA, shall be less than the containment design oressure by an adequate margin assuming the worst sinale active failure. This basis specifically applies to the containment spray system.
- 2. The containment atmosphere pressure 24 hr after the DBA shall be at an acceptably low value assuming the worst active failure. This basis applies to all containment heat removal systems.
- 3. The systems shall be capable of maintaining u low containment pressure indefinitely following the DBA.
This basis applies to all containment heat renoval systems.
- 4. The containment spray headers shall be capable of delivering spray water to the containment atmosphere in sufficient quantity and with an averace droplet diameter to ensure adequate heat removal to accomplish bases 1, 2, and 3 above.
- 5. The containment heat removal systems are designated Saf ety Class 2 and Seismic Category I and are designed in accordance with ASME III, Class 2. This basis applies to all containment heat removal systemn.
"Mininum ESE" and " Normal ESF" as used in this section are defined as follows:
" Minimum ESF" are those enoineered safety features that function following a loss of offsite power and railure ot one emergency diesel generator (reier to Table 6.1-7).
" Normal ESF" are those that function followinc loss of 9 of f site power with all emergency diesel generators operatina (refer to Table 6.1-7) .
6.2-21 [fT ar,
'O/ Amendment 6 3/28/75
SUESSAR-P1
'Ihe calculated containment peak pressures following the DBA are given in Table 6.1-3. The single active failure which gives the highest calculated peak pressure is also listed in this table. A single active failure of one diesel generator (minimum ES F) is assumed in the calculation for maximum containment pressure after a 24 hr period.
The sources and quantities of energy that must be removed from the containment to meet the design ba ses are given in Table 6.2.1-2. The ESF systems are assumed to become effective as specified in Ta ble 6.2.1-1. 6.2.2.2 System Design The containment spray system consists of parallel and separate subsystems. Each subsystem consists of one spray pump, two spray headers, spray nozzles as specified in Table 6.2.2-1, one ESF sump, and associated piping and valves. Initially, each contain: cent spray pump draws water f rom the RWST (in j ection phase). The quantity of water in the RWST is sufficient to provide water for both the containment spray system and emergency core cooling system (ECCS) . After the RWST is exha usted , suction is taken from an ESF sump (recirculation phase) . The RNST is an atmospheric tank that is located inside the annulus building. Sufficient venting is provided to preclude tank f ailure during drawdown conditions. The spray water which enters the refue_ing cavity drains to the reactor cavity since the refueling cavity water seal is in the raised position. From the reactor cavity, this water flows to the iloor of the containment through the cenetrations in the primary shield wall provided for the reactor coolant piping. Wa ter which enters the in-core instrumentation tunnel flows along 33 the containment floor exiting to the outer portion of the containment and the ESF sumps. The lx) rated water in the RWST has a maximum temperature of 120 F. Sodium hydroxide (NaOH) solutien from the refueling water chemical addition tank (CisT) is added to the containment spray to improve its iodine removal capability. Addition of th e NaOM occurs when the system is energized by the CDA signal. The resulting pH of the containment sprays is evaluated in Secti on 6.2.3.2. Westinqhouse-41 Tae chemical addition system uses gravity flow to add NaOH into the containment spray system piping. The caustic (NaOH) from the 6.2-22 Eaendment 33 6/30/77 s.o bus7 m au
SWZSSAR-P1 chemical addition tank (CAT) is introduced by means of Y connections into the containment spray pump suction pipes from the RWST as shown in Fig. 6.2.2-1. Upon receiving a CDA signal, redundant valves between the CAT and the Y connections open to allow immediate flow into the spray pump lines. The valves are closed during normal operation to prevent mixing of the NaO!I with ooric acid in the RWST. The fluid level of the CAT is at the 9 Amendment 33 6.2-22A,his b []) 6/30/77
SWESSAR-P1 maximum fluid level of the RWST. Since the density 7f the NaOH solution is at least 10 percent greater thar. the RWST .luid, the minimum initial driving force of the NaOF is at least 10 percent of the CAT height. The pipes between the CAT and the Y connections have a negligible pressure drop to ensure that there will be no significant resistance to flow f rom the CAT. both the RWST and the CAT are adequately vented to permit rapid drawdown; therefore, the tanks are in hydrostatic balance after the valves are opened. As water is pumped from the RWST, NaOH flows from the CAT. The liquid level of the CAT falls at a constant fraction of the rate of level decrease of the RWST. This ensures proper metering of the CAT liquid into the RWST liquid. Mixing takes place .in the fully developed turbulent flow of the spray piping before reaching the spray nozzles. Combustion-Engineering, Westinghouse-3S , Babcock & Wilcox The chemical addition system uses an eductor to add caustic plaCHI) into the containment spray piping. The NaOH from the chemical addition tank (CAT) is introduced into the containment spray pump suction pipes as shown in Fig. 6.2.2-1. Upon receiving a CDA signal, redundant valves between the CAT and the eductors open to allow immediate flow into the spray pump lines. The valves are closed during normal operation to prevent mixing of the NaOH with the boric acid in the contaimment spray piping. The eductor draws the NaOH at a constant rate into the containment spray pump suction. This ensures proper metering of the NaOH solution into the RWST fluid. Mixing takes place in the fully developed turbulent flow of the spray pump and piping before reaching the spray nozzles. The eductor continues to add NaOH from the CAT during the re .rculation phase with the sump water as required to ensure tant there is enough NaOH in the sump solution to attain a pH of 25 8.5. Spray and pump pH as a function of recirculation time is shown in Fig. 6.2.2-13. 9 , , o - OuJ cVJ 6.2-23 Amendment 25 4/30/76
SWESSAR-P1 24 All NSS Vendors The containment spray pump suctions are automatically switched to the ESF sumps when the RWST water level reaches the low-low-1 water level. At this time the motor-operated valves in the suction lines from the sumps are automatically opened, and after a short time delay, the motor-operated valves from the RWST are automatically closed, thus isolating the RWST and CAT from the containment spray pumps. In addition, the check valves in the lines from the RWST will close due to increased containment pressure. This assures water flow to the spray pumps at all times. No operator action is required to accomplish the switchover to the recirculation rP Tse. Failure to switch a spray train over automatically is de 2 ced by the low-low-2 RWST water level which is annunciated in _nc control room. At this time the operator switches the af f ected containment spray train to the recirculation phase manually from the control room. The ES F sumps are enclosed by a protective screen assembly. In addition, the check valves in the lines from the RWST w.? close due to increased containment pressure. Three stages of trash filtration are provided: trash bars, a coarse mesh screen, and a fine mesh screen as shown in Fig. 6.2.2-9. The screen sizes are given in Table 6.2.2-1. This ESF sump design camplies with Regulatory Guide 1.82 (Section 3A.1-1. 82) . The ESF sumps are physically separated as shown in Fig. 1.2-3. The fine mesh over the bottom 1 in. grating has a mesh spacing that is smaller the nozzles. than the largest sphere that can pass through all This prevents any particulate matter which could plug a spray nozzle from entering the sumps during the recirculation phas e . The mesh spacings of the sump screens are civen in Table 6.2.2-1. Each spray pump is connected to two spray headers as shown in Fig. 6.2.2-1. Spray is distributed to the containment atmosphere at two dif ferent elevations in the containment dome. Sufficient spray coverage is provided by the containment spray system even if one containment spray subsystem fails to operate. Spray coverage is evaluated in Section 6.2.3.2. Each of the headers is provided with nozzles a- speciried in Table 6.2.2-1. The nozzle orientations a r c- given in Fig. 6.2.2-10. Each containment spray discharge line which penetrates the containment contains a check valve inside and a motor operated isolation valve outside the containment. One quarter inch diameter drain lines, downstream from the check valves inside the containment, drain the containment spray risers should any water enter the risers during periodic testing. Flow through the drain lines during system operation does not significantly decrease the capacity of the containment spray system. O( ,] 6.2-24 Amendment 24 4/23/76
SWESSAR-P1 Each of the containment spray suction lines f rom the ESF sumps is provided with a motor-operated isolation valve and a check valve. The check valve precludes inadvertent dumping of the RWST contents directly into the containment when the recirculation phase is initiated. The lines from the ESF sump to the outside containment isolation valves are designed as described in Section 3.6.2.2.1 Paragrep B. The portion of each suction line embedded in the concrete mat between the ESF sumps and the safety features areas is wrapped with compressible packing to allow for seismic and thermal effects. All containment spray pumps and motors are located outside the containment in the saf ety f eature areas of the annulus building at an elevation sufficiently below the ESF sumps to ensure adequate net positive suction head (NPSH) . The pumps are of the vertical deep well type, each provided with a stainless steel well casing supported by a brackat connected to the concrete annulus build ing mat. The pump motors are accessible for inspection and aaintenance. The containment spray pumps, and the associated suction line valves and motors outside the containment, are designed und installed to account for differential movement between these components and the containment that may result from a seismic event. Restraints and sliding supports are used as appropriate. The containment heat removal systems are sufficiently redundant to perform their design f unction properly assuming a single @ i - m-u vJ Lgg
- 6. 2-24 A Amendment 24 4/23/76
SWESSAR-P1 active failure in the short term or an active or passive failure in the long term (af ter 24 hr) . The only interconnecting piping between the containment spray subsystems is the common test return pipe to the RWST. This affords a high overall system reliability sin. e a failure in one subystem cannot af fect the capability of the other subsystem (s) to perform the desired safety function. All test piping is normally closed off during plant operation by valves which are und2r administrative control. 6.2.2.3 Design Evaluation Heat Removal Systems Performance Analyses of the effects of the containment heat removal systems on containment pressure following various postulated accidents are done with the LOCTIC computer program (described in Section 6.2.1) . Calculations indicate that, while falling through the containment atmosphere, the smaller spray droplets approach 100 percent of thermal equilibrium with the containment atmosphere, whereas the larger droplets approach 99 percent equilibrium.(6) It is conservatively assumed in all containment design analyses that only 90 percent of the spray reaches thermal equilibrium with the containment a tmosphere (i.e., the spray efficiency is 90 percent) . These calculations also indicate that an average droplet diameter of 1,000 microns is adequate for heat removal. The actual droplet size distribution of the spray is shown on Fig. 6.2.2-12. Heat is also removed from the containment atmosphere by the containment a tmosphere recirculation system and transferred directly to the reactor plant component cooling water systen.. The heat removal rate vs time curve for the containment atmosphere recirculation coolers is shown in Fig. 6.2.2-3. Minimum ESF (see Table 6.1-7) is assumed in this calculation. The absorption heat rate vs time curve for the passive heat sinks is shown in Fig. 6.2.2-4. After the usable water inventory in the RWST is exhausted, the cont tinment spray, high head safety injection (HHSI) , and the low head safety injection (LHSI) systems are switched to the recirculation phase of operation. During this phase, the pumps take suction from the ESF sumps and heet is removed from the water on the containment floor by the RHR heat exchangers. There is no water held up in the refueling cavity during the recirculation phase. Water in the reactor cavity flows to the 33 or7 6.2-25 (103 c i Amendment 33 6/30/77
SWESSAR-P1 floor of the containment through the penetrations in the primary g shield wall provided for the reactor coolant piping. Water in W the upender area and storage area for lower internals is drained 33 to the containment floor through the fuel pool purification system drains (Fig. 9.1. 3-2 ) , which are open during plant operation. By separating these drains and providing drain guards with cros s-secuonal areas much greater than the area of the drain pipe, drainage of the refueling cavity is independent of break location. The heat rate vs time curve for the RHR heat exchangers is shown in Fig. 6.2.2-5. Minimum ESF is assumed for this calculation. The component cooling water temperature is shown in Fig. 6.2.2-11 as a function of time after the accident. NPSH Available to ESG Pumps Injection Phase Sufficient NPSH is available to the safety injection pumps and containment spray pumps during the injection phase. The following equation is used to calculate the available NPSH. Available NPSH = A + B - C - D where A = the pressure in the RWST (atmospheric pressure) B = the elevation head from thd discharge point on the RWST to the first stage impeller on the pump C = the vapor pressure at the maximum expected temperature (120 F) in the RWST D, = the suction piping friction loss assuming pump operation at the maximum flow rate Adequate NPSH is provided during the injection phase as shown in Table 6.2.2-3 which lists the required NPSH for each pump and the mini:mun NPSH available. Recirculation Phase For the recirculation phase the vapor pressure of the water in the sump is assumed to be equal to the containment pressure. Additional margin in available NPSH is ensured by: (1) hv credit is taken for the water level inside the containment.
'j
) .)
6.2-26 Amendment 33 6/30/77
Sh2SSAR-P1 (2) The vapor pressure of the sump water is actually less than Me containment pressure. (3) The friction losses at actual pump flow rates are less than at maximum flow rates. Thus, adequate NPSH is available for all ESF pumps during the recirculation phase of operation. The required NPSH for each pump and the minimum NPSH available are listed in Table 6.2.2-3. 9 6.2- 26A Amendment 33 6/30/77 (, 6 s,.
<ua
SWESSAR-P1 Contain men t Pressure Response The containment pressure transients for a DER of the hot leg. pump discharge line, and pump suction line are compared in Fig. 6.2.2-7. The pressure transients for a spectrum of breaks at the " worst break location" are shown in Fig. 6.2.2-8. The containment pressure one day after the DBA is calculated using the assumption of minimum ESF. The containment pressure transients for this calculation are shown in Fig. 6.2.2-6 for minimum ESF and normal ESF. The effect of secondary side steam generator energy addition to the containment atmosphere is included in the analyses. Iodine Removal By Containment Spray System The spray nozzles are selected to provide adequate iodine remo7al capability and containment ccverage. The drop size dictribution, spray coverage, and resulting iodine removal coefficients arc evaluated in Section 6.2.3.2. Leakage During the Circulation Phace Leaks in the suction and diccharge piping of the pumps (i.e., LHSI, HHSI, and containment spray system) are controlled as follows:
- 1. Large leaks in the discharge piping within the engineered safety features areas are detected by the control room operator. A decreased pressure reading indicates a pipe break; the operator then stops and remote manually isolates the leaking pipe from the containment and shuts down the pump in the affected loop.
- 2. Smaller leaks within the engineered safety feature areas are detected by an increased liquid level in the affected area sump. The sump .evel indicator will be saf ety-related with an alarm in the control room, and will be capable of detecting a 1 gpm leak within 1 hr.
A 50 gpm leak will be detected within a few minutes. 33 The operator could then shut down the affected ECCS train within 5 more minutes. Assuming an operator delay time of 30 min, the resulting water flooding is well below the amount that could interfere with the redundant ECCS train (Section 3. 8. 4.1) . The leakage during recirculation is based on the leakage of the ECCS system provided by the NSSS Vendor and the containment spray 6.2-27 Amendment 33 6/30/77
SWESSAR-P1 system provided by Stone & Webster. The assumptions used by the NSSS Vendor and Stone & Webster are given below: Babcock 6 Wilcox - ECCS Section 15.1. 3.2.4 of B-SAR-205 gives conservative leakage rates from equipment recirculating sump solution outside the containment structure. Combustion-Engineering - ECCS Section 6.3.1.5.P.1 of CESSAR specifies the maximum leakage rates listed below for C-E supplied equipment. High and Low Pressure Safety Injection Pump Seal 100 cc/hr/ pump Valves - backseat leakage 10 cc/hr/in. of seat diameter
- across the valve seat 10 cc/hr/in. of nominal valve size Westinghouse - ECCS Table 6.3-6 of. RESAR-41 and Table 6.3-3 of RESAR-3S list the maximum potential recirculation loop leakage external to the containment structure.
Stone 6 Webster The following maximum leak rates are part of the Design Specification for the valves in the containment spray system. Valves - backseat leakage - 10 cc/hr/in. of (stem) seat diameter Consistent with letters from the ACRSC1)(2) concerning vital piping which must f unction following a LOCA, failure of the suction piping is not considered credible during the short term following a LOCA. However, should a failure occur in a suction line in the long term, the water level sensors in the bottom of the engineered safety features area alert the operator to remote manually close the motor opera ted valves in the suction and discharge lines of the affected pump and to shut down the pump in tha t a rea. Water level indication is discussed in Section 9.3.3. ( eG
~'
m.
.u,
?
6.2-28 Amendment 33 6/30/77
SWESSAR-P1 padiological Consequences of Recirculation Liquid Leakage The radioactive releases from postulated leaks from equipment recirculating ESF surp water after a LOCA in the annulus building result in:
- 1. 0-2 hr exclusion area boundary doses of less than 8 percent of the thyroid dose and less than 2 percent of l0 the whole body dose calculated for any given distance in Section 15.1. 13.
- 2. 0-30 day low population zone dose of less than 16 per- g cent of the thyroid dose and less than 13 percent of the whole body dose calculated for any given distance in Section 15.1.13.
These results are based on the assumptions given in Table 6.2.2-4 Containment Atmosphere Recirculation System Dif ferential Pressure Buildup The containment atmosphere recirculation coolers normally draw suction directly from the containment atmosphere through three of the four vertical sides. These sides of the cooler housing are open to the containment atmosphere. This design precludes any differential pressure buildup acrocs the cooler housing during a LOCA. Since the ventilation ducts inside the containment are not used during a LOCA, they are not analyzed for differential pressure buildup . Failure Analysis A failure analysis for the components of the containment heat removal systems is given in Table 6.2.2-2. 6.2.2.4 Testing and Inspections The tests performed on the containment spray system are described below. The tes ts performed on the containment atmosphere recirculation sy stem are discussed in Section 9.4.5.1.4. The frequency of each test is given in Section 16.4.5 of the Technical Specifications. Containment Spray Pumps Preoperational testing of the containment spray pumps is performed according to the following outline:
- 1. Pipe plugs are inserted into each of the spray nozzle sockets in the spray headers.
//, o o 6.2-29 0 b ,) cQj Amendment 17 9/30/75
SWESSAR-P1
- 2. The blind plate is removed from each + tt flange of the system under test and a flow orifice is inserted.
- 3. The internals are removed from the containment isolation check valve of the containment spray subsystem not under test. The valve in the test line (to the RWST) of this subsystem is opened.
- 4. The pump is started.
- 5. Reference valves of the test quantities required for the 8
inservice tests specified in Section 16.4.5 are obtained. Fluid flow is from the RWST to the spray headers and back to the RWST through the spray headers, risers, and test line of the subsystem nct under test. Flow through each header can be measured by connecting a flow indicator to the taps on each side of the flow orifice. The pump developed head (discharge pressure minm suction pressure) and the measured flow should correspond to the full flow point on the pump head-flow curve. Subsequent to preoperational testing, each containment spray pump is periodically flow tested by recirculating water back to the RWST through the test line connecting the pump discharge line to the RWST. The strainer in the common test line at the RWST has a mesh spacing whicu is smaller than the largest sphere that can pass through all of the spray nozzles. A decreasing flow through the test line indicates strainer clogging. Absence of such a decrease indicates that there is no particulate matter capable of plugging the nozzles on the containment spray header. Acceptance criteria for the periodic flow test are specified in 8 Section 16.4.4. o ,~ (bD uG] 6.2-30 Amendment 8 3/28/75
SWESSAR-P1 Valves All safety class valves require periodic testing as specified in Section 16.4.5. Nozzles Means are provided for qualitative in place air flow tests of the containment spray nozzles. These tests are performed by connecting air compressors to test connections in the containment spray lines and by manually checking each individual nozzle for flow. Airflow tests are necessarily qualitative. The extremely large flow rates of air necessary to develop a measurable pressure drop across nozzles designed to spray water make quantitative airflow tests unfeasible. RWST The RWST is provided with a manhole for inspection access during refueling periods. The RWST water is sampled periodically to monitor water chemistry. Provisions are made so that the water can be purified, if necessary, by circulating it through che purification portion of the fuel pool cooling and purification system (Section 9.1.3) . The containment spray subsystems are completely independent and sufficient redundtncy is provided to permit testing and/or repairing of one subsystem without necessitating a shutdown of the reactor. i,y n n 6.2-31 bbj Lt U Amendment 22 3/17/76
SWESSAR-P1 Chemical Addition Westinghouse-41 Two preoperational tests will be performed with the CAT filled with water to ensure the proper pH of the containment spray solution during both the injection and recirculation phases. The first the to test will minimize the drawdown rate of the CAT in relation RWST by running all ECCS pumps and all containment spray pumps. This test will demonstrate that the level in the RWST does not drop faster than the level in the CAT. The second test will maximize the drawdown rate of the CAT in relation to the RWST by running only the containment spray pumps. This test will demonstrate that the CAT level does not drop faster than the RWST level. Testing or these two cases will verify that the tanks are in hydrostatic balance and that during a LOCA sufficient sodium hydroxide will be injected to meet the pH requirements. These tests will confirm that the level changes in the RWST and CAT are in the same relationship assumed during design. The di7 meters of these two tanks are measured for the as-built conditions to provide an evaluation of volumetric change in each tank with level. The above information is used to confirm that the flow rates discussed in Section 6.2.3.2.2 are met. If some devia tion exists in the as-installed system, the NaOH concentration in the CAT can be modified as described in Section 6.2.3.2.2 to provide the required pH. Combustion-Engineering, West inghous e-3S , Bab cockSWil cox A preoperational test with the CAT filled with water will be performed to ensure the proper operation of the eductor. This test will confirm the flow rates for the containment spray pumps and the rate of NaOH addition. The results of this preoperational test with water will be compared with data f rom manufacturers' tests correlating the eductor flow rates for water with those of NaOH. If some deviation exists in the as-installed condition , the NaOH concentration can be varied 15 percent to provide the required pH. All NSSS Vendors To maintain an adequate supply at the required concentration of NaOH, dests are made periodically to measure both the fluid level and the NaOH concentration of the CAT. The fluid level of the CAT is usually checked once a month. This is in addition to a redundant low level alarm and the
- aonitoring of fluid level in the control room. The NaOH concentration is measured once a year by taking a sample from the tank through a sampling connection. Experience indicates that this frequency of testing is sufficient.
"h e CAT contains a caustic (NaOH) solution with a concentration between 9 and 25 percent by weight with an ambient temperature (/ t'- J
'"1 m . I 6.2-32 Amendment 25 4/30/76
SWESSAR-P1 below 125 F. The corrosion characteristics of stainless steel with such NaOH solutions are such that the design characteristics of the CAT will not be degraded over the 40 year life of the plant. Reaction of NaOH with the stainless steel forms FeO-H20, which is insoluble in un alkaline solution. This reaction does not reduce the alkalinity of the solution and, therefore, does not impair the offectiveness of the NaOH solution. The cover gas for the tank is air which contains nominally .033 percent by volume CO2. Carbon dioxide reacts with the NaOH solution to form NaHCO3. This reaction could reduce the basicity of the solution and, potentially, reduce the effectiveness of the NaOH solution. Operating experience at the Surry Units 1 and 2 power plants has indicated that the above chemical reactions do not pose a problem. Monitoring of the Surry CAT has not shown any traces of FeO-H2O or NaHCO3. Also, the concentration of NaOH in the Surry CAT has remained at a constant level without deviation after years of operation. 6.2.2.5 Instrumentation Requirements The containment spray system is initiated by a CDA signal. Unless manually stopped, the system operates indetinitely. Motor operated valves on the discharge lines of the containment spray pumps and on the discharge line of the CAT open on a CDA signal. The following instrumentation is provided to allow monitoring of system performance during operation and testing. For a detailed description of the instrumentation, see Section 7.3.
- 1. Redundant level instruments, with control room indication (one channel recorded) , are provided for the 6.2-32A r7 o--) Amendment 25
- 30) L'C 4/30/76
SWESSAR-P1 RWST. Hig h , low, low-low-1, and low-lcw-2 RWET water level alarms are provided.
- 2. Tempera ture instrumentation is provided for the RWST with readout in the control room.
- 3. Redundant level instruments are provided in the refueling water CAT with a low level alarm and indication in the control room.
- 4. Temperature instrumentation is provided in the CAT with readout in the ccntrol room.
- 5. Flow indication is provided in the control room to monitor flow from ach containment spray pump.
- 6. Pressure indication is provided in the control room to measure the discharge pressure of each containment spray pump.
- 7. Additional flow indication in the test lines and local pressure gages on the suction and discharge piping of the containment spray pumps are used for pump testing.
- 8. Level indicators are provided in each ESF sump (one channel recorded).
- 9. Temperature indicators are provided in each ESF sump (one channel recorded) .
6.2.2.6 Materials The portions of the containment heat removal systems which are exposed to either the pumped fluids or to the atmosphere are made of stainless steel. Consequently,containment there are no radiolytic or pyrolytic decomposition products which can interfere with the safe operation of the ESF. Extensive experimental studias(3, *) have been made to determine the corrosion rates and the effect on the materials in the containment from the use of base borate spray solutions. It was found that the amount of aluminum in the containment should be minimized and, where used, should be restricted to equipment not required to function af ter the accident. Copper is relatively unaffected by base borate spray solutions. Concrete wetted by the spray solution does not have its strength impaired. This has been substantiated by several experiments. Further protection is afforded by the use of protective coatings within the containment. A la rge number of tests have demonstrated that these coatings can withstand the temperature and chemical action of the sprays. 9 g7 o-
" E' 6.2-33 Amendment 8 3/28/75
SWESSAR-P1 The possibility of stress corrosion cracking of the stainless steel piping was also investigated (3)(*). It was found that the higher pH borated solutions cause little or no stress corrosion cracking. However, tests with a low pH (approximately 5.0) borated water from the RWST resulted in numerous cases of cracking. 6.2.2.7 Interf ace ROquirements Sufficient water is available in the RWST to supply the safety injection systems and the containment spray system at the boron concentration requirements of the NSSS Vendor. This design meets all applicable interf ace requirements of SWESSAR-P1 Table 6.3-3.
- 1. The Utility - Applicant shall specify that the eductor manuf acturer perf orm flow tests with water and water -
sodium hydroxide, as described in Section 6.2.2.4.
- 2. The U ility - Applicant shall perform a preoperational test on the eductor system as described in Sections 6.2.2.4 and 14.0.
References for Section 6.2.2
- 1. Letter from Stephen H. Hanauer, Chairman, Advisory Committee on Rea ctor Safeguards, to Honorable Glenn T. Seaborg, Chairman, USAEC, dated May 15, 1969, subject: " Report on Edwin I. Hatch Nuclear Plant."
- 2. Letter from Stephen H. Hanauer, Chuirman, Advisory Committee on Reactor Safeguards, to Honorable Glenn T. Seaborg, Chairman, USAEC, dated May 15, 1969, subject: " Report on Brunswick Steam Electric Plant."
- 3. Griess, J. C. and Baccarella, A. L., " Design Considerations of Reactor Containment Spray Systems, Part III. The Corrosion of Materials in Spray Solutions," USAEC Report ORNL-TM-2412, Part III, Oak Ridge National Laboratory, December 1969.
- 4. Griess, J. C. and Creek, G. E., " Design Considerations of Reactor Containment Spray Systems, Part X , Corrosion Tests, with Low pH Spray Solution," USAEC Report, ORNL--TM-2 412 ,
Part X, Oak Ridge National Laboratory.
- 5. Safety Evaluation by the Division of Reactor Licensing, United States Atomic Energy Commission in the Matter of Commonwealth Edison Co., Zion Station, Units 1 and 2, Docket No. 50-295/304, ..ugust 21, 1968.
[' D'?J n ' y.. O 6.2-34 Amendment 25 4/30/76
SWESSAR-P1
- 6. Parsly, L. F., "The Heating of Spray Drops in Air-Steam Atmospheres," Design Consideration of Reactor Containment Spray Systems, Part VI, ORNL-TM-2412, Part VI, January 1970.
6.2.3 Containment Air Purification and Cleanup Systems The engineered safety features systems that fall into the category of containment air purification and cleanup systems are: 6.2-34A Amendment 25 bb) [' ," { 4/30/76
SWESSAR-P1
- 1. The supplementary leak collection and release system, described in Section 6.2.3.1.
- 2. The containment spray system described in Section 6.2.2 removes radioactive iodine from the containment atmosphere following a major LOCA, as described in Section 6.2.3.2.
- 3. The containment combustible gas control system, described in Section 6.2.5.
Nonsafety related systems that are classified as containnient air purification and cleanup systems are:
- 1. The containment purge air system described in Section 9.4.5.2.
- 2. The con tainment atmosphere filtration system described in Section 9.4.5.3.
6.2.3.1 Supplementary Leak Collection and Release System The function of the supplementary leak collection and release system (SLCRS) is to minimize the release of airborne radioactive contaminants from the containment structure following a loss-of-coolant accident (LOCA) and from the fuel building following a fuel handling accident and during refueling. The system is shown in Fig. 6.2.3.1-1 and principal component design parameters are given in Table 6.2.3.1-1. The system design and operation with optional enclosure building are 18 discussed in Sections A6.2.3.1 and B6.2.3.1. 6.2.3.1.1 Design Bases The design bases of the SLCRS are:
- 1. The .sys tem shall maintain a partial vacuum of at least 0.25 in. W.G. in the fuel building, annulus building, main steam and feedwater valve areas, and in all electrical tunnels contiguous to the containment structure following a LOCA.
- 2. The system shall maintain slightly negative pressure in the fuel building during rea ctor refueling when the annulus building filter is not available (see Section 9.4.6).
- 3. The sys tem shall maintain partial vacuum of at least 0.25 in. W.G. in the fuel building following a fuel handling accident.
6.2-35 (J U J L. U Amendment 18 10/30/75
SWESSAR-P1
- 4. The sys tem shall remove from the air stream methyl iodide (CH3I) in excess of 99.5 percent and elemental iodine in excess of 99.9 percent.
- 5. The HEPA filters shall have a minimum filter efficiency of 99.97 percent to remove particulates that are 0.3 micron or larger.
- 6. The system and fans.
shall consist of two redundant filter banks
- 7. The system shall be Safety Class 3 (SC-3) and Seismic Category I.
- 8. The SLCFS shall collect 100 percent of the containment leakage into the buildings contiguous to the containment structure.
6.2.3.1.2 System Design The SLCRS is designed to collect and process all potential leakage f rom the containment following a LOCA (Section 15.1.13). As indicated in Fig. 6.2.3.1-1, the areas served by this system are as follows:
- 1. Annulus building
- 2. Fuel building
- 3. Main steam and feedwater valve areas
- 4. Electrical tunnels in the annulus building The free volume, design leakage rate, and exhaust rate for each area are listed in Table 6.2.3.1 !4.
N If Option A or B is used, the enclosure building is included in the system design (Section A6.2.3.1.2 or B6.2.3.1.2) . The system consists of two 100 percent capacity main fans, two 100 percent removal fans, capacity filter banks, two charcoal filter decay heat isolation dampers, ducts, and controls. Each filter bank includes a demister, an electric coil, a prefilter, a carbon adsorber, filters and two high ef ficiency particulate air (HEPA) (one upstream and one downstream of the caron adsorbers). The prefilters have a minimum filter efficiency of 80 percent as determined by the National Bureau of STANDARDS (NBS) dust spot method. The prefilter is of water and fire resistant design pHTA Class 1) . The carbon adsorbers are of the gasket-less no.ntray type to reduce problems encountered during adsorbent replacement. The carbon adsorbers are designed for a flow velocity of 40 fpm to give sufficient residence time (0.25 sec/2 in. bed depth). Two
,,i <> 3 Lui L /
6.2-36 Amendment 18 10/30/75
SWESSAR-P1 inches of charcoal bed is sufficient for iodine removal as discussed in Section 9.4.2.2; an additional 2 in. is provided for additional capability, which may be required for weathering and aging as well as for economic reasons. The impregnated carbon is of a type which has been demonstrated to be capable of renoving in excess of 99.5 percent of methyl iodide (CH3I) and 99.9 percent of elemental iodine under entering conditions of less than 70 percent relative humidity. The anticipated operational pressure surges will not affect the carbon adsorbers. The electrical heating coil is used to maintain 70 percent relative humidity of air entering the carbon adsorber. Each HEPA filter is capable of removing at least 99.97 percent of the 0.30 micron or larger particles which impinge on the filter. The HEPA filter is of water and fire resistant design (NFPA Class I). The ductwork is airtight, welded, and Seismic Category I design. The building (or area) isolation dampers installed in parallel and in series meet the ningle failure criterion for air flow and system isolation. The system vacuum breakers installed in parallel and in series meet the single f ailure criterion to prevent excessive vacuum buildup in the building and to maintain desired negative pressure. The small duct with two dampers at the suction side of the main fans are installed for system balancing. The charcoal filter decay heat removal fans are designed to remove heat generated by radioactive iodine adsorbed onto charcoal. A radiation monitor is located downstream of the fiAters. For details of this airborne radiation monitor, see Section 11.4.2.3.1. The door openings into the SLCRS regions are of airtight con-struction. All these doors are under administrative control for industrial security and to ensure that a minimum pressure of -0.25 in. W.G. is achieved in the event of a fuel handling accidenn or a LOCA as described in Section 6.2.3.1.5. 24 Prior to fuel handling, if required, the system is started manually and operates continuously until the refueling activity is terminated. Should the fuel building exhaust air duct monitors sense high radiation, the system will start automatically, if not already running (for detailed description, see Section 9.4.6). Following a LOCA, the SLCRS starts upon receipt of a CIA signal and within 15 seconds reaches full speed. The CIA signal is initiated within 1 second following a LOCA, thus this time is assumed negligible in comparison with the 6.2-37 Amendment 24 hhj 2[] 4/23/76
SWESSAR-P1 15 second SLCRS starting time. The starting time assumes 10 seconds f or the startup of the diesel generators and 5 seconds for the SLCRS fans to reach full speed (Section 8.3.1.1.4) . When the fans reach full speed the exhaust rate exceeds the design leakage rate, thus negative pressures are produced in each of the areas served by the SLCRS. The system is assumed to be totally effective when the negative pressure in each area is 0.25 inches W.G. The system design as modified by Options A and B is described in Section A6.2.3.1.2 and B6.2.3.1.2, respe ctively . 6.2.3.1.3 Design Evaluation, To evaluate the ef f ectiveness of the SLCRS , it is necessary to determine the portion of leakage from the containment into the annulus building, where it is collected by the SLCRS. The potential leak paths from the containment are:
- 1. Containment penetrations
- 2. Pipes penetrating the containment
- 3. Equipment hatch
- 4. Personnel hatches
- 5. Liner welds The derAgns of the equipment hatch and personnel batches are such that a pressure inside the containment assists in sealing against le akage . In addition, the equipment hatch and personnel hatches open into the annulus building where the leakage would be collected by the SLCRS. The areas serviced by the SLCRS include all penetrations and valves that undergo Type B and C tests.
24 All containment leakage may not be collected because of the following two potential bypass conditions:
- 1. Containment structure leakage through the annulus building during the time prior to depressurization to less than -0.25 in. W.G.,
- 2. Bypass through containment isolation valves of systems which do not terminate in the annulus building.
Direct containment structure leakage prior to depressurization of the annulus building can occur for approximately 23 sec following startup of the SLCRS (Fig. 6 . 2. 3.1-2) or far approximately 38 see following the accident. Bypass leakage through containment isolation valves for all systems not terminating within the annulus building has been - 6.2 '8 . - c Amendment 24
/, ,l 3 / 4/23/76
SWESSAR-P1 cvaluated and is included in Table 6.2.3.1-5. The amount of potential leakage of each valve has been conservatively calculated by using the maximum permissible leakage allowed by ASME FI, Subsection IWV. All lines will be tested with gas pet IW\ . These values are indicated in Table 6.2.3.1-5 and their sur . nation results in a total passible bypass leakage of m proximately 0.0023 percent per day. The effect of this negligible amount of uncollected leakage on the calculation of the offsite radiological consequences of the accident is not significant. The containment purge air system (Section 9.4. 5. 2) is vented to the annulus building atmosphere during normal operation and therefore is not considered a bypass leakage path. Section 16.4.4 describes the leak test program performed to determine the actual bypass leakage. The capability of the SLCRS is based on the following data. The results from Type A tests conducted at various power plants in accordance with 10CFR50, Appendix J, Reactor Containment Leakage Testing for Water Cooled Power Reactors, are listed in Table 6.2.3.1-2. In all cases but two, the measured leak rate (Lcm) was 50 percent of the allowable limit or less. For the exception, Monticello, it is noted that main steam isolation valve repairs were required (these valves are cart of the RCPB for a BWR). For Dresden 1, it is noted that the initial test was done with an open line. For H. B. Robinson No 2, two 6 in. valves were repaired, decreasing the leak rate to 29 percent of the limit. It is concluded from these tests that the Type A measured leak rate (Lam) is usually less than 50 percent of the allowable leak rate (La) unless unusual and excessive valve leakage occurs. Table 6.2.3.1-3 lists the results of Type B and C testing conducted in accordance with Appendix J. The units which did pa ss as tested were at 48 percent or more of the limit (.6 La) for this test. For those which did fail, repair work on valves was sufficient to reduce the leakage to meet the limit. It is concluded from these data that containment leakage is dependent on the leaktightness of valves and penetrations. The periodic Type A test is an excellent method of detecting excessive leakage from the containment as pointed out by the comments in Table 6.2.3.1-2. Both the Type A and B & C tests perf ormed indicate that valves and penetrations are the major 6.2-39 , ^ n Amendment 24 OOJ LUd 4/23/76
SWESSAR-P1 source of leakage f rom the containment. In fact, the Type A and B S C leakage are of the same order of magnitude as shown by the percent La averages in Tables 6. 2.3.1-2 and 3. The penetration leakage typically represents the total leakage from the containment. The SLCRS complements the integrity of t? 9 containment structure by collecting and filtering all leakage from the containment penetrations following a DBA. The potential for leakage through the containment liner directly to the environment is considered negligible. The carbon steel liner for the containment walls is 3/8 in thick, and the liner for the dome is 1/2 in. thick. The liner plates act as a gastight membrane under any condition that can be encountered throughout the operating life of a plant. The liner plate is anchored to the concrete containment and at sufficiently close intervals so that overall deformation of the liner is essentially the same as that of the concrete containment. It is protected from potential interior missiles by interior concrete shield walls. The steel liner for the containment is described more fully in the following sections: Section No. General Description 3.8.1.1.3 Applicable Codes, Standards, And Specifications 3.8.1.2.2 Loads and Loading Combinations 3.8.1.3.2 Design and Analysis Procedures 3.8.1.4.2 Appendix 3B Structural Acceptance Criteria 3.8.1.5.2 Materials, Quality Control, and Special 3.8.1.6.3 Construction Techniques The liner welds meet the following codes and standards as sammarized in Section 3.8.1.2: ACI-359 " Proposed Standard Code for Concrete Reactor Vessels and Containments - Subsection CC-1000 to Subsection CC-6000," November 1974 Draft. Regulatory Guide 1.19 Nondestructive Examination of Liner Welds (Section 3A.1-1.19) . ASME Boiler and Pressure Vessel Code Section II Matcrial Specification Section III Nuclear Power Plant Components (referenced as ACI-3 5 9)
, .n .
Lu _. ci 6.2-40 Amendment 16 8/29/75
SWESSAR-P1 Section IX Welding and Brazing Qualifications Furthermore, the liner weld joints for the shell are overlapped with test channels which are welded to the shell as shown on Fig. 3.8.1-9. A series of test channels are welded together with dams welded on open ends. The test channels are pressurized to 50 psig to verify the liner weld integrity. The acceptance criterion is no leakage for a perioc of 120 min. Another consideration for evaluating the potential for bypassing the SLCRS is the total liner surface covered by the annulus build.ing . The annulus building which surrounds the containment covers 45 percent of the containment liner surface area. Because of the concentration of the liner welds around the penetrations, the annulus building covers nearly 65 percent of the linear length of the liner welds. Based on the descriptive analysis presented above, it is concluded that the leukage from the containment structure will be primarily through valves and penetrations and will be entirely collected and processed by the SLCRS with the exception of the bypass leakage previously described. To provide a conservative 16 safety analysis, credit for only 50 percent collection of containment leakage is assumed in the analysis of the LOCA (Section 15.1.13) . The design leakage rate for the centainment is 0.2 percent / day; thus the Appendix J limit on containment leakage is 75 percent of 0.2 percent / day o" 0.15 percent / day. Since the SLCRS takes credit for only 50 percent collection of containment leakage, Type A measured leakage as described in the technical specification (Se ction 16.4.4) will be limited to 0.1 percent / day uncollected leakage which is less than the 0.15 percent / day limit of Appendix J. Since the 75 percent factor has been applied to the containment design leakage rate of 0.2 percent / day, it need not be applied to the 0.1 percent / day uncollected leakage. Justification of this position is included in the following paragraphs. 16 The shell leakuge of a conteinment structure has never been directly measured and is probably below the measureable limit. Past experience has shown that it has always been possible to meet Type A test limits, even af ter previous attempts had f ailed, by reducing leakage through valves and penetrations . No modifications of the containment shell have ever been required to reduce leakage. The imposition of an additional 25 percent reduction on the uncollected leakage would therefore result in an unnecessarily conservative test of valves and penetrations , leakage from which is collected by the SLCRS. If a 75 percent factor is applied in Appendix J for @' considerations of thermal and seismic stresses ano deterioration over tim e , these have a f or greater effect on the valves and 6.2-41 ,77 on9 Amendment 16 It 0 ) L0L 8/29/75
SWESSAR-P1 penetra tions . Effects of these factors on the shell would be negligible. However, even if they did significantly affect the shell, lowering the limit on Lam for these reasons could only lead to reductions of leakage through valves and penetrations. It must be kept in mind that the definition of 0.1 percent / day uncollected leakage is an assumption used for analysis of the radiological consequences of the accident. For all this 0.1 percent leakage to actually represent uncollected leakage, the leakage throtxJh valves and penetrations would be required to be zero, obviously an improbable situat_On. Also, the SLCRS is effective in removing only the radiciodines from the leakage; the noble gases are not removed from the exhaust flow. Based on the definition of 0.1 percent collected and processed through 95 percent efficient filters and 0.1 percent released unprocessed, tha radioiodine release is O proportional to 0.005 pe; cent plus 0.1 percent, or 0.105 percent. If a facade were added to the basic arrangement (Option A of SWESSAR-P 1) , credit would be given for a ftF1 dual containment. In this case, all the leakage would be filtered, so that the radioiodine release would be proportional to 0.005 percent plus 0.005 percent, or 0.01 percent, which is one tenth of the base case. Alternatively, the maximtz:t allowable leak rate, La, could be increased to 2 percent, which if filtered again results in a radiciodine release proportional to 0.1 percent. Therefore, placing a facade over the part of the building that does not leak would allow the maximum allowable leak rate to be increased by a factor of 10, and the resultant radioiodine release from the structure by a factor approaching 10. This illustrates the conservatism already present in assuming that half the leakage is uncollected. Placing an additional conservatism on the uncollecteu leakage is thus unwarranted. The SWESSAR-P1 containment shell is of the same design as used on previous Stone & Webster plants. The shell for SWESSAR-P1 could conceivably have an even lower leak rate than in previous Stone & Webster designs, since the cylinder walls are 9 in. thicker. The SLCRS is designed to operate during the reactor refueling period, a fuel handling accident, and a IOCA. Fig. 6 . 2 . 3 .1--2 indicates pressure as a function of time in the annulus building, fuel building, and all contiguous areas served by the SLCRS following a LOCA considering a single failure as the failure of one train to start. No credit is taken for the capability of the SLCRS to mix the atmosphere in the annulus , te; tu> 6.2-42 Amendment 17 9/30/75
SWESSAR-P1 building, fuel building, and contiguous areas when evaluating radiological consaluences. A control room alarm sounds when any exterior door to the areas served by the SLCRS is opened. For further discussion, see Section 16.6. The active components in the system are redundant, and the ductwork provided to each train is also redundant, thus meeting the requirements of Regulatory Guide 1.52 (Section 3A.1.-1.52) . The Seismic Category I ductwork in the SLCRS is protected against passive failure due to seismic events and the failure of nearby piping and ductuork components. The design of the ductwork and support system precludes a passive failure from buckling of the duct and yielding of the duct and/or supports. To ensure protection from loss of function due to common events, the filter banks are physically separated with barriers placed between then. See Fig. 6.2.3.1-1 for indication of the damper failure position. The analytical model used in the analysis is based on the assumption that infiltration is proportional to differential pressure created by the exhaust fan between the inside and outside of the annulus building, fuel building, and all the contiguous areas following an accident. A detailed description of the analysis and assumptions used in the pressure analysis is included in Appendix 6A.3. 6.2.3.1.4 Testing and Inspections SLCRS components are tested and inspected both as separate compo. tents and as an integrated system. Instrument readings are taken to ensure that all air systems are balanced to exhaust the regr. ired air quantities at design conditions. Di f ferential pressure gages are strategically located throughout the annulus building, fuel building, and all the contiguous areas to demonstrate that a negative pressure will exist within the prescribed time. Gages are also provided for the enclosure building under Option A or B. 18 The preoperational and periodic inservice tests, to verify SLCRS capability to maintain the negative pressure within the required time, are conducted for (1) the annulus building (including electrical tunnels and main steam and feedwater valve enclosures) , (2) the fuel building, and (3) the annulus and fuel buildings together because of three possible SLCRS modes of operation. During the tests, the affected areas are isolated, and space ventilation systems stopped. 6 . 2 -4 2 A {[j {]4 Amendment 18 10/30/75
SWESSAR-P1 All door gaskets and structural joint seals are periodically inspected and replaced when required. The periodic leakage test to verify SLCRS capability to maintain the negative pressure in all related areas is performed with at least the same frequency as containment Type A testing. Capacity and performance of fans conform to the conditions and ratings and comply required with Air Moving and Conditioning Association (AMCA) test codes. SLCRS ductwork is leak tested after installation to ensure against any potential. The ductwork is bypass of all welded construction and is pressure tested to 1.5 times the fan zero discharge pressure. Every HEPA filter cell is given a standard diocty1phthalate (DOP) amoke test with 0.3 micron smoke particle diameter before leaving the manuf acturer's f acilities. A similar test is conducted af ter installation at the site to ensure that there is no leakage from upstream to downstream of the HEPA filter. Provision is made to inject a DOP test at the inlet to the HEPA filter. Each carbon using a Halon adsorber 11 and air is factory tested for efficiency and leakage mixture introduced upstream of the carbon adsorber type to confirm that and a halogen detector of the gas chromatograph the removal ef ficiencies are met. Filter cells are individually ds an integrated system a ter tested after fabrication and again ins tallation . Filter banks are periodically tested for leakage and efficiency while in place and defective in cells are replaced and all leal- eliminated. One cell each carbon adsorber bank is composed of test canisters which are removed annually in accordance with Regulatory Guida 1.52 (Section 3A.1-1.52) . Fans, motor operated dampers, and c;ntrols are tested periodically. Periodically each of the started atuomatically by redundant systems is a reach rated speed with all dampers in the simulated CIA signal and allowed to before being shut down. operating position 6.2.3.1.5 Instrumentation Applications Both SLCRS exhaust fans start on receipt of a CIA or a fuel building airborne radiation monitor signal. Only one fan is p required to meet the partial vacuum requirement of Section 6. 2. 3.1.1 (1) . The operator may, at his option, shut down eit.er of the exhaust fans. The SLCRS exhaust fans may also be started and stopped from the control room. A charcoal whenever air filter flowdecay through heata removal fan is started automatically filter bank is stopped. It is manually stopped following decay heat elimination. c; ' . 6.2-42B Amendment 24 4/23/76
SWESSAR-P1 A radiation monitor downstream of the filters and temperature and differential pressure indicators across the filters warn the operator of a potential problem so that he may take appropriate action which may include manual transfer to the standby filter bank. All outside doors of the annulus and fuel buildings and all doors between the annulus and fuel buildings have electrical switches which display any door opening on a control board. In addition to the visual indication on the control board, an alarm with a time delay will be energized to initiate necessary action related to door closure. G 6.2-42C , > a n ,r Amendment 24 buJ cuO 4/23/76
SWESSAR-P1 6.2.3.2 Containment Spray System - Iodine Removal The containment spray system (Section 6.2.2) provides water spray -to the containment during the unlikely event of a LOCA to depressurize the containment and to minimize the release of radioactive iodine and particulates to the environment. This section describes the iodine removal capability of the containment spray system. The analysis of the radiological consequences of the LOCA is given in Section 15.1.13. 6.2.3.2.1 Design Basis The design bases of the containment spray system for removing iodine f rcxn the containment atmosphere are:
- 1. The amount of radioactive iodine in the containment following a DER of the hot leg using TID-14844 source terms shall be reduced so that the outleakage results in a thyroid dose below the recommended limits of 10CFR100 (see Section 15.1.13). This accident is more severe than the cold leg DER since mass transfer to the spray droplets is at a slower rate because of the lower containment tE5nperatures encountered.
- 2. The worst single active failure shall not result in a reduction in containment spray coverage within the 24 containment or loss of item 1 above.
- 3. The minimum spray coverage of the containment f rom the 4rrangement of nozzles and headers at the elevated temperatures and pressures during an accident shall be approximately 95 percent of the cross sectional area above the operating floor and 90 percent of the ovcrall containment volume.
- 4. The minimum pH in the containment sump water duri .g the recirculation phase shall be 8.5.
- 5. The spray mixture during the in jection phase is a mixture of sodium hydroxide and boric acid with a minimum oH of 9.0 and a maximum pH of 11.0.
6.2.3.2.2 System Design The containment spray system consists of parallel and separate subsystems. Each subsystem consists of one spray pump, two spray headers, spray nozzles, one ESF sump, and associated piping and valves. The containment spray system design is discussed in detail in Section 6.2.2 and component data are given in Table 6.2.2-1. Sodium hydroxide (NaOH) solution from the refueling water CAT is added to the boric acid from the RWST as it enters a subsystem.
/ 7 on7 f
6.2-43 i '
'l Amendment 24 4/23/76
SWESSAR-P1 This is done to improve removal of iodine from the containment atmosphere by sprays during an accident, by raising the pH of the spray, and to buffer the boric acid. The NaOH is mixed uniformly with the boric acid by the turbulent flow. The initial pH of the spray (injection phase) is a minimum of 9.0. The final pH of the spray (recirculation phase) is a minimum of 8.5. Since for Westinghouse (RESAR-41) , the RWST and the CAT are in hydrostatic balance, the flow from the CAT into the CSS, Fcat, is: Fcat = BR (Fcss + Feces) where B = ratio of inside diameters squared (CAT /RWST) R = ratio of liquid densities OTNST/ CAT) Fcss = flow from RWST into CSS Feccs = flow from RWST into ECCS. The spray additive flow rates are given in Table 6.2.3.2-3. The g chemical compositions of the containment spray and sump solutions are given in Tables 6.2.3.2-4 and 6.2.3.2-5 (W-41 only) . The pH of the spray solution is calculated using the chemical equilibrium involved as outlined in Reference 17 for all flow conditions. These equations have been checked against experimental data (Reference 18). The mean spray droplet diameter used in the calculations was 1,000 microns. From Fig. 6.2.2-12, 96 percent of all droplets have less than a 1,000 micron diameter. The location of the spray headers and the arrangement of nozzles are given in Table 6.2.2-1. The nozzles used ace the Spray Engineering Company 1713A nozzle, or equivalent. Noiv.le supplier typical data can be found in Reference 19. The calculated spray g coverage of the containment is given in Section 6A.2. The chronology of events for the system is estimated in Table 6.2.1-1. 6.2.3.2.3 Design Evaluation The rate at which elemental iodine can be removed from the containment atmosphere by a rea ctive chemical spray can be calculated by Griffith's methodC8)~or by Parsly's computerized method,ca). Both methods are based on the experimental work of Taylor (3), who showed that overall mas", transfer rate at which elemental iodine is transferred into reactive solutions is controlled by the gas film resistance cc the vapors surrounding the spray drops, and of Ranz and MarshallC')C5), who developed a
, 7 OUa a nu ,d m
6.2-44 Amendment 24 4/23/76
SWESSAR-P1 correlation for calculating the mass transfer coefficient when the rate of transfer is controlled by the gas film resistance. The fundamental assumption in using an alkaline solution as a reactive spray to remove iod:ne vapor from the containment is illustrated in the following equations: OH +I O I + HOI 2 3 HOI : 21 + 3H + IO- 3 Two different cases are considered. In the first case, mass transf er is assumed to be gas-film limited, so that there is no resistance to mass transfer of iodine within the liquid drop because chemical reactions prevent the existence of elemental iodine in the drop. As soon as I2 has undergone reaction, it can no longer exert a significant partial pressure, thereby changing the partition very much in favor of the liquid phase and increasing the mass transfer rate. Experimental evidence from ORNL and BNWL tests (6)c7)(a)(e) confirms these theoretical 9 6.2-44A. , Amendment 22 l'..J on/ m u 3/17/76
SWESSAR-P1 sssumptions for NaOH-borated water solutions. In the second case resistance to mass transfer is assumed in both the gas film immediately outside the drop and also in the liquid immediately beneath the surface of the drop. In Case I, the gas-film-limited case, the mass flux is expressed in terms of a gas deposition velocity, whereas in Case II, an overall deposition velocity is used. All other factors in the expression (liquid flow rate, header height, contatnment volume, drop size, and terminal velocity) are independent of whether the process is gas-film or gas-and liquid-film controlled. The derivations that follow are indicative of the correlations used in the references and are presented here to show the methods used for computing the iodine removal coefficients. In the derivation it is assumed that iodine is uniformly mixed throughout the containment atmosphere and that all spray drops are spheres of the mean surface diameter. The assumption for uniform mixing is valid since the spray coverage for the arrangement of the spray headers and spray nozzles is approximately 95 percent of the containment. In the gas-film controlled case (Case I) , the dependence of mass transfer rate on film conditions is expressed by: V g= (2 + 0.6 Re ! Sc ! ) , cm/sec (1) Equation (1) is well substantiated by experiments in a variety of systems in which the gas film is controlling, as reported by Ranz and Marshall (*)(5). The surface area of drops (Sd) available for iodine absorption can be calculated from equation (2) which is based on the conservative assumption that all the drops are spherical and can be assigned a uniform certain diameter: nd 2 Fh 6Fh g _ AdFt = = , cm 2 (2) d Vd nd V/6 3 dV The containment atmosphere is assumed to be well mixed and all the drops are assumed to contain an excess of chemical reagent to react with the iodine and convert it to a practically nonvolatile form. Based on the foregoing, the rate of removal of elemental icdine from the containment atmosphere can be calculated on the basis of an exponential removal as the spray passes through the containment atmosphere by the relationship: C = Co ex -At (3) 6:: 2;o 6.2-45 Amendment 5 12/2/74
SNESSAR-P1 The iodine removal coefficient, A , is calculated by the relationship: A = Vgsd , sec _1 Vc (4) for the gas-film controlled case. Combining equations (4) and (2) : 6FhV A= 9 ' sec -1 VVdc (5) Using equation (5) for A, the calculated values of A were within 20 percent of the measured values for the containment systems experiment (11). For the conservative gas-and-liquid-film controlled case (Case II) , the gas film deposition velocity Vg is replaced by the overall deposition velocity Vd defined by the relationship: 1 1 1
-- = -- + (6)
V9 Vg KHL 1:here: 2n2 D g 9 KL- 3 d (7) and equation (5) becomes: 6FhV D (8) A= , sec _1 VVdc Using the gas-and liquid-film controlled case, the value for H is estimated. This is the gas-liquid iodine partition coefficient. A value for H can be predicted from experimental measurements and from chemical reaction equilibrium constants for iodine in water (23) 22). The values of H reported from the containment systems experiment, from experiment measurements, and frcxn considering only the rapid reactions of iodine in water are quite similar and lie between 104 and 105(82). For conservatism, H is taken to be 8,000. For long times (> 1/2 hour ) , the lowest allowed value of the amount the value of of iodine in the containment atmosphere will be limited by H in the following expression: C Vc Co Vc + HVs (9) f 7 c. c ; b0J L-l 6.2-46 Amendment 5 12/2/74
SWESSAR-P1 The values for the terms of Table 6.2.3.2-1 which represent the design conditions for the containment spray system and for the calculated iodine spray removal coefficient, A , are given in Table 6.2.3.2-2. Case I calculations involve equations (1) and (5). Case II calculations involve equations (1), (6) , (7) , and (8). Case III is the value of A calculated for Case II divided by a cumulative f actor of conservatism discussed below. The uncertainty in the predicted value of A is accounted for by dividing by a cumulative factor of conservatism. The quotient, AS, is the iodine spray removal constant which is used in the dose calculations for the LOCA (see Sectfon 15.1.13). Factors of uncertainty are assigned to each parameter in the expression for X , in terms of minimum performance and maximum excursion values, as suggested by Reterence 13. The cumulative l8 factor of conservatism is calculated by taking the square root of the sum of the sauares of the f actors of uncerta inty. These factors are summarized as follows:
- 1. Flow - A minimum expected value of 90 percent of rated flow for minimum safety features is used to account for diminished flow due to system construction, fluid density and viscosity variaticns, and possible system damage and
- 2. Fall Height - An uncertainty of 20 percent is included for consideration of the limited number of drops which strike interior surfaces and do not attain the minimum stated f all height.
- 3. Deposition velocity -
An uncertainty of 50 percent accounts for uncertainties in the calculation of h and K .
- 4. Terminal Velocity -
An uncertainty of 50 percent dCCountS for uncertainties in the calculati.on of V .
- 5. Containment Volume - With 95 percent of the containment volume covered by spray with minimum safety features operating, the uncertainty concerned with mixing and unitormity of distribution and with local depletion due to " channel effects" of successive spray drops is 10 percent.
- 6. Drop Diameter - An uncertainty of 50 percent is included for uncertainties in the spray drop distribution for normal and accident conditions and for drop size accretion from steam condensation and drop coalescence.
The average drop size accretion from steam condensation has been estimated at 10 percentca) and for drop coalescence, at 10 percent (20). 6.2-47 Amendment 8
, neq 3/28/75 (O$ l'L
SWESSAR-P1 This summary and the individual factors of uncertainty used are dl consistent with the Regulatory Staff evaluation for Diablo Canyon Unit 2(a). The camulative f actor of conservatism is 3.3. In actual tests run to date at ORNL and BNWL (7)('), the experimental half-lives for iodine removal were seldom found to exceed 1 minute for the variety of spray solutions tested, which included NaOH -H3B03 spray solutions. To estimate the expected spray removal coefficients f or this containment from those measured in the containment systems experiment,(21) Equations (5) &nd (8) can be rewritten in the torm: 1 KP (gpm) h ( f t) x s (min)
= (10) d (cm) Vc (f t3) where K is a constant.
The minirmm value of K for the containment systems experiment is 0.35.(11) Therefore, for this containment, the expected is on the order of 1.2/ min or 72/ hour. This expected is greater than the which is rsed in the thyroid dose calculations for the LOCA (see Section 15.1.13) by approximately the cumulativa factor of conservatism. 6.2.3.2.4 Tests and Inspections The tests and inspections of the containment spray system are described in Section 6. 2. 2. 4. 6.2.3.2.5 Instrumentation Application The Instrumentation application of the containment spray system is given in Section 6.2.2.5. 6.2.3.2.6 Materials It ts current practice t i enhance the ability of the sprays to remove iodine f rom the containment atmosphere by the addition of either sodium thiosulf ate or sodium hydroxide, although "xperiments indicate that the low pH borate solution from the nWST is almost as effective in removing iodine from the containment a tmosph ere . Sodium thiosultate quantitatively absorbs to iodine in un irreversible action, but is more susceptible radiation and may cause problems in long-term storage. Theretore, sodium hydroxide is the pref erred spray additive. The addition of soditu hydroxide to the spray water, which subsequently mixes with the water spilled trom the reactor coolant system, results in a final pH in the ESF sumps o approximately 8. 5. ,
- - -)
6.2-48 Amendment 8 3/28/75
SWESSAR-P1 8 The chemical composition and volume of the CAT and the RWST are given in Table 6.2.2-1. The chemical composition of the containment spray solution and the containment sump solution is given in Table 6.2.3.2-4. This table is based on the flow rates in Table 6.2.3.2-3 and the data on the accumulators and boron injection tank in Table 6.1-.7. NaOH - borate solutions show little change at high temperatures (130 C) with or without radiation (1 + )( 15 )(16 ) . The solutions are not susceptible to \B significant radiolytic or pyrolytic decomposition under conditions found in nuclear pcwer plant containments. References for Section 6.2.3.2
- 1. Griffiths V., "The Removal of Iodine from the Atmosphere by Sprays," UKAEA REPORT ARSB (s; R45, 1963.
- 2. Parsly, L.F., Jr. " Removal ot Elemental Iodine from Steam-Air Atmospheres by Reactive Sprays," USAEC Report O RNL-TM-1911, 1967.
- 3. Taylor, R.F., " Absorption of Iodine Vapors by Aqueous Solutions," Chem. Engrg. Sci. 10, pp 68-79, 1959.
- 4. Ranz, W. E. and Marshall, W. R . , Jr . , " Evaporation from Drous. Part 1," Chem. Engrg. Prog., 48 No. 3, pp 141-146, 1952.
- 5. Ranz, W.E. and Marshall, Jr., W.R., " Evaporation from Drops, Part 1," Chem. Engrg. Prog., 48, No. 4, pp 173-180, 1952.
- 6. Parsly, L.F. Jr. anc Fran reb, J.K. , " Removal of Iodine Vapor from Air and Steam-Air Atmosoheres in the Nuclear Safety Pilot Plant by Use of Sprays," USAEC haport ORNL-4252, 1968.
- 7. Row, T.H., " Spray and Pool Absorption Technology Prcqram,"
USAEC Report OPNLc4360, p 29, April 1969.
- 8. "E f f ectiveness of Spray System for Iodine Cleanup," A. % DRL Staf f Evaluation for Diablo Canyon Unit 2.
- 9. Cottrell, Wm . B . , "ORNL Nuclear Safety Research and Develop-ment Program Bimonthly Report for January - February 1970,"
USAEC Report ORNL-TM-2919, p 73, May 1970.
- 10. Pasedog, W.F. and Gallagher, J.L., " Drop Size Distribution and Spray Ef f ectiveness," Nuclear Technology, Vol. 10, p 418, April 1971.
- 11. Hilliard, R.K., et. al. " Containment Systems Experiment-Interim Report," BNWL-1244, pp 149-160, Feb. 1970.
- 12. Parsly, L. F., " Design Considerations of Reactor Containment h Spray Systems -
Part IV. Calculation of Iodine-Water Partition Coefficients," USAEC Report ORNL-TM-2412, Part IV, January 1970. ,,, ,, 6.2-49 6Oj c,9 Amendment 8 3/28/75
SWESSAR-P1
- 13. " Nuclear Safety Quarterly Report, February, March, April, 1970," BNWL-1315-2, May 1970.
- 14. Fittel, H.E. and Row, T.H., " Radiation and Thermal Stability of Spray Solutions," Nuclear Technology, April 1971, p 442.
- 15. Griess, J.C. and Bacarella, A.L., " Design Considerations of Reactor Containment Spray System - Part III The Corrosion of Materials in Spray Solutions," ORNL-TM-2412, Part III, 1969, p 15.
- 16. Row, T.H., " Spray and Pool Adsorption Technology Program,"
ORNL-4 360, 1969, p 21 and 22.
- 17. Byrnes, D.E., "Some Physiochemical Studies of Boric Acid Solutions at High Temperatures," WCAP 3713, September 1962.
- 18. Bell, M.J., Bulkowski, J.E. and Picone, L.F., " Investigation 22 of Chemical Additives for Reactor Containment Sprays,"
WCAP 7153-A, April 1975.
- 19. Spray Engineering Company, Cambridge, Mass. , Catalog 1713A.
O
,Q L' J
'S
"' til 6.2-50 Amendment 22 3/17/76
SWESSAR-P1 6.2.4 Conta inment Isolation Systems The containment isolation system isolates piping and instrument lines penetrating the containment boundary during a LOCA. This prevents radioactive release to the environment throuch these lines . 6.2.4.1 Design Bases The design bases for the containment isolation system are:
- 1. The containment isolation system shall meet the intent of AEC General Design Criteria 54, 55, 56, and 57 of Appendix A to 10CFR50 (Sections 3.1.54-57).
- 2. Instrument lines penetrating the containment shall be in accordance with Regulatory Guide 1.11 (Section 3A.1-1.11) .
- 3. The containment isolation valve arrangement shall ensure containment integrity assumina the occurrence <,f a single failure by providing at least two barriers between the atmosphere outside the containment and the containment atmosphere or the reactor coolant pressure boundary
- 4. Containment isolation valves shall be physically located within 10 ft of the containment wall if p ra ctical, thereby minimizing the length of piping between the valves and their penetrations.
- 5. Air-operated containment isolation valves shall f ail in the closed position on loss of control voltage to the associated solenoid valve or on loss of air. The closed position of all air operated containment isolation valves shall be the position of greatest safety. Motor-operated containment isolation valves shall fail "as-is."
- 6. The design pressure of piping and components within the isolation boundaries afforded by the contai nment isolation system shall be equal to, or greater than, the design pressure of the containment structure. Piping, valves, and containment penetrations shall be designed as QA Category I and Seismic Category I and a re constructed and installed in accordance with the requirements of ASME Section III, Code Class ?. (See Section 3.2.)
- 7. Containment isolation system components, including valves , controls, piping, and penetrations, shall be protected from internally or externally generated missiles, waterjets, and pipe whip. Details regarding 6.2-51 I', Amendment 5 12/2/74
SWESSAR-P1 the design of this protection are given in Sections 3.5 and 3.6. 6.2.4.2 System Design Table 6.2.4-1 lists the design details of the containment isolation systens for each individual penetration. The penetrations are listed by system in accordance with GDC55, 56,, and 57. Table 6.2.4--2 lists the instrument lines penetrating the containment; these lines are in accordance with Regulatory Guide 1.11 (Section 3A.1-1.11) . Fig . 6 . 2 . 4-1 shows the design details of each type of penetration listed in Table 6.2.4-1; in addition, each system diagram in this SAR clearly delineates the boundary of the containment structure and the containment isolation valves for each penetration. Table 6.2.4-3 lists the required testing for fluid lines penetrating the containment structure in accordance with the requirements of 10CFR50, Appendix J, Section II H, as clarified by Draft Standard ANS N274, " Reactor Containment Leakage Testing Requirements," dated April 22, 1974. Most of the containment isolation system conforms to the specifications of GDC 54 through 57. However, it is necessary to deviate from the specific arrangements described in the criteria in the following case:
- 1. ESF Sump Penetrations The ESF sump penetrations are for the suction piping of the high head safety injection, low head satety injection, and containment spray pumps. To ensare that water can be taken from the containment floor during the recirculation phase following a DBA, these sumps and piping penetrations are located in the containment mat.
Inside containment isolation valves are not only impractical, but defeat the intended purpose of these 12nes. Suction line isolation during the recirculation phase is required only in the event of a pipe rupture outside the con N inment. Further description of the ESF is given in other sections of this chapter. Table 6.2.4-3 states that the isolation valves in the main steam, feedwater, auxiliary feedwater, and steam generator blowdown systems do not require Type C testing. Appendix J to 10CFR50, Section II H, does not require these valves to be tested. Table 6.2.3.1-5 does not include these systems as leakage paths that 33 can bypass the SLCRS because the steam generator secondary side will be maintained at a higher preasure than the primary side in the long term following a IDCA. As discussed in Section 10.4.10, the auxiliary feedwater system is capable of maintaining the steam generator secondary side pressure higher than the primary side pressure following a LOCA. i, ,- - h .) 4- / I 6.2-52 Amendment 33 6/30/77
SWESSAR-P1 In the short term, once the primary side has blown down to containme:.t, the pressure dif ference can be maintained by normal auxiliary feedwater system operation. In the long term, once the primary side pressure has been reduced to atmospheric pressure, the pressure difference can be maintained by placing the steam generators in a flooded condition. A head of water over the tubes will maintain the pressure difference. A long term emergency water supply is provided to the auxiliary feedwater pump suction lines by the reactor plant service water system. This emergency supply is the ultimate heat sink. As such, it contains more than enough water to serve its intended heat removal function as well as having an inventory to allow makeup to the steam generators and maintain flooded conditions. For the Nestinghouse and C-E steam generators, a water level above the tube bundles can be demonstrated by observation of the 33 safety related level instrumentation on the steam generators. Pbr the BSW steam generators, maximum secondary side water level can be demonstrated when water discharges from the level sensing connection at the upper tube sheet, and when the safety related pressure instrumentation on the steam generators shows a pressure increase caused by the auxiliary feedwater pump approaching shutoff pressure. For all plants, the main steam piping loops to an elevation higher than the maximum steam generator level, thus ensuring the capability to maintain the maximum level. During the post-LOCA period, the higher pressure on the steam generator secondary side prevents primary-to-secondary leakage. Were is no Appendix J requirement to test the ability to maintain a post-IDCA flooded condition, nor to perform a hydrostatic test to ensure leak tightness of the secondary sysum valves. However, the ability to place the steam generators in a flooded condition is demonstrated when the steam generators are placed into wet layup. Compliance of the containment pressure monitoring system line penetrating the containment structure with Regulatory Guide 1.11 (Section 3A.1.11) is discussed in Section 7.3.3.9. The design, protection, and location of these Safety Class 2 lines minimize the likelihood of accidental damage, prevent f ailure from pipe whip, and allow periodic visual inspection. W e lines are sized (or provided with restricting orifices) to restrict leakage flow from the containment so as not to significantly af f ect of f site doses (in the event of a line rupture or isolation valve failure).
?DO Amendment 33 6.2-52A
{' U J U 6/30/77
SWESSAR-P1 A wide variation exists in the accident severity that would necessitate a plant shutdown: consequently, nhe containment isolation system is actuated at two stages depending on the severity of the accidents. The CIA and CIB signale provide this selectivity. The purpose of the CIA signal is to isolate sytems that are not required for an orderly and safe shutdown of the plant to protect equipment required for the CIA signal has to resume operation once the initiating cause been corrected. The safety injection signal (SIS) initiates the CIA signal and the emergency core cooling system (NSSS scope) . In the event of a major LOCA, after which normal heat removal systems do not provide adequate core and containment cooling, a CIb signal follows the CIA signal to complete the containment isolation (except for isolation valves required to be open for operation of ESF systems) . Containment penetrations for ESF tystems, which mitigate the consequences of an accident, are opened or closed, as required, to allow systems operation. Remotely operated containment isolation valves (actuated by an ESF actuation signal or remote-manually operated) have position 9 6.2-53 Amendment 17 7/7 pGg 9/30/75
SWESSAR-P1 indication and a manual control switch in the control room. Means are provided for manual initiation at the system level of the CIA and CIB functions from the control room. The arrangement intent is to keep the valves as close as possible to the containment. However, due to the number of valves, valve size, and accessibility requirement for inservice inspection, in some cases the isolation valve location is more than 10 ft from the containment wall. The only containment isolation valves which presently exceed 10 ft are the main steam isolation valves. A fuel transfer tube in the fuel transfer penetration between the refueling cavity in the containment and the fuel pool in the ~. u el building is fitted with a normally closed gate valve in the fuel pool and a blind flange in the refueling cavity to prevent leakage through the transfer tube during accident conditions. Spare penetrations are sealed with a welded closure. Descriptions of the design of piping, electrical, and containment access penetrations are given in Section 3.8.1. The most severe containment environmental conditions occur during a DBA. These conditions are stated in Section 3.11. Containment isolation valves inside the. containment are designed to maintain their integrity and leaktightness under these conditions. 6.2.4.3 Design Evaluation The design of the containment isolation system meets the requirements for system integrity, response, operation, and reliability. Isolation valve and piping design and location ensure containment integrity for any postulated accidents inside the containment. The design and internal arrangement of the annulus building provide protection for outside containment isolation valves from externally and internally generated missiles. All containment isolation valves are specif.ted to be able to operate and/or perform their intended safety function in the worst environmental conditions postulated to occur in the area in which the valve is located. Containment isolation valve types are selected on the basis of fluid system requirements, failure mode, seat leaktightness, closure time requirements, and past experience. Gate valves are used extensively for remotely operated containment isolation valves because of their deep stuffing box features for handling radioactive fluid, and availability in a larger range of pressure ratings and sizes. Globe valves are used exclusively on closed systems as containment isolation trip valves be :ause of the short stem travel and, therefore, quick closing capability.
, , , nn be. s t) 6.2-54 Amendment 5 12/2/74
SWESSAR-P1 Containment isolation trip valves are designed to be operable under normal operating environmental conditions during the life of the plant and during seismic conditions. They are designed to trip closed at the onset of a DBA and remain closed during post accident environmental conditions. All containment isolation valves which receive signals to close from either CIA or CT- have valve closure times as fast as possible consistent with the type of the valves and valve operators with consideration given to water hammer effects. Valve closure time for each valve is shown in Table 6.2.4-1. 6.2.4.4 Tests and Inspections All valves in the containment isolation system require testing as specified in Sections 16.4.2 and 16.4.4. 6.2.4.5 Materials The containment isolation system components are constructed from materials that can withstand the environmet tal conditions to which they may be exposed, as listed in Sect 3on 3.11. For the isolation valves inside the containment, these conditions include those prevailing during the 40 year lif e expectancy plus the post-DBA conditions of pressure, temperature, humidity, containment spray, and radiation. For the components outside the containment, including isolation valves and the pressure transmittera and switches for the ESFAS, the materials must withstand the pressure, temperature, and humidity conditions. The procurement specifications for components specify the design environmental conditions of each component. Each specification either: 1) specifies materials that are known to meet the specified design conditions, or 2) requires that the manuf acturer supply data which demonstrate that the materials to be used meet the design conditions or that they will not affect the safe operation of the canponent and the system as a whole. 6.2.5 Combustible Gas Control in Containment The combustible gas control system monitors the hydrogen concentration within the containment and maintains this concentration at a safe level following a LOCA. The combustible gas control system is shown in Fig. 6.2. 5-1 and the system component data are given in Table 6.2.5-1. nI (' t J >V 6.2-55 Amendment 8 3/28/75
SNESSAR-P1 6.2.5.1 Design Euses The design bases for the combustible gas control system are:
- 1. The ystem shall process containment atmosphere at a ra - -uf ficient to maintain the hydrogen ecncentration in the containment atmosphere below 4 volume percent 16I hydrogen following a DBA.
- 2. The combustible gas control system, with the exception of the backup dilution air subsystem, shall be designated as Seismic Category I and Safety Class 2 (SC-2) and designed in accordance with ASME III, Class 2. The dilution air subsystem is nonnuclear safety (INS) and nonseismic.
- 3. The system shall be protected f rt the effects of tornadoes and external missiles.
- 4. The design of the combustible gas control system shall 16 comply with Regulatory Guide 1.7 as stated in Section
, 3A.1-1.7.
6.2.5.2 System Design Two redundant 100 percent capacity hydrogen recombiner subsystems are provided to maintain the hydrogen concentration below 4 volume percent following a DBA. The hydrogen recombiner subsystems recombine hydrogen in the containment atmosphere with oxygen to form water. A dilution air subsystem is provided as a backup to the reccxnbiners. The dilution air subsystem discharges to the radioactive gaseous waste system (Section 11.3) . The type of recombiner (catalytic or electric) is described in Section 6.2.5 of the U t ility-Applicant's SAR. A general description of the combustible gas control system operation is given below. The hydrogen recombiners are housed in the annulus building which provides missile and tornado protection. The controls and power supply are locally installed near each recombiner in a shielded control area. The hydrogen recombiners are skid-mounted and can be moved in and out of the annulus building it required for maintenance or replacement. On a multiple plant site, only two hydrogen recombiner units are provided for the site. The hycirogen recombiner units are installed in accessible locations and can be moved to any one of the plants following a LOCA. A hydrogen analyzer is permanently installed and connected to the piping downstream trom the containment isolation valves in each recombiner suction line and is used to sample the containment atmosphere and measure its hydrogen concentration. Each analyzer is equipped with its own compressor and gas dryer. f, 7 - n7 UUJ s vL 6.2-56 Amendment 16 8/29/75
SWESSAR-P1 Following a DBA the combustible gas control system is isolated by the containment isolation system. At an appropriate time following the accident, the isolation valves can be remote manually opened and the hydrogen analyzers used to monitor the hydrogen concentration in the containment. The hydrogen recombiners can then be manually initiated to reduce this concentration. In the unlikely event that both recombiners f ail, the diluth,n air subsystem is initiated to reduce the hydrogen concentration in the containment. During operation of the dilution air subsystem, the contdinment atmosphere is drawn from a recombiner suction line. The gas then flows from the dilution air exhaust blower to the radioactive gaseous waste disposal system. This gas is then filtered to remove iodine and particulate matter before release to the ventilation vent. Dilution air is supplied to the containment by the ailution air supply blower through a recombiner discharge line. 6.2.5.3 Design Evaluation To ensure operability of the hydroge recombine r subsystem in the event of a single failure of any component, the system is arranged in two redundant 100 percent capacity trains. Each train is powered by a separate emergency bus. The analysis of hydrogen generation following a LOCA and the capability of the combustible gas control system, in both the recombiner and dilution modes, to maintain a hydrogen concentration below 4 volume percent are presented below. 22 Assumptions and Methods Used in the Analysis i The five major sources or hydrogen generation rollowing a DBA assumed for the analysis are:
- 1. Zirconium - water reaction The reaction of the zirconium clad with water is assumed to occur instantuneously after a DBA and the hydrogen evolved is added to the containment atmosphere immediately.
9 ( ' bb) ~
./ qd )T 6.2-57 Amendment 22 3/17/76
SWESSAR-P1 The amount Table 6.2.5-2. of zirconium assumed to react is listed in l
- 2. Pressurizer gas space and reactor coolant system water The hydrogen present in the pressurizer gas space and reactor coolant is assumed to be released to the containment atmospisere instantaneously af ter a DBA. The volume of hydrogen released from these sources is given in Table 6.2.5-2.
- 3. Radiolytic generation in the water collected on the con-tainment floor Radiolytic decomposition of water is given by the following chemical equation:
H O + Energy H +1/2Of The energy for this reaction is supplied by the decay of fission products which have escaped from the reactor core and dissolved in the sump water. The ficcion product distribution and energy absorption used in the analysis are in conformance with Regulatory Guide 1.7.
- 4. Radiolytic generation in the reactor core The chemical reaction presented in assumption (3) is also applicable to radiolytic decomposition of water in the rea: tor core. In this case the energy is supplied by the decay of fission products in the fuel.
The distribution of fission products and energy assumed in the analysis is given in Table 6.2.5-2.
- 5. Corrosion Corrosion of aluminum and zinc hearing materials, such as galvanized steel, within the containment may react with the containment spray and produce hydrogen.
The rates at which aluminum and zine react to produce hydrogen are given in Table 6.2.5-2. The zinc corrosion rate is considered to be temperature-dependent for the first day following the accident and is calculated from the following equation: 35 Zn Corrosion Rate = A exp (- BJ (mg/dm z/hr) T i,7 buJ
'"\h J'
6.2-58 Amendment 35 10/6/77
SWESSAR-P1 where: T is containment temperature, OR A and B are determined from Draft ANS Standard ANSI, N-275, " Containment Hydrogen Control," Working Group 56.1, Draft 4, June 1976 as tollows: Temp (F) A B 220 5 T 7.31 x 10:e 26,305.96 35 150 $ T <220 1.6 3 x 10 8 ' 26,849.78 The containment temperature transient is shown in Fig. 6.2.2-2 for the first day following the accident. After the first day, the zinc corrosion rate is assumed constant at 0.127 mg/dm2/hr (0.612 mils /yr) . The total amount of hydrogen in the containment is calculated by summing the hydrogen produced by each of the above sources. The partial- pressure of the hydrogen is calculated using the perfect gas law. The total pressure in the containment is then calculated by summing the partial pressures of the noncondensables. The volume percent of the hydrogen is calculated from the ratio of the partial pressure of the hydrogen to the total containment pressure. The partial pressure of the water vapor in the containment is neglected for conservatism. The containment atmosphere circulates through the recombiner at a constant flow rate (see Table 6.2. 5-1) . A separate calculation is made for both one and two recombiners operating. Each recombiner combines hydrogen and oxygen to form water vapor, The hydrogen free gas and the water vapor are returned to the containment. The hydrogen in the containment is assumed to consist of hydrogen produced by the sources listed above minus the hydrogen previously recombined. The total containment pressure is adjusted to account for the amount of oxygen and hydrogen removed by the recombiner. The volume percent of hydrogen for each case is then calculated from the ratio of the hydrogen partial pressure to the containment pressure (neglecting the partial pressure of water vapor) . The to tal hydrogen generated for each source is shown in Fig. 6.2.5-4 as a tunction or time. Corrodible Material inside Containment The maximum inventory of materials inside the containment that may contribute to the production of hydrogen after a LOCA is given in Table 6.2.5-2. The inventories of zirconium and aluminum are listed in Table 6.2.5-4 for each NSSS. The balance of plant specification of materials inside containment does not include zirconium or
-n-6.2-58A bUJ Jd) Amendment 35 10/6/77
SWESSAR-P1 aluminum. The NSSS Vendors
- SAR should Fe consulted for a listing of equipuent that uses these materials.
O 2r ,n l' u. 7J _g6 6.2-58B Amendment 35 10/6/77
SWESSAR-P1 Hydrogen Recomb _iner Performance Evaluation Fig. 6.2.5-2 gives the hydrogen concentration inside the contain-ment following a DBA as a function of time with no recombiners operating, with one recombiner operating (minimum engineered safety features), and vith two recombiners operating. One recombiner in operation is sufficient to maintain the hydrogen concentration in the containment a tmosphere below 4 volume percent. Although it is not necessary to operate the hydrogen recombiner until the Lydrogen concentration in the containment approaches 4 voluma percent, this figure shows the capacity of the system if the recombiners are actuated 15 days after the DBA. The hydrogen concentration at this time is approximately 2.8 volume percent. Dilution Air Subsystem Performance The dilution air flow rate required to keep the hydrogen concentration in the containment below 4 volume percent with no recombiners operating is shown in Fig. 6.2:5-3. Radiological Consequence of Containment Dilution RESAR-41 NSSS As shown in Fig. 6.2.5-3, containment dilution is not required until 42 days after the DBA, thus there is no radiation dose to the public as a result of purging the containment within the first 30 days. After the first 30 days, assuming the recombiners are not functioning, the radioactive releases from the containment purge following a loss-of-coolant accident result in a 30 day LPZ dose of less than 1 percent of the thyroid dose and less than 2 percent of the whole body dose calculated for any given distance in Section 15.1.13. These results are based on the assumptions given in Table 6.2.5-3. RESAR-3S As shown in Fig. 6.2.5-3, containrent dilution is not required until 35 days after the DBA, thus there is no radiation dose to the public as a result of purging the containment within the first 30 days. w 29 After the first 30 days, assuming the recombiners are not tunctioning, the radioactive releases from the containment purge following a lv.is-of-coolant accident results in a 30 day LPZ dose of less than 3 percent of the thyroid dose and less than 2 percent of the whole body dose calculated for any given distance in Section 15.1.13. These results are based on the assumptions given in Table 6.2.5-3. 6.2-59 /<] ~n7 Amendment 29 10/29/76
SWESSAR-P1 29 Limiting Case for C-E and PSW NSSS The additional dose due to purging that must be added to the 30 day LOCA dose is less than 2 percent of the thyroid dose and less than 3 percent of the whole body gamma dose calculated f or any given distance in Section 15.1.1?. The 30 day purging dose i.e., from 24 days (purge initiation) to 54 days is less than 11 percent of the thyroid dose and less than 15 percent of the whole body gamma dose calculated for any given distance in Section 15.1.13. These results are based on the assumptions given in Table 6.2.5-3. Hydrogen Mixing in Containment With two containment atmosphere recirculation coolers in operation, as is the case with minimum ESF oper ting, the air recirculation flow is approximutely 134,000 cfm. With this flow, the containment atmosphere is conpletely cycled through the coolers every 25 min. The containment atmosphere recirculation system operating in conjunction with the containment spray system ensures good containment utmosphere mixing and thus prevents hydrogen stratification in the centainment atmosphere. 6.2.5.4 Testing and Inspections A description of the tests and inspection performed on the combustible gas control system is given below. The frequency of these tests is discussed in Section 16.4.5. Recombiners The hydroien recombiners are qualified to meet or exceed the design parmneters listed in Table 6.2.5-1. The qualification tests performed in the selected recombiner are described in the Utility-Applicant's SAR. A performance test of each electric hydrogen recombiner will be made before the recombiner is accepted for shipment. This test is accomplished by placing the unit in operation, starting the hydrogen recombiner, and allowing atmospheric air to flow through th e unit. A flow of at least 50 ccfm is maintained and checked by a flowmeter. A satisfactory temperature rise through the recombiner indicates proper operation. For catalytic recombiners and electric recombiners not previously licensed, the recombiner is tested to ensure that the unit is capable of recombining hydrogen by supplying a hydrogen-air ntixture to the recombiner intake at a hydrogen concentration of 4 volume percent. An analysis is then performed on the discharged Lydrog en-a ir nd xture to determine its hydrogen (_ _ v 'J 6.7-60 Amendment 29 10/29/76
SWESSAR-P1 concentration. An effluent concentration of less than 0.04 volume percent indicates that the recombiner is operable. Periodic testing of electric recombiners is accomplished by placing each unit into normal operation and by noting the air flow rate and recombiner temperature. An air flow rate of at least 50 scfm and a recombiner temperature specified by the supplier are acceptable. Periodic testing of catalytic recombiners will require a hydrogen-air feed of 4 volume percent hydrogen to the recombiner. The discharge of the recombiner is analyzed for concentration of hydrogen. A hydrogen concentration of less than 0.04 volume percent shows that the recombiner is operable. Hydrogen Analyzers Ea ch analyzer, its sample system, and mounting cabinet will be subjected to a seismic vibration test and subsequently observed for physical integrity and operability during normal plant operation. Periodically, a gas sample cf a Knokm hydrogen concentration is supplied to the analyzer. The indicated concentration is then compared with the known value. Valves Safety Class 2 valvec require testing as specified in Section 16.4.5. 6.2.5.5 Instrumentatiqn Applications The combustible gas control system is operated manually and monitored locally. The ins truments and controls are locally mounted on a control panel in the annulus building. Redundant hydrogen analyzers are provided to monitor the hydrogen concentration of the containment atmosphere following a LOCA. Each analyzer includes a gas compressor to force the gas sample through the analyzer, and a refrigerant type gas drier to remove moisture trom the gas sample before it is analyzed. The analyzers are capable of measuring the hydrogen concentration to +0.1 volume percent. The instrumentation for each recombiner includes: Temperature indicator Flow indicator Pressure indicator { j)] )O) 6.2-60A Amendment 29 10/29/76
SWESSAR-P1
'nie dilution air subsystem instrumenta tion consists of flow indicators in the dilution air supply line and in the dilution exhaust line.
6.2.S.6 Materials Ma terials used in the recombiners are selected to resist radiolytic or pyrolytic decomposition; therefore, no interference with the sate operation of this or any other engineered safety feature will occur due to decomposition products. Tne dilution air subsystem is constructed of steel pipe with a steel blower. This material will generate neither radiolytic nor pyrolytic aecomposition products that will interfere with the safe operation of this or any other engineered safety feature. 6.2.5.7 Interrace kequirements Mie following information will be supplied in the Utility-Applicant's SAR.
- 1. Specification or the type of recombiner (catalytic or electric)
- 2. A description of the qualification tests performed on the selected recombiner.
- 3. The recombiner specifications will be in accordance with Table 6.2.5-1.
6.2.6 Containment Leakage Monitoring System Tie containment leakage monitoring system is primarily used for preoperational and periodic Type A (integrated) leakage rate testing of the cont 4unment structure and may be used for nonitoring the containment struc-ture leakage rate duriraj plant operation. Fig. 6.2.6-1 shows the system necessary to perform 34 the reference volume and absolute method testing. bej 0 $ 6.2-60B Amendment 34 7/22/77
SWESSAR-P1 6.2.6.1 Design Basis The design bases for the containment leakage monitoring system are:
- 1. The containment leakage monitoring system shall provide the capability for detection and measurement of a containment leakage rate of less than 0.1 percent of the containment structure free volume in 24 hr with a 95 percent confidence level.
- 2. The containment leakage monitoring system is classified nonnuclear safety (NNS) except for the containment penetrations and containment isolation valves which are Satety Class 2.
6.2.6.2 System Design Containment structure leakage rates (Type A) are measured using either of two test methods: the reference volume and/or 34 absolute. The makeup air method is used for instrument verification. The Utility-Applicant will select the test method to be used. The entire system shown on Fig. 6.2.6--1 is necessary to perform the reference volume method tes t . If only the absolute method 9 test is to be used, the sealed pressure bulbs, ditferential nanometer, containment leakage monitoring vacuum pump, associated valves, and piping are not necessary, and are not included in the design. The reference valume method is an application of the perf ect gas law. The containment structure leakage rate over a 24 hr test period is measured by the change in differential pressure between the containment atnosphere and a leaktight reference volume. The reference volume consists of sealed pressure bulbs located throughout the interior of the containment structure, providing a means for minimizing the effect on the test results of average temperature changes in the containment atmosphere during the test. Temperature changes nave a nearly equal effect on both the referc_nce volume and the containment a t:nosphere, and the reference volume pressure is af fected by containment temperature only. Therefore, temperature measurement is not critical to the accuracy of the leakage rate determined by this method. However temperature is a small f actor in the reference volume method equation and must be measured. Prior to testing, the bulbs are evacuated by the containment leakage monitoring vacuum pump, then tilled with instrument air. The use of dry air simplified calculations by eliminating the necessity to measure and correct for humidity changes in the reference volume. A correction for containment atnosphere humidity changes is still required. These data are suppMed to the computer system for a computerized @ solution of the ref er t.nce volume method equation. Manual 6.2-61 d7 ~' 1 Amendment 34 [J v J s i 7/22/77
SWESSAR-P1 calculations for leakage rate, using either the absolute or reference voluwe method, can be performed using appropriate data. We absolute method of leakage rate determination is a direct application of the perfect gas law. Containment atmosphere pressure and temperature measurements are taken hourly throughout a 24 hr test period, and the data are fitted by the method ot linear regression (i.e., linear least squares) to a linear "Iuation relating time to the air mass insida the containment s tructure. The slope of the line representing this equa tion is the containment structure leakage rate. The mean containment atmosphere temperature is obtained by averaging the readings or the resistance temperature detectors (RTDs) located throughout the interior of the containment structure. An absolute manometer measures the containment pressure. Ris manometer samples pressure from two open-ended pipes, in order to provide redundancy. Ilumidity sensors provide data for use in correcting leakage rates for changes in containment humidity. The above data are supplied to the computer system which calculates the absolute leakage rate of the containment structure. The makeup air method is used to verify the accuracy of the Type A test inetrumentation used with the above method. At the end of a Type A test, a mass of air is added to the containment through a ilowmeter. The mass of air added is between 50 and 100 percent of the mass corresponding to a leakage of La for one day. The value of the mass added, as measured by the Type A test instrumen ta tion, is then compared with the value determined by the flowmeter to verify the Type A test instrumentation accuracy. In addition to those ir.struments described above, instru: rents are provided to record atmospheric conditions external to the containment structure during testing. Such factors as barometric pressure, windspeed, and site temperature may af f ect the results of the tests. They are monitored to help analyze any inconsistencies and variations in the containment leakage rate measurements. The containment leakage rate can be determined during normal plant operation by establishing pressure at either plus or minus 0.5 to 1.0 psig. The air mass is determined from the indicated 34 containment pressure, temperature, humidity differential pressure, as a function of time. 6.2.6.3 Desi gn Evaluation The absolute and dit f erential manometers in the containment leakage monitoring system are expected to be sensitive to pressure changes of 10.001 and 10.00015 psi, respectively. The temperature detectors are expected to be sensitive to changes of 10.1 F, and the dewpoint sensors to dewpoint changes of 10.5 F. The actual sensitivities will be determined when the instruments have been selected. ,, , . , (i b ) ;ii 6.2-62 Amendment 34 7/22/77
SWESSAR-P1 Combining the above sensitivities with the inaccuracy involved in converting the electrical signals to numerical values in the computer yields an estirtate of the accuracy of a single calculation of air mass inside the containment structure. This analysis and previous nuclear power plant experience indicate that the containment structure leak rate can be measured to 10.009 percent per day with 95 percent confidenca using the 34 absolute or the ref erence volume method. , Nuclear plant tencing experience at the Surry Power Station (1) has proven that the minimum censitivity for the ref erence volume or absolute method is well within the design basis accident leakage rate of 0.2 percent of containment structure free volume in 24 hr. In addition, both methods have been found to be reliable in a Type A testing program. The makeup air method is used only to verity the accuracy of the instrumentation used in the other two types of tests. 6.2.6.4 Testing and Inspections Conui nment structure leakage is tested prior to initial opera-tion and periodically throughout the operating lif e of the plant. Testing includes performance of Type A tests to measure the con-tainment structure overall integrated leakage rate, Type B tests to detect and measure local leakage from certain containment structure components, and Type C tests to measure containment isolation valve leakage rates. The contairunent structure leakage tests are performed in accordance with the requirements of Appendix J to 10CFR50(2), as noted in Section 16.4.4. Acceptance criteria for the leakage tests are specified in Section 16.4.4. Type A Tests Preoperational and periodic Type A tes ts are performed at calculated peak containment internal pressure (Pa) . All equipment within the containment structure is designed to withstand periodic testing at P g without affecting operational capabilities or operating life. Instruments used to measure the containment structure leakage rate are calibrated before each Type A test as described in Section 7.7. Before the preoperational Type A tests, the sealed pressure bulb system, if used for the reference volume method, is filled with an air-Freon gas mixture and tested for leakage with a halogen detector. Before each Type A test, the sealed bulb system, if used, is pressurized to 48 psig and an absoluce method leak test is performed to determine the leak rate using an 34 absolute manometer for pressure measurement and the RTDs for temperature measurement. The Type A test program is described in Section 16.4.4. 6.2-63 I Amendment 34 7/22/77
SWESSAR-P1 Type B Tests Type B tests are performed on the following components:
- 1. Inboard and outboard persorinel access lock doors and seals
- 2. Equipment hatch seals
- 3. Fuel transfer tube blind flange seals Fluid system penetration design does not include resilient seals, gackets, sealant components, or expaasion bellows and, therefore, does not require Type B testing (Paragraph II.G.1 of Appendix J) .
In lieu of other Type B testing, each electrical penetration is provided with a leakage surveillance system which continuously pressurizes the penetration test chamber to a pressure not less than P (Paragraphs III .B.1 (c) and IIIB.3 (b) of Appendix J). Each component subject to Type B testing is equipped with a double O-ring seal and a test connection between seals. A pressure equal to Pa is applied to the test chamber with a halogen-air gas mixture, and leakage is checked with halogen detection equipment. Repair and retest are required in the event of any detectable leakage. Therefore, the sensitivity of the test equipment is used as a quantitative value for component
- leakage, Tyrie C Tests A list of all contain7ent isolation valves is given in Tables 6.2. b1 and 6.2.4-2.
Type C tests are performed on the containment isolation valves specified in Table 6.2.%3 as follows: Each valve to be tested is closed by normal operation without any preliminary exercise or adjustment. A section of piping which includes the containment isolation vaives is isolated from the remainder of the 21uld system, using valves or blanking flanges as necessary, and the piping is drained (if applicable) . The inside and outside containment isolation valves are tested individually with air at a pressure equal to P a. Test air is applied at a test connection on the inboard side (toward the inside of the containment structure) of the valve to be tested, and leakage air is vented through a test vent en the outboard side of the valve. A rotameter, connected to the pressure source, is used to measure leakage through the containment isolation valve as a function of time . In this procedure, air flow across ' he valve seat is m always in the inside--to-outside containment sg_ ure. =s.5-direction. (() b 6.2-64 Amendment 34 7/22/77
l,.-l .. $ N.- ..
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a SWESSAR-P1 The containment isolation valves, which isolate the containment 9 leakage monitoring system structure, lines penetrating the containment require testing as specified in Sections 16.4.2 and i 3 16.4.4. E Type Ce Tests Type C' tests arti performed as follows. A section of piping which includes the containment isolation valves is isolated from the remainder of the fluid system, _ using valves or blanking flanges as necessary. The configuration and other conditions of this isolated section of piping are the same as for the Type A test. ; Tne inside and outside containment isolation valves are i tested sianultaneously (both valves closed) with air at a pressure equal to P . Tes t air is applied at a test connection on the inboard side (toward the inside of the containment structure) of the inside isolation valve and - leakage air is vented through a test vent on the outboard - side of the outside valve. A rotameter, connected to the pressure source or the vent, is used to m isure leakage through the containment isolation valve as a function of time. In this procedure, air flow _ across the valve seat is always in the inside-to-outside containment structure direction. i Isolation valves in lines which are vented and drained during a Type A test may receive a type C' test at the option of the Utility-Applicant. However, valves in lines which bypass the = an=,ulus building shall not receive a type C8 test. 6.2.6.5 Instrumentation Applications The containwnt leakage monitoring system includes the following . instrumentation for measurement of the containment structure ' leakage rate:
- 1. One differential manuneter to measure differential pressure between tne reference volume and the -
containment ae.mosphere for computer input and indication (not required for absolute method testing) .
- 2. One absolute manometer to measure containment atmosphere =
absolute pressure for computer input and indication.
- 3. RTDs to measure containment atmosnhere temperature for "
computer input. 34 -
- 4. RTDs for main control board indication of containment 9 atmosphere temperature (no computer input) .
=
i 6.2-65 " Amendment 34
,<, 7/22/77 a f u, ;'. s,s
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SWESSAR-P1
- 5. Ilumidity transmitters to measure dewpoint temperature within the containment atmosphere: one for indication in the control room and the remainder for computer input.
34
- 6. One barometer, one thermometer , and one windspeed ana direction indicator to monitor atmospheric conditions outside the containment structure during testing.
- 7. One air flowmeter to totalize air flow during the makeup air test, local indication only.
'Ib calculate containment atmosphere mean temperature, the RTD signals are weighted with fa ctors assigned to represent the proportion of the containment structure voSune in which each RTD is located.
6.2.6.6 Materials The containment leakage monitoring system uses materials that are consistent with their location and expected environment. The resistance temperature detectors, dewpoint transmitters and 34 sealed pressure bulbs if used are located inside the containmcnt and are designed to f unction properly over the 40 year life of the plant. The remaining components are located in the annulus building. The procurement specification for each component sti l .ulates the design environmental conditions of the particular comr onent. 6.2.6.7 Interface Requirements 34 The specific containment leak rate test method, either absolute or reference volume, will be provided in the Utility-Appli-cant's SAR. References for Section 6.2.6
- 1. Virginia Electric and Power Company, Surry Power Station.
Unit No. 1, " Report of Containment Lea? Rate Test Startup," Docket No. 50-280, August 11, 1972.
- 2. " Reactor Containment Leakage Testing for Pater Cooled Power Reactors," as published in the Federal Register, Volume 38, No. 30, February 14, 1973 and as corrected on March 6, 1973.
i( ; ,,- (10 J .; i U 6.2-66 Amendment 34 7/22/77
F i ! SNESSAR-P1 L t- ' TABLE 6.2.1-1 ACCIDENT CHRONOLOGY PUMP SUCTION DER ICRMAL ESF WITH FAILURE OF ONE (IMTADDOTP SPRAY PUMP Time, oce Event 0 Accident occurs. 8 CDA signal, redundant $7 valves from CAT open, spray pump isolation valves begin to open and spray pumps start. I 16.1 Accumulators start injes-ting into reactor vessel. 21.5 First pressure peak occurs. l} 24.5 Blowdown over. 25 Safety injecticn system starts. 30 Containment atmosphere recir-culation system starts. 72 Accumulators empty. 100 Injection phase of the contain- l7 ment spray system begins. 120 Spray additive delivery into {1 the containment begins. 159 Core reflooding over, second peak pressure occurs. 1,500 Recirculation phase is auto- l7 matically initiated. Indefinite Recirculation phase ends. l . l I-t h t [.
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SWESSAR-P1 TABIE 6.2.1-4 MASS AND ENERGY RELEASE TO TTE REACTOR CAVITY 150 IN.a PUMP DISCHARGE LDR ' 9 Tilne, maec Mass Rate, 103 lbm/sec Energy Pate, 106 Ptu/sec 0.0 0.0 0.0 3.01 13.91 7.88 5.00 15.63 8.E6 10.00 19.4i 11.03 13.01 24.59 13.97 16.00 23.52 13.34 20.01 25.09 14.25 24.03 24.11 13.67 40.05 26.18 14.65 53.06 27.04 15.34 60.07 25.82 14.64 73.09 25.07 14.20 9 83.12 25.30 14.34 91.03 24.84 14.07 110.07 23.60 13.36 155.15 24.81 14.06 180.07 23.70 13.42 200.08 23.98 13.58 280.05 24.04 13.62 360.01 23.70 13.43 460.12 23.83 13.50 590.00 24.00 13.59 i f 4 W 1 of 1 Amendment 9 i i 4/30/75 Y
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SWESSAR-P1 TABLE 6.2.1-5 MASS AND ENERGY RELEASE TO STEAM GENERATOR COMPARTYENT PUMP DISCHARGE SER 9 Time, maec Mass Rate, 1031trn/sec Energy Rate, 10*Ftu/sec 0 0.0 0.0 1 30.50 17.27 2 33.45 18.93 3 33.28 18.82 5 32.32 18.25 10 32.80 18.49 15 44.20 24.91 19 47.82 26.98 24 50.11 28.26 31 64.77 36.60 35 62.34 35.20 40 69.18 39.11 49 74.48 42.14 56 77.60 43.93 64 77.36 43.78 73 77.77 44.02 81 78.05 44.17 100 77.38 43.79 120 78.54 44.45 125 78.91 44.66 130 78.89 44.65 150 78.27 44.30 175 78.38 44.26 180 78.34 44.34 230 77.47 43.84 340 77.36 43.18 350 77.54 43.88 360 77.41 43.81 490 76.29 43.17 700 74.54 42.18 800 73.43 4 1.54 1000 71.91 40.69 1060 71.56 40.50 1100 71.54 40.50 1250 ~1.16 39.74 h 00 6 .89 37.93 1750 63.1% 35.92 5 2000 59.24 33.69 2250 55.99 31.89 2563.4 52.00 30.01 i' ,' i 3 l'y M 1 of 1 Amendment 9 { 4/30/75 hhb L-
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SWESSAR-P1 TABLE 6.2.1-6 i e MASS AND ENEPGY PELEASE % PRESSURIZER
, CUBICLE - SURGE LINE DER Time, Psee Mass Rate, 103 lbm/sec EnerqY Pate, 106 Btu /sec 0 0 0 1 18.89 12.97 2 18.91 12.98 3 18.87 12.95 4 18.84 12.93 7 18.82 12.91 11 18.95 13.00 12 22.86 15.61 16 24.32 16.59 17 24.82 16.93 18 24.80 16.91 19 24.01 16.38 20 22.93 15.65 28.1 22.61 15.43 37.1 22.65 15.46 45 22.74 15.52 52 22.82 15.57 60 22.89 15.62 67.1 22.88 15.62 80 22.27 15.20 90 21.62 14.77 7 110 21.42 14.63 115 21.63 14.7 8 120 21.86 14.9 2 125 21.95 14.99 130 21.85 14.9 2 155 20.11 13.75 180 19.29 13.20 2 10 19.23 13.15 320 18.73 12.82 440 18.63 12.75 550 18.56 12.70
, 650 18.50 12.6 6 800 18.40 12.58 1000 18.25 12.47 '. 1300 18.05 12.32 I 1700 17.76 12.11 2000 17.49 11.92 2500 17.11 *o.65 g . 3000 16.72 11.35 w 1 of 1 Amendment 7 2/28/75
, o I.
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SWESSAR-P1 TABLE 6.2.1-7 MASS AND ENERGY RELFME TO PRESSURIZER CUBICLE - SPRAY LINE DER Time, maec Mass Rate, 103 lbm/sec Energy Rate, 106 Dtu/sec 0 0 0 1 3.839 2.598 2 4.149 2.776 3 4.242 2.829 4 4.262 2.840 5 4.212 2.811 6 4.056 2.720 7 3.876 2.616 8 4.081 2.734 10 4.285 2.851 18.06 4.303 2.8S8 32 4.732 3.105 33.1 4.732 3.105 34.1 %.724 3.100 39 4.611 3.033 56 4.55' 2.997 1 68.05 4.664 3.062 69.2 4.665 3.063 70 4.665 3.062 85.1 4.615 3.033 100 4.478 2.953 105 4.489 2.959 125 4.741 3.106 160 4.56S 3.003 290 4.519 2.974 340 4.448 2.933 440 4.431 2.921 57cl 4.392 2.897 630 4.402 2.901 790 4.395 2.895 950 4.387 2.887 1200 4.368 2.873 1750 4.287 2.819 2000 4.242 2.790 2290 4.208 2.768 2370 4.208 2.767 2460 4.203 2.763 2750 4.188 2.752 2930 4.183 2.749 3000 4.179 2.746 t I t' [ w 1 of 1 Amend: rent 7 1 2/28/75 i V m: J' bbJ
.--y.-- ,
SWESSAR-P1 TABLE 6.2.1-0 CONTAIlmDrr PEAK PRESSURE AND TMPMATURE FDLIOtfING MAIN STEAM PIPE AND FEED %ATFP PIPE BREAR INSIDE LVNTAINME?rf Percent Full !*ercent rull Single Active Peak Pressure, Peak AcTident _Pwe r _ Break Area Failure neig Tergre g a t ure. F Feedwater 0 100 MSIv** 21.9 220.4 Pipe Break
\
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.sh L%
C' g V. ( i ['- ) c,
**M.nin steme isolation valve f ailure to close. g l11 M 1 of 1 Arenament 19 17 /1 ? / M
SWESSAR-P1 TABLE 6.2.1-9 MASS AND ENERGY RELEASES TO CONTAINMENT MAIN STEAM PIPE DER (!OTE 1) (STT.AM GENERA'IOR COMPARISON DATA) W C-E Steam generator secondary side inventory, linn (maximum) 210,000 >269,000 Steam generator =2H == diameter, in. 198 265 Steam generator secx>ndary olde height (tube sheet to outit r, nozzle), in. 688 662 Number of main steam lines per steam gu.arator 1 2 Steam generator nozzle minimum ID, in. 16 28 Main steam line ID, in. 30 30 to Main steam line cross sectional g,9 area, fte 4,9 Flow restrictor area, fta (steam generator site of break) 1.4 2x4.3=8.6* Assumed marinn (100 percent) Lreak area, its 4.9+1.4=6.3 2x4.3+2x4.9=18.4
- The C-E steam generators are not equipped with flow restrictors.
However, P.s two steam generator outlet nozzles are 28 '.n . ID which are emnected to main steam lines which are 30 in. ID. Note 1: These data will be provided when available t .xn the NSSS vendor. W 1 of 1 Amendment 10 5/15/75 LOS Li
SWESSAR-F1 TAB 12 6.2.1 10 PWS5 AIO DEPGY RFJ2ASES 10 CONTATIOTNT FEIDtIATD FIFE DER Steam Generator Main Feedwater Feedwater Pipe Main Steam B lotr20wrt Backflow Ttztal Floes Energy Mass Energy Mata Energy Mass Energy
- Mass knergy Pass pate Date Fate Rate Rate Rate Rate Rate Date Bate 108 10*
108 108 10* 108 10*
.- 108 10* 10*
< Ilme. Sec }Jw/$fc Stu/$eg pan /Sec Pt u/Sg Mur /Sec Ptta/Sec hh Ft g ee Uaq/see Ptu/sec 10.a55 0.0 0.0 0.0 0.0 20.91 10.357 5.97 a.3er
(
~
0.0 10.eS5 . 10.455 0.0 0.0 0.0 0.0 20.91 10.357 10.0 10.455 5.97 a.307 s E 10.455 4.387 0.0 0.0 20.91 10.357 10.1 10.455 5.97 0.0 0.0 4 10.eSS 4.387 0.0 0.0 20.91 10.357 11.6 10.e55 5.97 5.97 0.0 0.0 0.0 0.0 10.455 5.97 11.7 10.455 0.0 0.0 10.455 5.97 19.2 10.455 S.97
*-1 3.075 3.638 3.075 3.630 C- ;" 19.3 0.0 0.0 3.075 3.6 38 3.075 3.638
;; f 31.5 E .,p-j <* 31.6 0.0 0.0 0.0 0.0 E ,- /
[. = *AS Integrated 200,736 114.62sx 104 SSO 43.87x 15.6s0 6.5Fr 37,515 se.3ser 354,rso 209.46r 106 Ot u g-w# g,t Ammant s IJun 106 Ptu IJa 10* Pt u L2mm 106 Peu usa 10* Pt u Unn A-CN y-4 q;I p .) C~^ 1 of 1 A w nshment 9 u 4/30/75
SWESSAR-P1 TABLE 6.2.1-1 ACCIDENT CHRONOLOGY PUMT SUCTION DER NORMAL ESF WITH FAILURE OF ONE COMPONENT COOLING WATER PUMP Time, see Event 0 Accident occurs. 8 CDA signal, redundant valves from CAT open, spray punp isclation valves begin to open and spray pumps start. 16.1 Accumulators start injecting into reactor vessel. 20.5 First pressure peak occurs. 24.0 Blowdown over. 25 Safety injectican system starts. 17 30 Containment atmosphere recir-culation system starts. 60 Accumulators empty. 100 Injection phase of the contain-ment spray system begins. 120 Spray additive delivery into the containment begins. 151 Core reflooding over, second peak pressure occurs. 900 RHR pump recirculation phase. 1,500 Charging pump recirculation phase manually initiated. 4,650 Containment spray recirculation phase manually initiated. Indefinite Recirculation phase ends. W-3S 1 of 1 Amendment 17 9/30/75
, ;n Q L. ) J'I
. ShT.SSAh-P1 s g l, . .
pMO ' M TABLE 6.2.1-2 ghs $1$\ .R . \ - ENTRGY PAIATET TAALE g, , DOUBLE-ENDED PIMP SUCTION BRIAK TTO .g 3) ' NORMAL ESP WITH FAILURE OF ONE CisGM WENT OtlOLYNG !sA1TR PUMP Mk Q h ' . 3' i MILLIONS OF BTU Nornhs1 ESF Except Minisass _bnt( ainment Sprays MiniJetus ESF Reat Sources Time = 0.0 sec 24.4 see 150.27 see 210.0 sec Reactor coolant (inc1mieg pressur i zer) 302.29 17.34 29.74 35.00 RCS hot metal Thin metal 25.02 20.51 11.72 11.72 1 hick metal 30.00 30.80 25.36 23.45 Steam generator 412.08 410.92 344.60 332.09 Core stored heat 29.36 12.10 5.00 5.00 Ref ueling water storage tank 315.59 315.59 303.73 293.51 , Accumulator 18.51 13.79 0.00 0.0 8 Beat Sinks containment atmosphere water 2.01 240.68 273.39 258.64 Conteirusent atmisghere air 2.98 8.74 9.06 8.92 Cbntainment floor water 0.0 47.91 76.93 106.28 Concrete sinks 0.0 5. 0 '. 22.68 28.74 Contalrunent liner 0.0 22.51 60.87 67.47 Miscellaneous steel 0.0 0.0 0.0 0.0 Reactor vessel support shield tank water 19.42 19.46 19.67 19.76 Heat inputs Decay heat 0.0 9.86 31.70 39.40 Heat Outputs Residual heat ress> val heat exchangers 0.0 0.0 0.0 0.0 cont airunent recirculation alr coolers 0.0 0.0 5.07 7.54 C' Reat lost to atm>spMre 0.0 0.0 0.0 0.0 Reat f rrue steam generator meanndary 0.0 4.09 4.09 2.17 k"1 L I CD W-33 1 of 1 Amendment ~20 1/23/76
?!
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SWESSAR-P1 TABLE 6.2.1-4 MASS AND ENERGY RELEASE E THE REAC"IOR CAVITY 150 IN.a PUMP DISCHARGE LDR Ti2ne , msee Mass Rate, 108 1hn/sce Energy Rate, 106 btu /sec 0 0 0 2.51 12.48 7.00 12.53 22.52 12.63 15.02 22.35 12.52 20.05 25.46 14.26 25.01 24.63 13.77 45.01 27.39 15.30 70.03 25.40 14.16 77.50 25.94 14.47 U 105.14 22.83 12.70 110.03 22.97 12.78 125.16 23.99 13.36 170.15 22.14 12.32 212.51 23.10 12.86 237.53 22.67 12.61 250.08 22.92 12.76 275.09 22.34 12.43 312.65 22.91 12.75 325.07 22.70 12.63 350.22 22.76 12.67 387.50 22.66 12.61 400.29 22.71 12.64 412.57 22.70 12.63 437.62 22.86 12.72 462.56 22.71 12.64 475.05 22.74 12.65 500.02 22.89 12.74 550.03 22.75 12.66 575.14 22.87 12.73 W-3S 1 of 1 Amendment 17 9/30/75 i ? 7"7 bb) JJJ
SWESSAR-P1 TABLE 6.2.1-4 A MASS AND ENERGY RELEASES TO THE REACTOR CAVITY 144 IN? PCrr LEG LDR Time fmnec) Mass Rate, 10311w/see Energy Rate, 10* Btu /sec
- 0. O. O.
- 1. 10. 12 6.58
- 3. 10.08 6.55
- 5. 10.08 6.54
- 7. 13.89 9.03 10, 15.67 10.19
- 12. 15.83 10.29
- 14. 17.91 11.65
- 16. 19.07 12.41
- 18. 18.15 11.82
- 19. 17.47 11.35
- 20. 16.70 10.88
- 25. 15.97 10.36
- 30. 17.24 11.19
- 35. 16.47 10.67 20
- 40. 16.34 10.60
- 45. 16.65 10.80
- 50. 16.24 10.63
- 55. 16.11 10.44
- 60. 16.09 10.43
- 65. 15.90 10.30
- 70. 15.69 10.17
- 75. 15.42 9.99
- 80. 15.23 9.87
- 85. 15.01 9.72
- 90. 14.64 9.48
- 95. 14.30 9.?6 100. 14.02 9.08 135. 12.71 8.22 150. 11.76 7.61 170. 10.99 7.11 200. 10.36 6.71 250. 10.05 6.50 300. 9.99 6.46 350. 10.04 6.50 400. 10.01 6.48 450. 9.99 6.46 500. 10.01 6.48 W-3S 1 of 1 Amendment 20 1/23/76
, , ' ~ \
(') V J
's ') J M
SWE5SAR-P1 TALLE 6.2.1-5 MASS AND ENERGY RELIASE TO STEAM GENERATOR COMPARTME15' PUMP DISCHARGE SEP Time, msg Mass Rate. 10 ltr/see Enerrry h te. 1060tu/sec 0 0 0 2.5 33.25 18.63 5 32.47 18.15 10 31.94 17.78 15 33.13 18.4 17.5 43.8 24.38
- 20. 47.75 26.55 22.5 54.56 30.32 27.5 58.15 32.3 35 62.7 ~ 4.85 37.5 64.3 35.76
- 45. 71.9 40.0
- 50. 74.94 41.71 57.5 76.29 42.46 4
- 65. 75.24 41.85 72.6 73.99 41.15 85 75.77 42.15 92.5 75.53 42.0 102.6 75.23 41.82 120 73.31 40.77 137.5 75.91 42.23 150.4 75.06 41.75 157.6 75.0 41.72 175. 75.44 41.92 185. 74.77 41.58 195. 73.65 40.95 205.4 72.96 40.56 227.5 73.83 41.06 237.5 74.18 41.26 250. 73.9 41.11 270. 74.14 41.24 302.6 72.4 40.26 365. 74.25 41.3 407.6 72.84 40.51 445.2 73.18 40.7 500. 72.67 40.41 540. 72.9 40.54 630. 72.1 40.1 850. 69.84 38.83 1500. 61.2 34.08 K-3S 1 of 1 Amendment 18 10/30/75 h f a
SWESSAR-P1 TABLE 5.2.1-6 wh S AND ENERFr RELEASE W PRESSURIZER RELIEF TANK COMPARD 1NT SURGE LINE DER Time. Mass rate, Energy Rate, i r::S ee) 1031hn/nec 10
- fitu/sec 0 0 0 2.5 15.16 10.27 25 18.83 12.65 50 19.15 12.85 75 18.65 12.52 100 18.26 12.33 125 18.25 12.25 150 16.94 11.39 9 175 16.23 10.94 200 15.59 10.52 225 15.55 10.44 250 15.29 10.32 275 14.78 9.98 300 14.37 9.72 325 14.30 9.67 350 14.27 9.65 375 14.24 9.64 425 14.21 'J . 61 450 14.21 9.61 475 14.19 9.58 500 14.17 9.58 650 14.11 9 . ra 800 14.06 9.50 1000 13.95 9.42 1 of 1 Amendment 17 W-3S 9/30/75 bu) s JU
SWESSAR-P1 TABLE 6.2.1-7 MASS AND ENERGY RELEASE TO PRESSURIZER CUBICLE SPRAY LINE DER Time Mass Rate Energy Rate n/see 10 lb/see 106 htu/seg 0 0 0 5 5.23 3.20 25 5.54 3.36 50 5.68 3.43 75 5.79 3.48 100 5.70 3.43 125 5.73 3.45 150 5.60 3.38 175 5.48 3.30 200 5.50 3.32 225 5.52 3.33 g 250 5.47 3.30 275 5.53 3.33 300 5,36 3.25 325 5.33 3.22 350 5.36 3.24 375 5.3 3.20 400 5.37 3.24. 425 5.34 3.22 450 5.27 3.18 475 5.27 3.18 500 5.24 3.16 770 5.18 3.13 1000 5.1 3.11 1200 5.13 3.1 1400 5.1 3.07 1600 5.06 3.06 1300 5.05 3.04 2000 5.03 3.04 H-3S 1 of 1 Amendment 17 9/30/75
*) O NOI
SWESSAP-P1 TABLE 6.2.1-8 CONTAINMENT PEAK PRESSURE AND TEMPLPA1UPE F0110etDiG MA3N STIAM PIPL AND FE2.DWATER PIPE BREAK INSIDE COtFTAINTENT Time of Time of Time Fan Time Peak Single Pres- FWTV FWI Signai MSIV SLI Coolers Spr.sys Percent Break Start Full Area (Fta) Active sure,E*3 Peak Isolation Generated Iso *ation ed Generat- Start Wmp,(#3 (Sect by (Sect Py JSec) [Sec) A nident Power Iypets) _ Failuge Esta Peedwater 0 2.33 DER MSIV8*3 22.8 222.8 Pipe Break
' 11 357.9 10.0 MSSS 10.0 MSSS 3'.6
. 106.0 Main Steam 102 1.4 DEReal MSiv 33.5 12.0 NSSS ' 37.2 319.9 Pipe Break 102 0.6 LERt P 3 MSIV 30.3 123.4 12.0 NSSS 239.0 SPLIT MSIV 30.8 336.5 17.5 CPtB83 35.2 CP 40.0 102 0.86 10.0 NSSS 10.0 NSSS 31.6 106.0 104 1.4 DER SPRAYt*3 14.1 157.9 10.0 NSSS 31.6 106.0 COf f
- 3 34.0 362.2 10.0 N.iSS 102t** 1.4 DER 319.9 12.0 NSSS 12.0 MSSS 38.2 429.0 70 0.5 LDR MS1V 31.5 219.0 16.4 CP 33.6 CP 39.0 70 0.903 SPLIT MSIV 30.3 330.5 13.5 NSSS 13.5 NSSS 40.0 679.0 30 0.4 IfR MSIV 33.0 112.7 38.2 229.0 332.8 15.7 CP 31.6 CP 30 0.942 SPLIT MSTV 29.9 13.5 NSSS 13.5 MSSS 40.0 179.0 30473 0.4 IDH SPhAY 38.5 321.1 NSSS 10.0 NSSS 31.4 101.0 0 1.4 DER MSIV 31.8 353.1 10.0 1454.0 IDF MSIV 26.2 298.9 14.5 NSSS 210 CP 75.5 0 0.1 CP 75.5 CP 49.2 779.0 0 0.4 SPLIT MSIV 26.7 3G9.9 27.3 101.0 DER SPRAY 34.3 353.1 13.0 NSSS 10.0 NSSS 31.4 0 1.4 NOTES:
(1) For DLPs the forward flow area is 1.4 f t e end the reverse f low area is 4.2 f t8, corresponding to the 27.7 in. ID main steam line. For LDRs the value listed is the flow area to each side of break. For splits, the value listed is the total flow area. (2) Fur the DERs the blowdown data were ocunputed with the MARVEL code arh1 transmitted by letter PDA-851-SMSP-056,
- ' ' dated 5/3/76 f rom Westinghouse to Stone & Webst er .
,~31 ( 3) For LDRs and splita the blow &wn data are deternis - ' rom Westinghouw St andard 12.2, Rev. 1. (4, Main steam isolation valve f ailure to close.
-'d (5) Failure of one containment spray pump.
4 (6) Lim' ting steam line break case f or maximum containment atm.ospt pre terverature. ( (7) Limiting steam line break case f or maximum containment stru> sphere pressure. Amendment 29 W-35 1 of 2 10/29/76
91 96 27
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SarESSAJt-P1 irAnIJE 6.2.1-tA (Ctern It"P?ttrD tHIp st tw13x;tnp stryn PgrIMLJ}rV7FT9mTe s s WW JWTACT UNIIS TUTAL Time Mass Rate Energy Rate Masa Rate Energy Rate Mass Rate Energy Rate Mass Rate Ehergy mate Sf(M 4s _JLaljLe,s_ 10* Btu W /sec 106 Pt a/peq _ ItW eec 10* Btu /sec linn/see 10* btu /sec 190.0 200.0 250.0 ft 300.0 1200.0 1945.10 1945.19 10000.0 (h L.. -4 t -4 J= N' teDTE : s a n joe ge m of main f eedwater line inveentory e3ded to initial ruptured loop steam geerator inventory. w_33 2 of 2 Natwnt 28 9/b/76
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SWESSAR-P1 TABLE 6.2.1-1 ACCIDENT CHRONOLOGY 5.585 FTa PCMP SUCTION IDR MINIMIM F.SF f34 Time, see Rvent 0 Accident occurs 9 CDA signal. Spray ptznp isolation valves begin to open and spray pump starts. 23 Core flooding ' is start in3ecting into reactor ve.,sel. 23 First pressure pek occurs. 30 Contain:nent atmosphere recirculation system starts. 3r 31 Blowdown over. Safety injection system starts. 81 Core flooding tanks empty. 100 Injection phase of the containment spray systems begins. 120 Spray additive delivery into the containment begins. 171 Core reflooding over. 3,420 Recirculation phase initiated. Indefinite Recirculation phase ends. B&W 1 of 1 Amendment 34 7/22/77
. ~/
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EaGSSAR-F1 TABLE 6.2.1-2 ENERC) BALJuK3 TAALE (8 3 5.585 FT* PtMP SUCTION LDR - MINIMist ESF MILIJONS OF BTb I.i_me s o sec 31 acc 170 sec 3360 pec Heat Sources Reitsella; water storage tank 295.33 292.71 295.33 158.36 Beat Sinks Countainment atmospinere unter 2.30 276.19 301.04 %4.99 la Contain=t atmusphere air 3.21 9.57 9.81 8.31 Qmtainment floor water 0.0 54.65 83.30 617.98 Qecrete sinks 0.0 5.70 24.04 1M .61 Containment liner 0.0 23.55 58.00 77.24 heat Inputs dCS blowdown 0.0 364.24 475.60 848.54 ETS spillage 0.0 0.0 U.0 149.24 Pump and tan heat 0.0 0.00 0.06 1.96 Beat Outgasta Contain-t air coolers 0.0 0.04 6.61 136.85 m a -Y. - k-1' y, sa8xass P.neryy satance in asss ver.aora s sure Ammtamt 14 B&W 1 of 1 7/27/77
StfESSAlt-P1 TABLE 6.2.1-3 SUBGMPAJtTMENT DESIGN HLESSURE DIFFDtENTIAIE f
- 3 Congnater Peak Calculated Destgri Pressure Subocumoartaset Brent Omsidered Code Pressure Dif f erential. paid Differential. paid Results
- 1. Pressuriser cahicle Pressure profile surge line DEREal THREED mode 1 6.3 8.8 38 pode 2 6.3 8.8
- 2. Surge line com- surge line DET THREED Fig. 6.2.1-6 Partsanat Uniform guessure pode 3 11.3 15.8
- 3. Reactsr coolant surge line DER 883 THREED drain tanat evuepa rtmen t uniform pressure Mode 5 10.6 14.0 18
- 4. steam generator Bot leg DERF 83 THREED Fig. 6.2.1-8 wrtaset beloir operating floor Pressure profile leode 10 46.5 65.1 Inde 2 21.0 29.4 mode il 14.8 20.7 mode 12 14.7 20.6 3g Mode 14 46.5 65.1 C Bode 13 Mode 15 21.0 14.8 29.4 20.7 t-.
Inode 16 14.7 20.6 L'
- 5. steam generator Bot leg DERt3D THREED Fig. 6.2.1-4 shield wall above operatlag floor Uniform pressure a Q neude 5 28.0 39.2 lts D 6. wa=.e belo. tr. not leg DERc=> TimEED steen generator support pedestal B< 1 of 2 Amendment 4 12/21/11
8 5 3 l 1 l n 9 a 1 e Mm ta t 2 l l u 6 i f .f h nn e s e . o roe wo t i s2 fe1 R g gtie n i vr w e m F ed u d A rens unas ets sgle si pr a es p l ei d ree pdhn v a ra up tg e rs i ss, ains s l ie i el ul d ra cl e n g Pi iel m t w gu i t 1 7099159735 7 nn ge 6 0 r a d ai nm se 5531s24064 5 5 pl e yar d ir 3 2723e2 4 5 se 1111 7 1 ef sinm e Df iha e t>r i D hsettoh p p
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SWESSAR-P1 TABLE 6.2.1-4 MASS & ENERGY RELEASE TO ' HIE REACTOR CAVITY 2.5 Fta PIMP DISCHARGE LDR Mass Rate Energy Rate Time, sec (IIrn/sec) (Btu /sec) O. .0 .0
.105000E-02 .319082E+05 .180959E+08 l33 .203000E-02 .253163E+05 .143271E+08 .301000E-02 .335536E+05 .190287E+08 .406000E-02 .364837E+05 .207130E+08 .504000E-02 .360289E+05 .204510E+08 .602000s-02 .368739E+05 .209315E+08 .707000E-02 .432890E+05 .245935E+08 .805000E-02 .458883E+05 .260843E+08 .903000E-02 . 4 82661E+ 05 .2 74 4 97E+ 0 8 .100100E-01 .493678E+05 .280835E+08 .110o00E-01 .4 8 7731E+ 05 .2 77412E+ 0 8 .120400E-01 .475349E+05 .270303E+08 .130200E-01 .466381E+05 .2L5176E+08 .140700E-01 .461326E+05 .262305E+08 .150500E-01 .461037E+05 .262170E+08 .'60300E-01 .467512E+05 .265905c+08 .e70100E-01 .477316E+05 .271544E+08 .180600E-01 .481889E+05 .274179c+08 .190400E-01 .477089E+05 .271433E+08 .200200E-01 .469116E+05 .266862E+08 .210700E-01 . 4639 9'/E+ 05 263917E+08 .220500E-01 .46176PE+05 .262621E+08 230300E-01 .460266E+05 .261751E+08 .240100E-01 .459597E+05 .261339E+08 .250600E-01 .459612E+05 .261332E+08 .260400E-01 .458573E+05 .260726E+08 .270200E-01 .456176E+05 .259346E+08 .280700E-01 454662E+05 .258475E+08 .290500E-01 .456295E+05 .259410E+08 .300300E-01 .460503E+05 .261825s+08 .310100E-01 .465560E+05 .264736E+08 .320600E-01 .470674E+05 .267673E+08 .330400E-01 .475672E+05 .270553E+08 .340200E-01 .480346E+05 .273249E+08 .350*/00E-01 .4 8 4751E+ 05 .275792E+08 .360500E-01 .488803E+05 .278168E+08 .370300E-01 .492821E+ 05 .280455E+08 .380100E-01 496310E+05 .282472L*08 .390600E-01 .498434E+05 .283702E+08 .400400E-01 .498686E+05 .283852E+08 B&W 1 of 3 Ameninent 39 7/14/78 (i '
O (! b ) J 'l U
SWESSAR-P1 nELE 6.2.1-4 (corr) Mass Rate Energy Rate Time, sec flbm/sec) (Btu /FEL f36
.410200E-01 .497520E+05 .293183E+08 .420700E-01 .495343E+05 .281930E+08 .430500E-01 .492429E+05 .281752Eto3 .440300E-01 .489128E+05 .27835.as08 .450100E-01 .485611E+05 .276326E+06 .460600E-01 .481856E+05 .2 7416 4E+ 08 .470400E-01 .478008E+05 .271950E+08 .480200E-01 .474376E+05 .269861E+08 .490700E-01 .471091E+05 ,267972E+08 .500500E-01 .468623E+05 .266555E+08 .600600E-01 .469253105 .266968E+08 .700700E-01 .469208E+05 .266944E+08 .800100E-01 .468304E+G5 .266456E+08 .900200E-01 .448274E,05 .254962E+08 .100030E+00 .447239E+05 .254389E+08 .110040E+00 .452244E+05 .257280E+08 .120050E+00 .454538E+05 .258617E+b2 .130060E+00 .463901E+05 .264014E+08 .140070E+00 .468558E+05 .266710E+08 .150010E+00 .465662E+05 .265058E+08 .160020E+00 .460479E+05 .262069E+08 .170030E+00 .458019E+05 .260858E+08 18C040E+00 .454387E+05 .259573E,08 .190050E+00 .452781E+05 .257643E+08 .200080E+00 .460015E+05 .261817E+08 .225050E+00 .462736E+05 .263379E+08 .250040E+00 . 4 61690E+ 05 .262762E+08 .275030E+00 .458557E+05 .260932E+08 .300020E+00 .460403E+05 .262020E+08 .325010E+00 .461834E+05 .262769E+08 .350070E+00 . 4 59365'4 + 05 .261299E+08 .375060E+00 .459433E+05 .261299E+08
.400050E+00 .459043E+05 .261021E+08 .450030E+00 .457451F+05 .260019E+08 .500010E+00 .458993E+05 .260787E+08 550060E'00 .460520E+05 .261533E+08 .600040E')0 .1618050+05 .262133E+08 .650020E+00 .461634E*05 .261892E+08 .700070E+00 .463708E+J5 .262944E408 .750050E+00 .464265E+05 .263131E+08 .800030E+00 .465386E+05 263648E+08 .850010E+00 .465951E+05 .263859E+08 .900060E+00 .467049E+05 .264391E+08 .950040E+00 .466936E+05 .264241E+08 BSW 2 of 3 Amendment 36 12/21/77
, n l l , .I h
SWESSAR-P1 TABLE 6.2.1-4 (COtm Mass Rate Energy Rate Tim , sec (1hD/sec) (Btu /Sec) 36
.100002E+01 .472681E+05 .2 67452 E+ 0 8
.110005E+01 .467010E+05 .264164E+08
.120001E+01 .486435E+05 .263827E+08
.130004E+01 .466251E+05 .2637792+08
.140007E+01 .466128E+05 263830E+08
.150003E+01 .465451E+05 .263619E+08
.160006E+01 .464524E+05 .263303E+08
.170002E+01 .462639E+05 .262469E+08
.180006E+01 .460994E+05 .261797E+08
.190001E+01 .459306E+05 .261115E+08
.200004E+01 .457619E+05 .26C447E+08 Source: B-SAR-205 Table 6.2-18A, A=nd w t 18, 9/26/77 BSW 3 of 3 Ameniment 36 12/21/77
/s 47 ', ~ n UU) _aU
SNESSAR-P1 TABLE 6.2.1-4A MASS S ENERGY RELEASE TO THE REAC' TOR CAVITY 2.5 FTr HOP LM IDR Time, see Mass Rate Energy Rate J1bn/sec) fBtu/sec) l36
- 0. O.
.105000E-02 O.
.366202E+05 .239192E+08
.203000E-02 .318048E+05
.001000E-02 .207521E+08
.299120E+05 .195086E+08
.406000E-02 .305279E+05
.504000E-02 199133E+08
.317045E+05 .206866E+08
.60200 0E-02 .309551E+05
.707000E-02 .201938E+08
.296313E*05 *
. 93245E+08
.805000E-02 .294560E+05
.903000E-02 .192097E+08
.307459E+05 .200571E+08
.100100E-01 .325417E+05
.110600E-01 .212389E+08
.339266E+05 .221521E+08
.12 040 0E-01 .343152E+05
.130200E-01 .224087E+08
.338790E+05 .221208E+08
.14 0700E-01 .331295E+05
.150500E-01 .216265E*08
.325159E+05 .212223E+08
.16 0300E-01 .320994E+05
.170100E-01 .209482E+08
.317297E+05 .207049E+08
.18 0600E-01 .312926E+05
.190400E-01 .204176E+08
.308400E+05 .201205E+08
.200200E-01 .305225E+05
.210 70 0E-01 .199113E+08
.304247E+05 .198472E+08
.220500E-01 .305518E+05
.230300E-01 .199308E+08
.307946E+05 .200905E+08
.240100E-01 .310432E+05
.250600E-01 .202540E+08
.312295E+05 .203766E+08
.260400E-01 .313186E+05
.270200E-01 .204353E*08
.313165E+05 .204340E+08
.280700E-01 .312362E+05
.290500E-01 .203813'+08
.310868E+05 .202831E+08
.300300E-01 .308822E+05
.310100E-01 .201500E+08
.306397E+05 .199893E+08
.320600E-01 .303683E*05
.330 40 0E-01 .198111E+08
.301100E+05 196415E+08
.340200E-01 .298991E+05
.350700E-01 .195031E+08
.297366E+05 .193965E+08
.36050 0E-01 .296336E+05
.370300E-01 193290E+08
.295899E+05 .193001E*C8
.380100E-01 .29593FF %
.390600E-01 193029E+0B
.29631T '5 ,
.193275E+08
. 400 400E-01 .29688.s+05
.410200E-01 .193652E+08
.297503E+05 .194057E+08
. 4 20700E-01 .298083E+05 .194439E*08 B&W 1 of 3 Amerriment 36 12/21/77
(, L j JJI
SWESSAR-P1 TABLE 6.2.1-4A (CONT) Mass Rate Energy Rnte Time, sec flbs/sec) _ IBtu/sec) 36
. 4 30500E-01 .29999fE+05 .195693E+08
. 44 0 30 0E-01 .3077793+05 .200799E+08
. 45 0100E-01 .315121E+05 .205620E+ 0 8
.460600E-01 .321501E+05 .209812E+08
.470400E-01 .326600E+05 .213165E+08
. 4 8020 0E-01 .330278E+05 .215584E+08
. 49 0700E-01 .332917E+05 .217321E+08
.500500E-01 .334557E+05 .218 401E+ 0 8
.600600E-01 .326702E+05 .213236E+08
.700700E-01 .296248E+05 .193239E+08
.800100E-01 .310306E+05 .196567E+08
.900200E-01 .316452E+05 .206520E+08
.100030E+00 .304420E+05 .198622E+08
.110040E+00 .286963E+05 .187178E+08
.120050E+00 .285246E+05 .186065E+08
.130060E+00 .284224E+05
.140070E+00 .185407E+08
.28?F09E+05 .184363E+08
.150010E+00 .281393E+05
.160020E+00 .183582E+08
.281083E+05 .183394E+08
.170030E+00 .281090E*05
.180040E+00 .183415E+08
.280898E+05 .183306E+08
.190050E+00 .280492E+05 .183058E+08
.200060E+00 .290140E+05 .182847E+08
.225050E+00 .279626E+05
.250040E+0i .182546E+08
. 2 7759 8E+ 0 5 .181272E+08
.275030E+00 .276090E+05
.300020E+00 .180343E+08
.274868E+05 .179603E+08
.325010E+00 .273207E+05
.350070E+00 .178577E+08
.2724 4 2E
- 05 .198130E+08
.375060E+00 .271015E+05
.400050E+00 .177267E+08
.269959E+05 .176637E+08
.450030E+00 .267139E+05
.500010E+00 .'74877E+08
.283684E+05 .172722E+08
.550060E+DO .262170E+05
.600040E+00 .171833E+08
.261010E+05 .171166E+08
.650020E+00 .259855E+05
.700070E,00 .170489E+08
.258819E+05 .169878E+0B
.750050E+00 .257997E+05
.800030E+00 .169393E+08
.257054E+05 .168818E+08
.850010E+00 .255802E+05
.900060E+00 168033E+08
.254948E*05 .167507E+08
.950040E+00 .254536E+15
.100002E+01 167268E+08
.253982E+05 .166936E+08
.110005E+01 .252620E+05
.110001E+01 166089E+08
.251040E+05 .165113E+08
.130004E,01 .249870E+05 164402E+08 B5W 2 of 3 A:mendn.ent 36 12/21/77 rn
;y b
SWESSAR-P1 TABLE 6.2.1-4A (CONT) Mass Rate EnertJy Rate Time, ser (lbrn/sec) (Btu /sec) l35
.140007E+01 .247640E+05 .162984E+08
.150003E+01 .245639E+05 .161696E+08
.'.50000E+01 .244271E+05 .160806E+08
.170002E+01 .242306E+05 .159502E+08
.180005E+01 .24085GE+05 .158518E+08
.190001E+01 .239750E+05 .15774 0E+ 0 8
.200004E+01 .239124E+05 .157264E+08 Source: B-SAR-205 Table 6.2-18A, Am=Mm=_t 18, 9/26/77 B&W 3 of 3 Amendment 36 12/21/77
, -r ;
} .)
SWESSAR-P1 TABLE 6.2.1-5 MASS AND ENERGY RELEASE TO STEAM GENERATOR COMPARTMENT HOT LEG DER Time. Sec Mass Rate fibm/sec) Energy Rate (B tu/sec) 0.0 -
.00105 .191794E+06 .125116E+09
.00105 - .00203 .174382E+06
.00203 - .00301 .113684E+09
.00304 .170807E+06 .111345E+09
.00406 .167289E+06
.00406 - .00504 .109043E+09
.163801E+06
.00504 - .00602 .160574E+06
.106761E+ 0 9
.00602 - .00707 .158420E+06
.104650E+09
.00707 - .00805 .103238E+09
.156538E+06 .102004E+09
.00805 - .00903 .154723E+06
.00903 - .01001 .152940E+06
.100815E+09
.01001 - .01106 .151802E+06
.996475E+08
.01106 - .01204 .151266E+06
.989028E+08
.01204 - .01302 .150820E+06
.985515E+08
.01302 - .01407 .150403E+06 .982591E+08
.01407 - .01505 .150017E+06
.979855E+08
.01505 - .01603 .149683E+06
.977323E+08
.01603 - .01701 .149384E+06
.975129E+08 35
.01701 - .01806 .149108E+06 .973169E+08
.01806 - .01904 .148863E+06 .971362E+08
.01904 - .02002 .148651E+06 .969754E+08
.02002 - .02107 .148453E+06
.968366E+08
.02107 - .02005 .14 8 273E + 0 6
.967070E+08
.02005 - .02303 .148110E +06
.965885E+08
.02303 - .02401 .147956E+06 .964819E+08
.02401 - .02506 .147797E+06 .963803E+08
.02506 - .02604 .147607E+06 .962757E+08
.02604 - .02702 .147521E+06 .961505E+08
.02702 - .02807 .147988E+06 .960932E+08
.02807 - .02905 .148443E+06 .963962E+08
.02905 - .03003 .148787E+06 .966907E+08
.03003 - .03101 .149037E+06 .969124E+08
.03101 - .03206 .149197E+06 .970728E+08
.03206 - .03304 .149260E+06 .971743E+08
.03304 - .03402 .149230E+06 .972121E+08
.03402 - .03507 .149105E+06 . 9 71883E+ 08
.03507 - .03605 .148890E+06 .971028E+08
.03605 - .03703 .148597E+06 .969581E+08
.03703 - .03801 .148225E+06 .967630E+08
.03801 - .03906 .147758E+06 .965157E+08
.03906 - .04004 .147214E+06 .962063E+08
.04004 -
.04102 .958458E+08
.146615E+06
.04102 - .04207 .145925E+06 .954506E+08
.04207 - .04305 .145167E+06 .949950E+08
.04305 - .04403 .144377E+06 .944956E+0B
.939750E+08 B&W 1 of 3 Amendment 35 10/6/77 t,p: IEk-as
SWESSAR-P1 TABLE 6.2.1-5 (CONT) Time, Sec Mass Rate (lbm/sec) Enercy Rate (Btu /sec)
.04403 - .04501 .143539E+06 .934230E+08
.04501 - .04606 .142617E+06 .928167E+08
.04606 - .04704 .141647E+06 .921784E+08
.04704 - .04802 .140668E+06
.04802 - .04907 .915349E+08
.139617E+06 .908446E+08
.04907 - .05005 .138537E+06 .901351E+08
.05005 - .06006 .134063E+06
.06006 - .07007 .871937E+08
.128670E+06 .836477E+08
.07007 - .08001 .125122E+06
.08001 - .09002 .813385E+08
.123538E+06 .803373E+08
.09002 - .10C03 .123043E+06
.10003 - .11004 .800594E+08
.123050E+06 .801178E+08
.11004 - .12005 .123657E+06
.12005 - .13006 .805710E+08
.125057E+06 .815370E+08
.13006 - .14007 .126969E+06
.14007 - .15001 .828251E+08 128752E+06 .840147E+08
.15001 - .16002 .130074E+06
.16002 - .17003 .848890E+08
.130945E+06 .854629E+08
.17003 - .18004 .131506E+06
.18004 - .19005 .858332E+08 g
.131810E+06 .860382E+08
.19005 - .20006 .131801E+06
.20006 - .22505 .860429E+08
.130972E+06 .855171E+0s
.22505 - .25004 .129349E+06
.25004 - .275C3 .844783E+08
.128216E+06 .837516E+08
.27503 - .30002 .127 92 7E+0 6
.30002 - .32501 .835648E+0F
.128098E+06 .836709E+
.32501 - .35007 .127946E+06
.35007 - .37506 .127107E+06
.835612E. s
.37506 - .40005 .830042E+08
.40005 - .125835E+06 .821755E+08 45003 .124874E+06
.45003 - .50001 .815584E+08
.126208E+06 .823963E+08
.50001 - .55006 .126330E+06
.55006 - .60004 .122172E+06 .823436E+08
.60004 - .65002 .794711E+08
.118304E+06
.65002 - .70007 .118901E+06
.767796E+08
.70007 - .75005 .769272E+08
.120050E+06 .774276E+08
.75005 - .80003 .118038E+06
.80003 - .85001 .759844E+08
.114027E+06
.85001 - .90006 .111543E+06
.733321E+0B
.90006 - .95004 .716451E+08
.110997E+06 .711840E+08
.95004 - 1.00002 .110045E+06 1.00002 - 1.10005 .705022E+08
.107326E+06 .687419E+08 1.10005 - 1.20001 .104835E+06 1.20001 - 1.30004 .672146E+08
.102310E+06 .657405E+08 1.30004 - 1.40007 .100151E+06 1.40007 - 1.50003 .645326E+08
.977129E+05 1.50003 - 1.60006 .958273E+05
. 6 314 4 0E+ 0 8
.620364E+08 B&W 2 of 3 Amendment 35 10/6/77
-r "
~
(: j JJJ
SNESSAR-P1 TABLE 6.2.1-5 (CONT) Time. Sec Mass Rate fibm/sec) Enerav Rate (Btu /see) 1.60006 - 1.70002 .941115E+05 .609 50E+08 1.70002 - 1.80005 .917355E+05 .595594E+08 1.80005 - 1.90001' .895732E+05 .582570E+08 1.9C001 - 2.00004 .868652E+05 .565933E+08 35 Source: B-SAR-205 Table 6.2-16, Amendment 15, 5/20/77 B&W 3 of 3 Amendment 35 10/6/77 bb. O
SWESSAR-P1 TABLE 6.2.1-6 MASS AND ENERGY RELEASE M Tu, PRESSURIZER CUBICLE SURGE LINE DER Stagnation Enthalpy Total Mass Total 35 Mass Rate Rate Energy Rate Btu /lb (10* lbm/see) (10* Itrn/sec) (106 Btu /see) 14" Sche-dule 160 Pressurizer 697.8 0.825 1.9 52 Side 13.124 35 Hot Leg uS3.9 1.127 Side Mass pipe after and energy rupture.rates are held constant for ttle first 2 seconds Source: 15 B-SAR-205 Table 6.2-32, Amendment 16, 6/28/77 B&W 1 of 1 Amendment 35 10/6/77
-r7 bb U#'
SWESSAR-P1 TABLE 6.2.1-7 MASS AfD ENERGY RELEASE TO THE PRESSURIZER CUBICLE - SPRAY LINE DER Stagnation Total Total Enthalpy Mass Rate, Mass Rate k35 Energy Rate Btu /lbn 10* 1hn/see 10* 1hn/sec 10* Btu /see Pressurizer Side 1120.4 0.033 Cold Leg Side 570.4 0.153 0.186 1.235 Mass and energy rates are held constant for the first 2 seconds after pipe rupture. Source: B-SAR-205 Table 6.2-32, Amendment 16, 6/28/77 B&W 1 of 1 Amendment 35 10/6/77 . m () U ) 's o
SWESSAR-P1 TABLE 6.2.1-8 CtWPAINHEWT PEAK PRESSURE AND TD4PERAMIRE PO1JDetING MAIM STEAM PIPE BMAK 1RSIDE CWTAIMMLVF Peak Time of Time of Time Fan Time Pw1V MSIV Coolers Break Pressure Peak Isolation Sprays Ar9a fFtal Psia Isolation Start Start Temp, F ESsc) (Sect M ).. ISect l11 8.55E8B 33.7 404.4 21.9 9.9 31.0 114.0 2.0tel 25.9 413.7 24.0 12.0 35.2 141.0 0.33 s> 24.7 361.0 42.5 30.5 56.0 NAI3) 39 NOTESI (1) Initial containment pressure = 15.7 gaia (2) Initial containment pressure = 14.7 psia (3) Since containment spray setpoint is 40 pela and the peak pressure reached is 24.7 peig, containment spray setpoint is not reached and sprays do not actuate. 31 , ~ ~ . C.' s. k-(, ' s.3 B&W 1 of 1 Amendment 39 7/14/78
SWESSAR-P1 TABLE 6.2.1-9 MASS AND ENERGY RELEASES TO LVNTAINMLWT MAIN STEAM PIPE PREAK 8.55 Pr2 DER, 102 PERCENT FOWER, MSIV 6 DG FAILURE Time After Rupture Mass Release Rate Energy helease Rate see Ib/see htu/sec 0.0-0.001 1.524+4 0.001-0.002 1.865+7 1.869+4 2.287+7 0.002-0.003 1.861+4 2.277+7 0.003-0.004 1.854+4 0.004-7.005 2.267 +7 1.840+4 2.257+7 0.005-0.006 1.b39+4 0.006-0.007 2.247+7
- 1. 8 31+ 4 2.238+7 0.007-0.008 1.824+4 2.228+7 0.008-0.009 1. 817 + 4 0.009-0.010 2.219+7 d.810+4 2.210+7 0.010-0.011 1.804+4 2.202+7 0.011-0.012 1.797+4 2. H 3 +7 0.012-0.013 1.791+4 2.184+7 0.013-0.014 1.789+4 2.176+7 "
0.014 4 .015 1.777+4 0.015-0.016 2.167+7 1.771+4 2.160 +7 0.016-0.017 1.766+4 0.017-0.018 2.152+7 1.760+4 2.144+1 0.018-J.019 1.754+4 2.137+7 0.019-0.020 1.748+4 0.020-0.021 2.129+7 1.743+4 2.122+7 0.021-0.022 1.738+4 0.022-0.023 2.115+7 1.7 2+4 2.108+7 0.023-0.024 1.727+4 0.024-0.025 2.102 +7 1.722+4 2.095+7 0.025-0.026 1.717+4 0.026-0.027 2.089+7 1.712+4 2.083+7 0.027-0.028 1.707+4 0.028-0.029 2.076+7 1.703+4 2.070+7 0.029-0.030 1.698+4 0.030-0.031 2.064+7 1.694+4 2.058+7 0.031-0.032 1.689+4 0.032-0.033 2.053+7 1.685+4 2.047+7 0.033-0.034 1.680+4 0.034-0.035 2.042 +7 1.676+4 2.036+7 0.035-0.036 1.672+4 0.036-0.037 2.031+7 1.668+4 2.025+7 0.037-0.038 1.664+4 0.038-0.039 2.020+7 1.660+4 2.015+7 0.039-0.040 1.656+4 0.040-0.041 2.010+7 1.652+4 2.005+7 0.041-0.042 1.648+4 0.042-0.043 1.999+7 1.644+4 1.995+7 0.043-0.044 1.641+4 1.390+7 B&W 1 of 4 Aht 37 5/22/78 bub U
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SWESSAR-P1 TABLE 6.2.1-9 (CONT) Time After Rupture Mass Release Rate Energy Release Rate Sec lb/sec Btu /sec 0.044-0.045 1.637+4 1.985+1 0.045-0.046 1.633+4 1.980+7 0.046-0.047 1.629+4 1.976+7 0.047-0.048 1.626+4 1.971+7 0.048-0.049 1.622+4 1.966+7 0.049-0.050 1.618+4 1.969+7 0.05-0.06 1.599+4 0.0s-0.07 1.93e+7 1.565+4 1.893+7 0.07-0.08 1.533+4 1.852+7 0.08-0.09 1.503+4 1.813+7 0.09-0.10 1.475+4 0.10-0.11 1.777+7 1.448+4 1.74 4 +7 0.11-0.12 1.424+4 1.712+7 0.12-0.93 1.402+4 1.685+7 0.13-0.14 1.382+4 0.14-0.15 1.662+7 1 364+4 1.640+7 0 .15-0.10 1.347+4 1.620+7 0.16-0.17 1.331+4 1. 6 01 + 7 0.17-0.18 1.315+4 1.580+7 0.18-0.19 1.298+4 1.560+7 U 0.19-0.20 1.282+4 0.20-0.225 1.539+7 1.258+4 1.509+7 0.225-0.250 1.233+4 1.477+7 0.250-O.275 1.214+4 1. 4 55 +7 0.275-0.300 1.199+4 0.300-0.325 1.438+7 1.189+4 1.425+7 0.325-0.350 1.183+4 0.350-0.375 1.415+7 1.177+4 1. 409 +7 0.375-0.400 1.174+4 1.404+7 0.400-0.450 1.171+4 0.450-0.500 1.u03+7 1.178+4 1.409+7 0.500-0.550 1.184+4 0.550-0.600 1.423+7 1.199+4 1.443+7 0.6UO-0.650 1.212+4 0.650-0.700 1.460+7 1.221+4 1.470+7 0.700-c.750 1.237+4 1.478+7 0.750-0.800 1.450+0 1.740+7 0.800-0.850 1.470++ 1.764+7 0.850-0.900 1.452<4 1.7 '42 +7 0.900-0.950 1 Q +4 1.722+7 0.950-1.00 1.419+4 1.703+7 1.0-1.1 1.626+4 1.1-1.2 1.951+7 1.625+4 1.950+7 1.2-1.3 1.519+4 1.3-1.4 1.823+7 1.521+4 1.825+7 1.4-1.5 1.527+4 1.5-1.6 1.832+7 1.503+4 1.803+7 1.6-1.7 1.482+4 1.778+7 B&W 2 Of 4 Arundnent 37 5/22/78 bb
SWESSAR-P1 TABLE 6.2.1-9 (CONT) Time Atter Rupture 'tss kelease Rate Sec lb/nec Energy Release Kate Btu /see 1.7-1.8 1.448+4 1.8-1.9 - 1.423+4 1.737+7 1.9-2.0 1.393+4 1.708+7 2 . 0 -2 . 2 1 210+4 1.672+7 2.2-2.4 1.154,4 1.452+7
- 2. 4 -2. 6 1.098+4 1.382+7 2.6-2.8 1.048+4 1.318+7 2.8-3.0 1.016+4 1.258+7 3.0-3.2 1. 221 +7 1.000+4 1.202+7 3.2-3.4 3.4-3.6 9.882+3 1.18 8 +7 3.6-3.8 9.759+3 1.173+7 9.643+3 3.8-4.0 9.543+3 1.159 +7 4 . 0 -4 . 2 9.45*+3 1.147+7 4.2-4.4 9.378+3 1.137+7 u.4-4.6 1.128+7 4.6-4.8 9.320+3 1.121+7 4.8-5.0 1.007+4 1.208+7 5.0-5.5 1.132+4 1.358+7 p
- 5. 5-6 . 0 9.485+3 1.138+7 6 . 0 -6. 5 8.761+3 1.067+7
- 6. 5 -7 . 0 8.424+3 1.03b+7 7.0-7.5 8.150+3 1.00 7+ 7 7.5-8.0 7.903+3 9.780+o 8.0-8.5 7.70b+3 9.5b4+6 7.4 95+ 3 8.5-3.0 7.299+3 9.313+6 9.0-9.s 7.164+3 9.078+6 9.5-10.ucan 7.006+3 8.914+6 10.0-11.0 6.287+3 8.718+6 11.0-12.0 4.711+3 7.818+e 12.0-13.0 3.Bbb+3 5.888+6 13.0-14.0 3.320+3 4.868+6 14.0-15.0 3.177+3 4.168+6 15.0-16.0 2.986+3 3.812+6 16.0-17.0 2.576+3 3.583+6 17.0-18.0 2.428+3 3.227+o 18.0-19.0 2.392+3 3.112+6 19.0-20.0 2.372+3 3.066+b 20.0-22.0 2.351+3 3.037+6 22.0-24.0 2.335+3 3.005+6 24.0-26.0 2.285+3 2.980+6 26.0-28.0 2.131+3 2.910+b 28.0-30.0 1.901+3 2.714+6 30.0-32.0 1.587+3 2.423+6 32.0-34.0 1.200+3 2.023+6 34.0-36.0 8.560+2 1.534+o 36.0-38.0 5.580+2 1.099+6 38.0-40.0 2.260+2 7.165+5 2.905+5 B&W 3 of 4 hs%t 37 5/22/78
? ,,
[t f
SNESSAR-P1 TABLE 6.2.1-9 (CONT) Time After Rupture Mass Release Rate Energv Relea ? Rate 37 See Ib/.sec btu /sec 40.0-42.5 0.031+2 0.046+5 ta3 Leak from unaffected SG isolated at 9.9 seconds by closure of main steam isolation valves. 37 B&W 4 of 4 Amendment 31 S/22AB
. n n -
SWESSAk-P1 TABLE 6.2.1-9A MASS AND 12tERGY M f 5 MPR 'IO CO1TIAlhMLdT
)AIN STEAM PIPE BREAK, 2.0 FTa MSLB, 102 PEhCurr POWER, MSIV & DG FAILURE Total Time Mass kate Fee Lnergy hate lb:n/see 106 BTU /sec 0.00-1.0 3,670. 4.498 1.00-2.0 3,718. 4.582 2.00-3.0 3,732. 4.612 3.00-4.0 3,740. 4.630 4.00-5.0 3,82u. 4.7b0 g 5.00--6.0 3,921. 4.889 6.00-7.0 3,984. 4.973 7.00-8.0 4,034. 5.04b 8.00-9.0 4,111. 5.151 9.00-10.0 4,104. 5.219 10.0-11 4 4,174.
11.0- a, 5.235 4,183. 5.zu8 12 . .a.0 4,187. 5.255 13 .4 -14.0 3,943. 4.926 14.0-15.0 3,649. 4.556 15.0-20.0 3,344. 4.206 20.0-25.0 3,020. 3.836 25.0-30.0 2,73 0. 3.500 30.0-35.0 2,440. 3.14 35.0-40.0 2,040. 2.62 40.0-45.0 1,040. 45.0-50.0 1.30 380. 0.440 50.0-55.0 40.0 0.040 55.0- 0 0.0 86W 1 of 1 Amendment 37 5/22/78
. < t, o f b
SWESSAR-P1 TABLE 6.2.1-9B MASS AND ENERGY RELEASES TO CONTAINMENT, MAIN STEAM P1PE BREAX, 0.33 FTa MSLB, 102 PERCDir POWER, MSIV & DG FAILURE Time Mass Rate Energy Rate (Sec1 (Lin/Sec) (105 Btu /Sec) 0 . 0 - 1. 0 679.5 8.319 1.0-2.0 687.5 8.511 2.0-3.0 672.0 9.040 3 . 0 -4 . 0 719.0 8.980 4.0-5.0 708.0 8.847 5.0-10.0 711.4 8.740 10.0-15.0 714.8 8.948 15.0-20.0 714.0 8.880 20.0-25.0 712.0 8.920 25.0-30.0 716.0 8.920 3 0.u-3 5.0 710.0 8.900 35.0-40.0 712.0 8.880 g 40.0-45.0 738.0 9.220 45.0-50.0 846.0 10.50 57.0-60.0 859.0 10.54 60.0-70.0 822.0 9.99c 70.0-80.0 745.0 9.080 80.0-90.0 685.0 8.360 90.0-100.0 637.0 7.822 100.0-110.0 601.0 7.340 110.0-120.0 563.0 6.910 120.0-130.0 530.0 6.500 130.0-140.0 501.2 6.153 140.0-150.0 474.7 5.813 150.0-160.0 450.0 5.510 160.0-170.0 428.0 5.240 170.0-180.0 408.9 5.005 180.0-190.0 396.0 4.847 190.0-390.0 396.0 4.847 390.0- 0.0 0.0 (1) Af fected steam generator bicarn completely dry B&W 1 of 1 Amendment 39 7/14/78 t (_, . o-
SWESSAR-P1 TABLE 6.2.1-1 ACCIDENT CHRONOLOGY PUMP SUCTION DER MINIMUM ESF Time, see Event 0 Accident occurs. 8 CDA signal, redundant valves frcxn CAT open, spray pump isolation valves begin to open and spray pumps st. art. 15.2 Safety injection tanks start to flood reactor vessel. 20.3 Fira.t pressure peak occurs. 24.0 Bloudown over. 30 High pressure safety injection system starts. 30 Containment atmosphere recir-culation system starts. 50 Low pressure safety injection system starts. 72.5 Safety injection tanks t..+ty. 9 100 Injection phase of the contain-ment spray system begins. 120 Spray additive delivery into the containment begins. 210.2 Cbre reflooding over 284 Second peak preseure occurs. 3,479 Cbntainment spray and HPSI re-circulation phase are autmta-atically initiated. LPSI system stops automatically. C-E 1 of 2 Amendment 15 8/8/75 60' 000
SWESSAR-P1 TABLE 6.2.1-1 (CONT) T.me, see Event 7,200 L>SI recirculation phase man- ;g ually initiated (cooldown phase). Indefinite Recirculation phase ends. C-E 2 of 2 Amendment 10 5/15/75
<,r)
- bQ- VU/
SWESSAR-P1 TABLE 6.2.1-2 ENERGY BALANCE TABLE DOtTELF-FNDI D PUMP SLK7 ION lip f.AK FILLIONS OF DTU Norinal ESF F.xcept rinimum Heat Sources Conta1.nmey _tSpra_ys_ _ Pinirsn_EEF Time = 0.0 sec 24.0 sec 210.2 sec 429.0 sec kcactor (hlant (includes pressurizer) 349.69 24.22 47.62* *
- RCS bot met al Ia>p metal 94.12 92.71 86.14* **
Vessel walls & int er nals 113.72 109.92 89.91* *
- Steam generator metal 200.31 195.95 177 12* *
- Steam generator inventory 209.86 184.78 106.81* 90.74 Core stored heat 31.82 13.09 6.52* *
- Refueling water storage tank #13.n1 413.01 4 19.59 Saf ety in jection t ank wat er 393.03 42.33 32.95 0.0 0.0 g Heat Sinks Containment atmosphere wat er 2.25 282.00 329.20 Containment atmosphere air 138.18 3.12 9.27 9.b9 9.77 lbntainment floor water 0.0 58.40 127.42 Concrete sinks 173.18 0.0 2.86 21.47 38.32 Containment liner 0.0 17.44 57.31 Misec114neous steel 72.89 0.0 0.0 0.0 0.0 Neutron shield tank water 22.53 22.58 22.97 23.45 Heat Inputa Decay and fission heat 0.0 Pump heat 7.27 39.92 54.14 0.0 0.0 0.04 0.12 Heat Outputs Shutdown heat exctanger s 0.0 0.0 0.0 0.0 C Conta irunent recirrelation air coolers 0.0 0.0 8.05 18.60 Heat lost to ata eher e 0.0 0.0 0.0 0.0
(:~ Flow to turbine 0.0 2".67 31.67 31.67 J. =
- Minimum ESF data not available, mar inn un ESF data at end of ent r ainment masumed t o al pl y.
** Data not narrently availat le in CESSAF.
C3 ( .) CD C -E 1 of 1 Amendw nt 10 1/21/76
3 0 4 5 6 8 1 66 1 1 1 1 1 27
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)
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SWESSAR-P1 TABLE 6.2.1-3 (CONT) Ctnuput er Peak Calculated Design Pressure Subc<mrpart ment Break Considere! Code Mesure Dif f erential paid Dif f erent ial,Jaid, Pressure grofile" mMes 152 85.4 119.6 n> des 3,6,9, 6 12 46.7 65.4 Nodes 4,5,10, 6 11 36.4 51.0 Nr>de s 7 6 8 138.4 193.8
- 6. Pipe penetrations (guard pipes) Note 3 2,250 hote 1. Break occurs in hade 5 of rig. 6.2.1-3.
- 2. Break occurs in Node 3 of Fig. 6.2.1-2.
- 3. Analysis is not performed in the guard pine. The piges are designed to wit hstand the max === kCS pressure.
- 4. The asymmetric Iwding is used in the structural design of the' reactor cavit y . it's particular pressure profile results f x(se a laeak surrounded II by Ntxx : ,2,7, and 8. liowever, th- cvyncrete shield wall is designed to withstand the asynsmetric loading resulting f run this pressure profile rot.ated to any angle in the plan view.
N. wh-, C'.2
=
- C-F 2 of 2 Amenenent 20 C' 8/6/76
SWESSAR-P1 TABLE 6.2.1-4 MASS AND ENEFGY PELEASE TO THE REACER CAVI'IY* I. Hot Leg SES Time, msec Mass Rate, 103 lbm/sec Energy Rate, 106 Itu/ser 0 0.0 0.0 1 15.63 10.1 5 58.28 37.5 l20 10 113.5 73.08 15 109 .9 70.80 20 107.4 69.15 30 104.1 67.00 40 102.3 65.79 50 101.8 65.45 75 104 .1 66.94 100 106.5 68.51 150 101.7 65.41 200 97.30 62.62 250 93.13 59.95 300 89.32 57.52 350 84.83 54.63 400 82.52 53.13 450 80.84 52.04 500 79.94 51.46 l20 II. 350 in.: Pump Discharge LDR 0 0.0 0.0 1 10.59 5.889 5 26.64 14.70 5.2 26.99 14.89 10 28.91 15.95 15 28.65 15.81 20 45.12 24.98 30 48.67 26.97 40 51.81 28.73 50 51.87 28.76 75 49.28 27.30 100 44.65 24.69 150 46.74 25.85 200 45.36 25.08 256 46.70 25.83 30 45.45 25.12
- Releases are calculated by the Combustion Engineering CEFLASH-4A computer code.
C-E 1 of 2 Amendment 20 1/23/76
, r,< *
() 'T A
SWESSAR-P1 TABLE 6.2.1-4 (CONT) Time, msec Mass Rate, los Ibn/sec Energy Eate, 106 Btu /sec 350 45.94 25.40 ig 400 45.59 25.21 c50 45.42 25.11 500 45.18 24.98 C-E 2 of 2 Amendment 19 12/12/75
, ..T r' ' - )
.I IL
SWESSAR-P1 TABLE 6.2.1-5 MASS AND ENERGY REIEVE
% STEAM GENERA'IOR COMPARTMEhT HOT LEG SES l20 Time , esec W as Pete, 1031bm/see EnercTy Rate, 10*Ptu/sec 0 0 0 1 15.63 10.10 5 58.28 37.55 10 113.5 73.08 15 109.9 70.80 20 107.4 69.15 3G 104.1 67.00 40 102.3 65.79 50 101.8 65.45 75 104.7 66.94 100 106.5 68.51 150 101.6 65.41 200 97.30 62.62 250 93.13 59.95 300 69.32 57.52 350 84.63 54.63 403 82.52 53.13 450 80.84 52.04 500 79.94 51.46 550 79.09 50.90 C-E 1 of 1 Amendment 20 1/23/76 1 ?. n- -
+ w, b i _,,
SWESSAR-P1 TABLE 6.2.1-6 MASS AND ENERGY RELEASE '[O PRESSURIZER CUBICLE - SURGE LINE DE7 "ime, Msee Mass Rutt, 103 lbm/sec Energy Rate, 106 Btu /sec 0 0 0 5 15.51 10.41 10 15.57 10.45 15 16.65 11.15 20 16.83 11.27 40 16.52 11.07 60 16.20 10 "5 80 15.93 10.68 100 16.02 10.74 120 15.72 10.54 140 15.42 10.35 1 60 15.33 10.28 g 180 15.33 10.28 200 15.10 10.13 220 15.00 10.06 240 14.91 10.01 260 14.75 9.901 280 14.60 9.803 300 14.40 9.676 320 18.29 9.602 340 14.17 9.518 360 14.09 9.467 380 13.99 0,.404 400 13.88 9.325 420 13.83 9.296 440 13.79 9.267 460 13.77 9.252 480 13.75 9.240 500 13.74 9.228 C-E 1 of 1 Amend nent 15 8/8/75 l '. ge ) I
SWESSAR-P1 TABLE 6.2.1-7 MASS AND SERGY BFT.FASE M PRESSURIZER CUBICLE - SPRAY LINE DER Time, (msec) Mass Rate, (103 lbm/sec) Enerev kate. (106 Btu /sec) 0 0 0 5-220 3.41 2.104 240-780 3.40 2.096 15 800-1,400 3.39 2.088 s C-E 1 Of 1 Amendment 15 8/8/75 bbh b
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SWESSAR-P1 TABLE 6.2.1-10 MASS ADO ENERGY R212ASES 70 COBrTAJIOOJrf, FIE!*ATER PIPE DFB Steet Generator Main Feedwat er F*euttrater Pige N in Steam Fj otr Bl e amb nen J gjtfjow hM al Mame Energy _ Mass Energy Mass amergy Mas Energy Mass Enerrry Date kate Rate Rate Rate Rate Rate Ra;e. Rat 108 10* 10s 10 Ion 10 e. 10s 10s 10s kate. 10 IlP*a S*s M8%ffrs RiWJer MavSe Bt.NFrs MeVFes Pt_uge_c 1.r.igee _stypes uw.ec Beu/sec 0.0 7.201 4.0 s 7 7.201 3.s01 0.0 0.0 0.0 0.0 14.402 7.tes 10.0 7.201 4.087 7.201 1.101 0.0 0.0 0.0 0.0 14.402 7.198 10.01 7.201 4.087 0.0 0.0 7.201 3.101 0.0 0.0 14.402 7.188 13.1 7.201 4.087 7.201 1.101 0.0 0.0 14.402 7.188 13.11 7.201 4.007 4.0 0.0 0.0 0.0 7.20 4 )97 37.2 7.201 4.007 0.0 0.0 7.20 4.087 ,3 27.21 0.0 0.0 1.900 2.325 1.960 2.325 55.3 1.960 2.325 1.960 2.325 55.31 0.0 0.0 0.0 0.0 Int egrat ed 267,877 152.04 72,010 31.01m 23,323 9.613 35,476 42.08 397,606 214.75 Amsman t s um a 10 6 stu um 10* Bt u uun x 10* Bt u um x106 htu um x 10
- Bt u
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" Amem1 ment 13 6/10/75 CD
SWESSAR-P1 TABLE 6.2.1-11 SING 12 FAILURE ANALYSIS RESULTS DESIGN BASIS BREAK IDCATION Cbntainment Peak Pressure, psig Single Active Failure BT,W C-E W-41 W-3S Diesel generator 38.7 43.4 42.2 38.6 One component cooling 36.6 41.9 43.3 39.6 water train 3 One contai.nment spray 36.0 41.7 43.5 39.6 system train 1 of 1 Amendment 36 12/21/77
' 1 \}
4, u.9-
SWE SS AR -P 1 TABLE 6.2.1-12 MAIN FEEDWATER ISOLATION TIIE5 C-E B6W W-41 W-3S(2) W-3S(3) Signal de; y time, sects) 5 5 5 8.5 5 Valve closure time, sec 20 15 5 5 5 Time isolation completed, 25 20 10 13.5 10 sects) g Notes: (n > Time af ter start of accident (2) Spray pump failure is the limiting pressure case. The pressure setpoint is reached at 6 sec. The addition ot a 2.5 see signal processing time yields the 8.5 sec delay time. (33Cerponent cooling water failure is the limiting temperature case. 1 of 1 Amendment 27 6/30/76 o6 , i tf r bLd
SWESSAR-P1 TABLE 6.2.1-13
SUMMARY
OF REAC"IOR CAVITY SUBCOMPARTMENT VENT LOSS COEFFICIENT Vent Bend Contraction Expansion Friction Area K Input to K K K K TIREED
- _
1-2 0.029 0.41 1.0 0.010 4.66 0.449 1-5 - 0.33 1.0 0.05 15.28 0.38 1-9 -- 0.50 1.0 0.18 0.74 0.68 2-3 0.029 0.41 1.0 0.010 5.82 0.449 2 -6 -- 0.33 1.0 0.05 15.27 0.38 2-9 -- 0.50 1.0 0.18 1.50 0.68 3-4 0.029 0.41 1.0 0.010 4.57 0.443 3-7 -- 0.33 1.0 0.05 15.26 0.38 0.68 4 3-9 -- 0.50 1.0 0.18 1.52 4-8 -- C.33 1.0 0.05 15.27 0.38 4-9 --- 0.50 1.0 0.18 1.30 0.68 5-6 0.056 0.17 1.0 0.018 16.47 0.244 5-9 --- 0.50 1.0 0.18 0.74 0.68 5-9 - 0.21 1.0 0.05 20.40 0.26 6-7 0.056 0.17 1.0 0.018 17.33 0.244 6-9 -- 0.50 1.0 0.18 1.50 0.68 6-9 -- 0.21 1.0 0.03 20.00 0.26 7-8 0.056 0.17 1.0 0.018 17.33 0.244 7" --- 0.50 1.0 0.18 1.52 0.68 7-9 --- 0.21 1.0 0.05 14.40 0.26 8-9 -- 0.5C 1.0 0.18 1.50 0.68 8-9 - 0.21 1.0 0.05 20.00 0.26 Note - HVFM-1 is used for all the vents listed above.
- An expansion loss a*. the vent exit with a coefficient of 1.0 is built into DIREED.
W_ 1 of 1 Amndent 10 5/15/75 nn1 bud bL i
SWESSAR-P1 TAnLE 6.2.1-14 VENT AREM, K-FACTORS, AND VENT FLOW HDDELS USED IN THE PRESSURIZER CUBICLE, PRESSURIZER SUPPORT SKIRT, AE) PRESSURIZERRET.TRP TANK COMPAR E NT ANALYSES Contraction plus friction Expansion Area, K input to Flow Vent K K ft2 THREED* Model 1-2 0.5 1.0 40 0.5 HVFM-1 HVFM-1 I2 1-6 0.5 1.0 194 0.5 2-1 0.5 1.0 40 0.5 HVFM-1 2-3 0.5 1.0 9.4 0.5 Note 1 3-2 0.5 1.0 9.4 0.5 UVFM-1 3-4 0.5 1.0 40 0.5 HVFM-1 ' 3-6 0.5 1.0 25 U.S HVIH-1 4-3 0.5 1.0 40 0.5 HVFM-1 4-5 0.53 1.0 260 0.53 1".TM- 1 4-6 0.5 1.0 73 0.5 HVFM-1 5-4 0.53 1.0 260 0.53 HVFM-1 5-6 0.5 1.0 23 0.5 HVM(-l
- An expansion loss at the vent exit with a cx>ef ficient of 1.0 is built into THREED.
Note 1 - HVD(-1 is used in the pressurizer suppert skirt and pressurizer relief tank compartment analyses. Frictionless Moody flow is used in the pressurizer i cubicle autysis for the surge line DER. l l W 1 of 1 Amendment 12 6/16/75 r, o Li
, (c. !{ N L. L
SNESSAR-P1 TABLE 6.2.1-14A NODE LENGTH-TO-AREA RATIOS FOR 'n1E PRESSURIZERRELIEF TANK COMPARTMENT ANALYSIS WITH RELAP 4 Node L/A . ft-1 16 1 0.13 2 0.40 3 0.078 4 0.072 5 0.037 6 0.0036 1 of 1 Amendment 16 W_ 8/29/75
, r: o 7 i !C uur UuJ
SWESSAR-P1 TABLE 6.2.1-15 VFW AREAS, K-FACTORS, AND VEW FIDW MODELS USED IN THE STEAM GENERATOR COMPARTMENT AND SilIELD WALL Al'ALYSES Bend Contr. Expan. Frict. Area, K Input Flow Vent K K K K fta To TimEED* Mode 1 1-2 - - 1.0 .36 528 .36 INFH-2 1-5 - 44 1.0 .02 329 46 Note 1 93 1-6 - 49 1.0 .024 106 .514 Note 2 1-9 - .5 1.0 .024 25 .524 INFM-1 1-10 -
.5 1.0 .024 25 .524 INFM-1 2-3 -
.46 1.0 .024 109 .t84 Note 3 l13 2-9 -
467 1.0 .024 74 491 INFH-1 2-10 - 49 1.0 .024 203 .514 INFM-1 3-4 -
.3 1.0 .033 211 .333 INFM-1 4-10 - -
1.0 .011 211 .011 INPN-2 5-1 2.5 .35 1.0 .02 329 2.87 WFH-1 5-6 - 47 1.0 .02 137 49 s4ote 4 li3 5-7 -
.28 1.0 .02 349 .3 INFM-1 6-1 -
.377 1.0 .037 106 .414 INFM-1 6-5 - 45 1.0 .037 137 49 INFM-1 6-9 -
.28 1.0 .037 17 1 .32 W-1 6-10 -
.27 1.0 .037 17 5 .31 M-1 7-8 - .5 1.0 .024 270 .524 uvFy-1 7-10 -
49 1.0 .024 614 .514 INFH-1 8-7 -
.5 1.0 .024 270 .521 INFM- 1 8-9 - .5 1.0 .024 237 .524 FNFM-1 8-10 - 44 1.0 .024 614 .514 INFH-1 9-8 -
.5 1.0 .024 237 .524 INFH-1 9-10 - 49 1.0 .024 614 .514 INFH-1
*An expansion loss at the vent exit with a mefficient o1 1.0 is built into THREED.
Note 1: INFM-1 is used in the analysis of the steam generator canpartment below the operating floor. Prictionless Moody flow is used in the analysis of the volumes above the in-core instrumentation drive ruom and below the I3 steam generatar support pedestal. Fute 2: BVFN-1 is used in the analysis of the operating floor and the volume above the in-core instruraentation drive room. Frictionless Moody flow is used i*s the analysis of the volume below the steam generator support pedestal.
}i 1 of 2 Amendment 13 6/30/75 0k
SWESSAR-P1 TABLE 6.2.1-15 (CONT) Note 3: HVFM-1 is used to analyze the steam genc'ator cxxnpartment l13 below the operating floor. Prictionless Moody flow is used in the analysis of the steam genfrc.or shield wall above the operating iloor. Note 4: HVm-1 is used in the analysis of the steam ger.erator ccxnpartnet below the operating floor and the volume above the in-core instru:nentation drive room. 13 Frictionless Moody flow is used in the analysis of the voltane below the steam generator support pedestal. H 2 of 2 Amendment 13 6/30/75 e n'
;g .'[, . , bcJ
SWESSAR-P1 TABLE 6.2.1-13
SUMMARY
OF REACTOR CAVITY SUBCOMPARTMENT VENT LOSS COEFFICIDfr Vent Bend Contraction Expansicti Priction Area K Input to K K K K THREED
- 1-2 0.029 0.41 1.0 0.010 4.66 0.449 1-5 - 0.33 10- 0.05 15.28 0.38 1-9(1) ---
0.50 1. 0.18 0.74 0.68 2-3 0.029 0.41 1.0 0.010 5.02 0.449 2-6,,3--- 0.33 1.0 0.05 15.27 0.38 2-9 --- 0.50 1.0 0.18 1.50 0.68 3-4 0.029 0.41 1.0 0.010 4.57 0.449 3-7 - 0.33 1.0 0.05 15.26 0.38 3-9 --- 0.50 1.0 0.18 1.52 0.68 4-8 -- 0.33 1.0 0.05 15.27 0.38 4-9 --- 0.50 1.0 0.18 1.50 0.68 5-6 0.056 0.17 1.0 0.018 16.47 0.244 5-9(t) -- 0.50 1.0 0.18 0.74 0.66 5-? --- 0.21 1.0 0.05 20.40 0.26 6-7 0.056 0.17 1.0 0.019 17.33 0.244 6-9(2) -- 0.50 1.0 0.18 1.50 0.68 6-9 -- 0.21- 1.0 0.05 20.00 0.26 7-8 0.056 0.17 1.0 0.218 17.33 0.244 7-9(r) --- 0.50 1.0 0.18 1.52 0.68 7-9 -- 0.21 1.0 0.05 19.40 0.26 8-9 --- 0.50 1.0 0.18 1.50 0.68 8-9 -- 0.21 1.0 0.05 20.00 0.26 Note - Hvm-1 is used for all the vents listed above.
- An expansion loss at the vent exit with a coef ficient of 1.0 is Imilt into THREED.
8These vents not considered in the analysis of the pump discharge break. 2These vents not considered in the analysis of the hot leg break. W-3S 1 of 1 Amendment 17 9/30/75 bbk bb
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92SSAR-P1 1 TABLE 6.2.1-14A NODE 1ENGTH@-AREA PATIOS AND FLOW AREAS FOR TIIE PPISSURIZER CUBICLE ANALYSIS WITH RELAP4 lbde L/A, ft-a Flow V ea, ft2 1 0.1 464 11 2 0.065 461 3 0.0a1 511 4 0.0287 511 5 0.004 30100 W-3S 1 of 1 Amend:nent 17 9/30/75 e n,, l' i l
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SWESSAP-P1 TABLE 6.2.1-15A NOCE LE?CTI@-AREA RATIOS AND FLOW AREAS FOR THE STEAM GENEPA WR COMPAR1 MENT ANALYSIS WITH RELAP 4 Node L/A ft-8 Flow Area it2 1 .086 243 2 .24 100 18 3 .11 184 4 .029 809 5 .033 723 6 .036 307 7 .023 237 8 .13 723 9 .003 3.02x10* W-3S 1 of 1 Amendment 18 10/30/75 t y %ss %
SWESSAR-P1 TABLE 6.2.1-16 PEAK CO2CAIt@iENT STRUCTJP.E TEMPERATURES WORST STEAM PIPE BREAK ACCIDEhT 1.4 FTa DER 0.4 FTa LDR 102 Percent Power 30 Percent Power CCh Pump Failure Spray Pump Failure Cbntainment Liner 231.1F 242.8F Cbntainment Concrete 215.7 222.6 28 Internal Thin Steelta3 247.3 255.5 Internal Cbncrete 231.8 244.9 ca) Material total thickness = 25 mil W-3S 1 of 1 Amendment 28 8/6/76 (, (, tt o") uJ
SWESSAR-P1 TABLE 6.2.1-17 ACCIDENT CHRONOLOGY STEAM PIPE BREAK 1.4 PT2 DER, 102 PERCENT POWER, CCW PUMP FAILURE Time, See Event 0.0 Accident occurs. Ruptured loop steam generator and turbine plant piping blowdown into containment is initiated. 2.5 Main steam and feedwater isolation setpoints reached. 3.351 Turbine plant piping inventory exhausted. Intact loop steam generators begin blowdown into containment. 5.0 MSIVs and WIVs start to close. 10.0 Main steam isolation valves (MSIVs) and feedwater isolation valves (WIVs) fully closed. Intact loop steam generators isolated. 14.2 Containment spray pressure setpoint (40 psia) 28 reached. 31.6 Containment atmosphere recirculation coolers start. 100 Peak containment temperature reached. 10 6 Containment sprcy enters containment atmosphere. 324.2 Ruptured loop steam generator blowdown complete-(dryout occurs). Peak containment pressure reached. 1800 Operator initiatcs termination of auxiliary feedwater flow to the af fected steam generator. 1862.5 Auxiliary f eadwater isolation valve is fully closed. Auxiliary feedwater boiloff to containment terminated. W-3S 1 of 1 Amendment 28 8/6/76
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SWESSAR-P1 TABLE 6.2.1-17A ACCIDENT GRONOLOGY STEAM PIPE BREAK
- 0. 4 P'T LDR, 30 PEACENT POWER, SPRAY PUMP FAILURE Time, see E' vent 0.0 Accident occurs. Af fected and intact 26 steam generators begin blowdown into centainment.
6.0 Main steam line and feedwater line isolation setpoints reached. 8.5 MSIVs and Di1Vs start to close. 13.5 MSIVs and PVIVs fully closed. Intact loop steam generators isolated. Main steam line inventory between break and ruptured loop MSIV starts to exhaust to the containment. 16.03 Main steam pipe inventory exhausted. All reverse flow ceases. 40.0 Containment atmosphere recirculation coolers start. 689.0 containment spray pcassure reaches setpoint of 40 psia. 779. Ontainme t pray enters containment atmosphere. 28 1800. Oprator initiates tennination of auxiliary feedwater ficw to the af fected steam generator. 1862.5 Auxiliary feedwater isolation valve is fully closed 1945.2 huptured loop steam generator boilotf cacelete (dryout occurs) . Peak contairunent pressure and temperature reached. W-3S 1 of 1 Ameadment 28 8/6/76 I, t
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FWESSAR-P1 TRmLE 6.2.1-Is PJtEAKS CDNSIDERED IN THE SUDCOMPAP1MElff ANALYFES RESAR-3SttD SWFJSAR-F 1 SuMsoportsent E(.pe g etiin.ei Type genLin ,* n Type 31owdown Data mt Leg 4660.5 LDR 144 ISR htes 3, 6 Reactor Cavity (594 .0 1ER 150 *IbR lentes 3, 6 Pisap Discharge 660.9 DERt89 1DR B.ates 8, 9 Steam Generator a:)t Leg not Isq 660.5 sts 660.5 szs isote ? Pump Buctics: 754.8 DEk IDR Mtites 3. S. 9 Ptump Discharge 594.0 DER 594.0 eSER le>tes 7, 9 PRR 153.9 DER Note 8 Aconsulator 78.5 DER taote O Pressuriser Surge Line 153.9 DER teote O Spray 12.6 DER 12.6 D Dess geyt, 7 Pressuriser surge 153.9 CER 153.9 *CER Ik>te 7 Pressuriser Beller Tank 1s tecrrtSi
- 1. Inreak areas and blowdown data are not presented directly in REEAR-38, but are referenced frtus stytes 4 and 5, respectively.
- 2. Por DIRa, the area given is one cross section only.
- 3. *A break area Isas than the cross sectional flow area of the pipe . . . will be cumsidered if it is justifiable by . . . considerations of physical constraints due to concrete walls or stnetural steel.
For exasuple, reetzaint on the pips lateral 34ption at the reactor vessel notale due to the shield wall prevents developnient of full double-ended guillotine treak, and therefore ef f ective cross sect ional flow area of the loop pipe (which is less than the cross sectional flow area of the loc.p pipe) will be used . . . WCAP B082-P-1, page 2-52. 4 WCAP-41724 and WCAP-8082-P-Ag " Pipe Breaks for the IDCA Analysis of the Westinghouse Primary (bolant toop.*
- 5. 18 CAP-8 312-A and WCAP-8264-P-Af
- Westinghouse Mass and Energy Release Data for Ccultainment Design.*
(
- 6. Istter PDA-594%'WSP-033, dated S/3 3/75 (sic) , f rom weetinghouse to Stzme s Webster.
- 7. leCAP-9312-A without the toestinghouse 10 percent margin.
S. A ccuparisen of the blowdown rates for this break with the limiting break for t his suW>artment indicates that this break is enveloped by the SNESSAP-P1 limiting break. {) CJ 9. Break area and type as .**termined by Westinghouse as ag5211 cable to the SWESSAR-P1/RESAR-3S &asign. Istter DOP-37-SMSP-061, dated 6/29/76, f ras west inghouse tn Stone & Webst er.
,r.
- Denotes limiting break f or this autmartment 1 of 1 Asendment 28 W-3S 9/6/76
SWESSAR-P1 TABLE 6.2.1-13
SUMMARY
OF REAC"N)R CAVITY SUBCOMPARTMEW VENT II)SS COEFFICIEWS FOR TITE THREED ANALYSIS Contraction Expansion Area K Input to Vent K K [Ft a) Flow
,2 FE ED* Model 1-2 0.18 1.0 16.5 1 -3 0.18 INm1 0.13 1.0 19.2 0.13 1-6 0.18 1.0 INm1 1-17 17.1 0.18 IWm1 C.05 1.0 2-1 0.?O 30.2 0.05 INW 1 1.0 16.5 0.10 INm ;
2 -4 0.46 1.0 2-11 0.29 4.6 0.46 INFM1 1.0 5.0 0.29 INr , ' 3-1 0.13 1.0 19.2 0.13 3 -4 0.06 1.0 INF. . 3-5 0.10 23.2 0.06 INW 1 1.0 20.0 0.10 3-17 0.05 1.0 INFM2 4 -2 0.46 30.2 0.05 INm1 1.0 4.6 0.46 4 -3 0 . 0.' 1.0 TNm1 4-16 0.03 23.2 0.05 1NFM 1 3.0 15.4 0.03 5 -6 0.13 1.0 INm2 5-16 0.06 14.2 0.13 INm1 1.0 23.2 0.06 30 5-17 0.05 1.0 23.2 INFM 1 6-7 0.18 0.05 INm2 1.G 17.1 0.18 6-1 0.18 1.0 INW1 6-17 0.05 16.5 0.18 INFh 4
.0 30.2 0.05 7-10 0.13 1.0 19.2 INm1 7-14 0.18 0.13 INm1 7-17 1.0 16.5 0.18 INm1 0.05 1.0 30.2 8 -9 0.1.5 0.05 INm1 1.0 19.2 0.13 8-11 0.18 1.0 16.5
}Wm1 is-17 0.0% 0.18 INm1 1.0 30.2 0.05 9-10 0.10 1.0 INFM1 9-12 0.06 20.0 0.10 INFM2 1.0 ^3.2 0.06 9-17 0.05 1.0 30.2 0.05 INFM 1 10 -7 0.13 INFM1 10-13 1.0 19.2 0.13 HVN1 0.06 1.0 23.2 10-17 0.05 0.06 INm1 1.0 30.2 0.05 11-8 0.10 1.0 INm1 11-12 16.5 0.10 INm1 0.46 1.0 12-9 0.05 4.6 0.46 INm1 1.0 23.2 0.05 12-13 0.03 1.0 15.4 INFM1 13-14 0.46 0.03 INFM2 1.0 4.6 0.46 13 -10 0.05 1.0 INm1 14 -7 0.10 23.2 0.05 INFM1 1.0 16.5 0.10 14 - 13 0.46 1.0 INm1 4.6 0.46 INFM1
- An builtexpansicra into THREED. loss at the vent exit with a coefficient of 1.0 is B&W 1 of 2 Amendment 30 1/28/77 u 05
SWESSAR-P1 TABLE 6.2.1-13 (CO?TP ) Cbntraction Expansion Area K Input to Flow Vent K K iPt 8) THREED* Model 15-6 0.10 1.0 16.5 0.10 15-14 0.29 HVml 16-5 1.0 5.0 0.29 MVFM1 0.05 1.0 23.2 0.05 30 16-15 HVFM1 0.46 1.0 4.6 0.46 HVFM1
- An expansion loss at the vent exit with a coefficient of 1.0 is built into THREED.
MW 2 of 2 Amendment 30 1/28/77 (;.4 p d ' 0,
SWESS AR-P 1 TABLE 6.2.1-13A
SUMMARY
OF REACTOR CAVITY SUBCOMPARTMENT VENT LOSS CDP.FFICIENTS AND EFFECTIVE L/As POR RELAP4 MOD 3 Contraction Expansion Area K Input Effective Vept K K (Pt r) ___ to RELAP L/A ft-1 1-2 0.18 .098 16.5 0.278 .201 1-3 0.13 .055 19.2 0.185 .359 1-8 0.18 .241 17.1 0.421 .315 1-17 0.05 .997 30.2 1.047 .127 2-1 0.10 .179 16.5 2 -4 0.279 .201 0.46 .416 4.6 0.876 .748 2-11 0.29 .724 5.0 1.014 .543 3 -1 0.13 .029 19.2 3 -4 0.159 .359 0.06 .050 23.2 0.110 .176 3-5 0.10 .226 20.0 0.326 3-17 .285 0.05 .997 30.2 1.047 .117 4-2 0.46 .311 4.6 4-3 0.771 .748 0.05 .063 23.2 0.113 .176 4-16 0.03 .355 15.4 0.385 5 -6 447 0.13 .029 19.2 0.159 .359 30 5-16 0.06 .050 23.2 0.110 .176 5-17 0.05 .997 30.2 1.047 .117 6-7 0.18 .241 6-15 17.1 0.421 .315 0.18 .098 16.5 0.27S .201 6-17 0.05 .997 30.2 1.047 .127 7-10 0.13 .055 19.2 0.185 .359 7-14 0.18 .098 16.5 0.278 .201 7-17 0.05 .997 30.2 1.047 .127 8 -9 0.13 .055 19.2 0.185 .359 8-11 0.18 .098 16.5 0.278 .201 8-17 0.05 .997 30.2 1.047 9-10
.127 0.10 .226 20.0 0.326 .285 9-12 0.06 .050 23.2 0.110 .176 9-17 0.05 .997 30.2 1.047 .117 10 -7 0.13 .029 10-13 19.2 0.159 .359 0.06 .050 23.2 0.110 .176 10-17 0.05 .997 30.2 1.047 .117 11-8 0.10 . F;; 16.5 11-12 0.279 .201 0.46 .416 4.6 0.876 .748 12-9 0.05 .063 23.2 0.113 .176 12-13 c.03 .335 15.4 0.385 .447 I 13-15 0.46 .311 4.6 0.771 .748 13 -10 0.05 .063 23.2 0.113 14-7 .176 0.10 .179 16.5 0.279 .201 14-13 0.46 .416 4.6 0.876 .748 15-6 0.10 .179 16.5 0.279 .201 B&W 1 of 2 Amendment 30 1/28/77
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SWESSAR-P1 TABLE 6.2.1-13A (C0!rr) Contraction Expansion Area K Input Effective Vent 3g K K _ fPt ) to REIAP I/A ft-s 15-14 0.29 .270 5.0 0.560 .829 16 -5 0.05 .063 23.2 0.113 .176 16-15 0.46 .311 4.6 0.771 .748 MW 2 of 2 Amendment 30 1/28/77
, r. .' ,
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SWESSAR-P1 TABLE 6.2.1-14 VENT AREAS, K-FACTORS, AND VENT FIDW MODELS USED IN THE PRISSURIZER CUDICLE ANALYSIS WITH THREED Area Bend Contr Prict K input THREED Vent Fta K K K to THREED* Flow Model 1-2 221 - 0.52 0.02 1.54 IN m-1 1-7 1.2 - 0.50 0.03 0.23 HVM4-1 1-7 42 1.59 0.38 0.03 2.00 HVR4- 1 2-1 221 - 0.60 0.02 0.62 2-3 MVM4- 1 16 - 0.49 0.03 0.52 HVR4-1 2-3 24 - 0.76 0.02 0.78 2-4 55 HVm- 1 0.74 0.02 0.76 HVD4-1 2-6 28 1.09 0.47 0.03 1.59 2-7 21 IND4- 1 0.31 0.45 0.03 0.79 HVFM-1 3-2 16 - 0.48 0.03 3-2 24 0.51 HVm-1 0.75 0.02 0.77 HVD8.- 1 3-4 42 - 0.45 0.03 N 3-5 126 0.48 HVm-1 0.33 0.03 0.36 HVm-1 3-7 123 - 0.42 0.03 0.45 INM4- 1 4-2 55 - 0.60 0.02 0.62 4-3 42 - HVm-1 0.44 0.03 0.47 HVm-1 4-6 28 1.09 0.44 0.03 1.56 4-7 28 - IWm-1 0.44 0.03 0.47 INEi-1 4-7 28 - 0.46 0.03 5-3 126 0.49 inn 4-1 0.31 0.03 0.34 HVP4-1 5-7 124 2.8 0.0 0.03 2.83 Innd-1 6-2 28 1.09 0.47 0.03 6-4 28 1.59 HVm- 1 1.09 0.47 0.03 1.59 IND4-1 6-7 1,250 - 0.47 0.03 7-1 1.2 - 0.50 HVm-1 0.47 0.03 0.50 MVm-1 7-1 42 2.57 0.50 0.03 3.10 7-2 21 INFM-1 0.28 0.50 0.03 0.81 HVMi- 1 7-4 28 - 0.47 0.03 0.50 7-4 2b - HVFM-1 0.47 0.03 0.50 IRTM-1 7-5 124 3.22 0.5 0.03 3.75 HVMi-1
- The i.e.,
THREED code include one velocity head loss at exit of vent, the expansion K=1. B&W 1 of 1 Amendment 30 1/28/77 i, (- .,
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SWESSAR-P1 TABLE 6.2.1-15 VEITF AREAS, K-FACTORS, AND VENT FLOW MODELS USED IN TliE STEAM GENERATOR COMPAR1HEfff AND SHTRTD WALL ANALYSES 76-Node Model Area Bend THREED Omtr Expan. Frict. K Input Flow Vent ft K K K K % THREED* Hrxiel 1-6 140.0 1.2 .5 .997 .03 1.73 INFM-1 1- 7 60.0 1.2 .42 .902 .04 1.66 INFM- 1 1-7 175.0 1.2 .39 .728 .04 1.63 1-9 100.0 1.95 .5 INFM-1
.790 .04 2.49 INFM- 1 2-1 62.5 -
.366* .885 .02 2-3 0.386 INFM-1 47.0 -
.466* .920 .02 0.486 INFM-1 2-10 120.0 -
.130 .079 .02 0.150 2-11 120.0 INFM- 1
.130 .056 .02 0.150 3-4 391.0 -
.25 -
INFM-1
.03 0.28 INFM- 1 3-6 185.0 -
.252 .997 .03 0.282 3-6 185.0 -
.252 INFM-1 3-6 32.0 0.32
.997 .03 0.282 INFM- 1
.43 .999 .01 0.76 3-6 32.0 0.32 INFM-1
.43 .999 .01 0.76 INFM- 1 34 4-5 252.0 -
.15 .236 .03 0.18 INm-1 5-6 550.0 - -
.941 .04 0.04 3-6 686.0 -
.5 INFM-2
.996 .1 0.6 INFM-1 8-7 100.0 0.75 .33 .740 .04 1.12 INFM-1 9-8 220.0 -
.186 .184 .04 0.226 10 -1 32.0 -
.663*
INFM-1
.940 .02 0.683 INTM-1 10-2 120.0 -
.127 .093 .02 10 -3 0.147 INFM- 1 82.5 -
.487* .861 .02 10-12 170.0 0.501 INm- 1
.062 .183 .02 0.082 INFM-1 10-14 100.0 -
.18 1 11-1 52.5
.161 .02 0.201 INFM-1
.484* .903 .02 0.504 11-2 120.0 INm- 1
.106 .003 .f 2 0.126 INFM- 1 11-3 53.7 -
.484 .908 .02 0.504 INFM-1 11-12 155.0 -
.055 .157 .02 0.075 12-1 140.0 -
.585* INm- 1 12 -3 223.0
.752 .02 0.605 INN- 1
.469* .648 .02 0.489 12-6 32.0 0.32 .43 INFM-1
.999 .01 0.760 INm-1 12-10 170.0 -
.192 .019 .02 12-11 0.212 INFM-1 155.0 -
.179 .015 .02 0.199 12 -16 170.0 -
.152 INFM-1
.114 .02 0.172 13-1 62.5 -
.366* .885 INm-1 13 -3 47.0 -
.02 0.386 INm-1
.466* .920 .02 0.486 13 -14 120.0 INFM-1
.130 .079 .02 0.150 INm-1 13-15 120.0 -
.130 .056 .02 14-1 32.0 0.150 INm- 1
.663* .940 .02 0.683 INFM-1 B5W 1 of 2 Amendment 34 7/22/77
'~
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SNESSAR-P1 TABLE 6.2 1-15 (CONT) 16-tiode Model TH REED Area Bend Contr Expan. Frict . K Input Flow Vent ft* K K K K To THREED* Mt> del 14 1 82.5 -
.481* .861 .02 0.501 INFM-1 14 -10 100.0 -
.18 1 .161 .02 0.201 INFM-1 14 -13 120.0 -
.127 .083 .02 0.147 INFM-1 14-16 170.0 -
.062 .183 .02 0.082 INFM-1 15 -1 52.5 -
.484* .903 .02 0.504 INFM-1 15-3 53.7 -
.484* .908 .02 0.504 INm-1 15-13 120.0 -
.106 .083 .02 0.126 INFM-1 15-6 155.0 -
.055 .157 .02 0.075 IN m-1 38 16-1 140.0 -
.585* .,752 .02 0.605 INFM-1 16 -3 223.0 -
.469* .648 .02 0.489 INFM-1 16-6 32.0 0.32 .43 .999 .01 0.760 INFM-1 16-12 170.0 -
.152 .114 .02 0.172 IN m-1 16-14 170.0 -
.192 .019 .02 0.212 INFM- 1 16-15 155.0 -
.179 .015 .02 0.199 INFM-1 8An expansica loss at the vent exit with a coef ficient of 1.0 is built into THREED.
- This contraction K-factor includes K=0.33 for the 75 percent open grating at the node junction.
B&W 2 of 2 Amendment 34 7/22/77 a 7 (> i -t u l*tL. u
?
m *
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mN kN M E6 E o ) 5,
- e--neNOpmee...
MNNeeedem**mme SempeOme.Me-eNo#OedeNOeemeeO W MmeeededNNeNeeecedeeedese*de b 0000**O0000000 000000000*O*O000000*PO000000 e e e e e e e e e e e e e a e e e a e e e e e e e e e e e e a e e e o e e e a e e e W M 00000000000000 0000000000000000000000000000 E h Om W> M J NNeSeded Seped OMONeNkpNMmbep edadeemNNom U ed g SNeeceomeeeeee weNMO*#eemer*****edesceOe*** demMNMNpMNNMPM p e e M @ N m N m M p e te m p b N P N N m N P d e m M N e e a e o e a e e e e e o e o e e e e o e e e o e o e o e e e e e o e e e o e e e a M - NNNMPPOOO#***O OpwOpO*ComoeOM*wCOO*POO*pOOO .* - o r e meeeNNNNemm.*m e weNNNN~NN~NNN NNNN~NN~NNNN N
- M OOOOOOOOOOOOOO O*O000000000000000000000000O w N g e e e e e e o e e e e o e e e e e e e e e e e e e e e e e e e e a e e e e e a e e e Q e O w hh "
8* MNGOeOS Mkeed *doeO****mMSMNeee#em OedO**** b eONeeNN epeem eeem ededOeodete***E NNeeddee 4 f W Sekke#OOl@ee#N e e e e e e e e e e e e e paeeme "e o Oe**SopwMeeOOpa%COce*Oe e e e e e e e a e o a e o e a r } e e o e e e e U M d400 Ne dommeecmMe NN dmm>Needecee NeNdeOOm**mN demeMempHNNee* medNedericacedmemedemedeeNJ W 4AM*dreO*epeOA#e***me**de**O W e e e e e e e e e e e e e e e e e a e o e e e e e e o e a e e o e e e e a e e e e L W e NN e N m* MM N m 4 k J NNMee e e e1 3 3 g g ag e e e pow
- OO g g e g g g g i eie1 1 0 g ol i E i I 4 l l t I t t O O N OOOOd O0000000 e o e e eO e e e e o e e o OOOOOOeOCAOPOOOOOOO+OOOOASOO e e e a e e e e e e e e e e e e o e ; e a e e e e a e e Z Q e OceON eOO*eeNNN OdOONONOONomeSmMOcoNMOONNCOO E %
s *eJM4dMNN,emeMMe w> Sp* ww N eeON,NemOeNeceNMNehdeNNmeONS
@d*N e 99 9 #wN wep we woe Ne N Qwd de Omd Op FMepMe* e *NmeppNmppmppeewmappepew Met e I t I e 4 4 I t t i l e I $ e e o e e e e s ,3 d 6 I e e i e g e e eOOOOOpawwNNNNNNemmmespee P peppHNNNMMMMMe WWeepperoppppppppppppppppppe m e
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SWESSlut-P1 TABIJ 6.2.1-15A (GMT) 16-tKOE Mm AREA BEND CDerTH EXPflu F1t1CT R INPUT INERT 1= TERM VENT FTa L K K K 1T) REIAP @R REIAP. FT-s 15-1 52.5 - 484* 903 .02 9.387 0.123 15-3 53.7 - 484* .908 .02 1.392 0.123 15-13 120.0 - .106 .083 .02 0.199 0.049 14 15-16 155.0 -
.055 .157 .02 0.212 0.065 M- 1 140.0 - .585* .752 .02 1.337 0.042 16-3 223.0 -
.469* .648 .02 1.117 0.040 16-6 32.0 0.32 .43 .999 .01 1.749 0.055 16-12 170.0 -
.152 .384 .02 0.266 0 . 0 .* '.
%-14 170.0 -
.192 .019 .02 0.219 0.060 h- 15 155.0 -
.179 .035 .02 0.194 0.065 This contraction K-factor includes K*0.33 for the 75 percent open grating at the node junction.
Os es .l-B&H 2 of 2 Aw-1:arnt 34 9 9
E LSShR-P1 TABLE 6.2.1-1b PEAK CC2fTAIRMENT STRDLM TEMPERATURES, WORST STLAh FIPh BREAK ACCIDMIT, 104 PERCENT PCAGh, MS1V A:fD DG FAILukd 8.55 Fta 2.0 Fta 0.33 Ft2 DER BreaX hrea;< Contat rtwnt TA ner, F 179 lb2 186 Cont.alnment Ch e ete, F lo9 154 176 Internal concrete, F 199 184 196 Anternal Thin Steel, F 267 257 269 1* W 1 of 1 Ameninen t 39 7/14/78 e a F I -' f) - t
SWR m R-P1 TektR o.2.1-17 E CIDE.NT G RONOLOGY STLAM LINE hKhnK 8.55 FT2 Drm , 104 PERCLNT PUWER, MSIV & DG FAALUdt Tree , Sec Event D.u accident occurs. blowoown into containtaent begins. 2.* Main steazir and feedwater isolation setpoints reacneu. 4.9 MSIVs and MFIVs start to close:, 9.9 Main steam isolation valves (MSIV) tullt closed. Intact loop steam generator isolateo. 21.>* Main reedwater isolacAort valves Iully closeo. 3g 31.0 Containment atmosphere recirculation coolers start. 38.m Pen.K contain wns armaphoric romperature reacLeo. Peax cent inwnt pressure reacned. 93.4 kuptureo loop steam generator Dlomiown cunplete.4 (dryout occurs). 114.u Conte nment sprays start.
'The requireo closure time for MF1V As 15 seconds pee Section 10.4.7.1) . For analysis purposes, D6W ano sow useo a 38 17 secono closure rim to provide auditional conservatis:n.
O m*W 1 of 1 Amenuent 3a 5/31/78 (, b !) i'yr' U
dVM't'%x-P 1 TAMLK 6.4.1-17a ACCIDENT CHM 11011JGY ST1.hM LINr. BhdAA 4.0 FT2 MSLB, 102 PERChNP POWER, MSIV & DG Foi die Tinks , see Event 0.0 acclaent occurs. begins. tilcudown unto containaa_nt 4.5 Main steam reached. and reenwater isolation setWinu 7.0 Main steam isolation valves (MSIV) and auun teeawater isolation valves start to close. 12.0 M5AVs fully closeo. Intact loop steam generator isolated. 24.U* MF1Vs tully closed. 38 35.2 uxitainment atmosphere recirculation coolers start. 44.8 Peax cour a nunent atmosF;*xic taperature and peut conem - nt pressure reacnea. 141.u containment sprays start.
- Die rt=;utred closure time for 18FIV is 1S secunds (see Section 10.4.7.1) . For analysis purposes, at;h 3g anc SuW use<1 a 17 secona closure time to provide
.*Mi tional conservatism.
b4W 1 or 1 ameniment 38 5/31/78 i I l
SWESSAR-P1 TABLE 6.2.1-17B ArmnMir CHRONOIDGY, STEAM LINE BREAA, 0.33 FTa SLB, 102 PERCENT POWER, MSIV & DG FAIIAIRE Time faec) Event 0.0 M:cident occurs. Ruptured loop steam generator and turnine plant piping blowdown into pontain - t is initiated. 23.0513 niin steam isolation valves (MSIVs) and main feedwaterreached. setpoints isolation valves (MFIVs) isolation Auxiliary feedwater (APW) initiation signal setpoint reached. 30.5 MSIVs 1solated. fully closed. Intact loop steam generator 38 42.5ta) K'IVs fully closed 4d.0tz) Reactor shutdown complete 56.0 containment atmosphere recirculation coolers start. t>3.0 sdW flow to both steam generators started. 290.0 FOGG isolates ruptured loop steam generator. 389.0 Peak containwrit pressure reached. Peak containment tanperature reached. 390 0 Ruptured loop steam gener.ator blowdown completed (dry-out occurs) . (1) Values reflect pressure first equals 4.0estimate psig. of time when containment ta>
' Die required closure time for MFIV is 15 sectmda Section 13.4.7.1) . For analysis purposes, B&W and(see S&W use a 17 second closure tim ** to provido additional conservatism.
(33 Reactor shutdown containment is assuned pressure reaches to4.0occur psig.25.0 seconds after B&W 1 of 1 Amendment 39 7/14/78 nin Uut l UtU
5 3 y y g) g) tr. r . reea rtea olmnt ttea oent 5 cu d tl ea 31 aod cn d 1 aid 8 e rlas ne e ne t/ n6 8 e a rlas 8 e/ esee e a ' n0 a 2 8 esse t MI R Rssal i1 R a I 2 1 E EUeae a D D(vmr D D E Rsssl LUeae za t R Euet s2 t m e D(vmr D DDDD n A
"5) 8 n
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> t t t 8 y o o t 8 y tI N o o o os 8 N u t N N I 7
S 2 I 1 S 0l Y 1e 6 L - o 6, A N Wn 4 4 Ai 6 3 6 A Bf s s 6 3 5 5333 a e e e et e e r v e r t o t t t t e t e eeee t u A N 5 0 o o o o o o oooo tttt ri 1 N N N N P i N NNNN A A P 8 M e 1 O p RRRRR RRRRR R RRRR 1 - C RRRR RRRR P 1 B U 5Y EEDDD 0T 2 DSIII EEDDD DSILI E I EEDD DSII EEDD DSIA EEEE R 2 S - DDDI 2 A R f S 6 E A o S I 1 F E E T - 3 W L B 8, 1 S B N A I n 1 4 T i 68 26 8 1 2 5 4303 D f 573 57 3 8 573 68 52 2 2 5899 E e 78950 52510 3 e 789 6727 R E c t 7890 1570 e a 3515 I 57321 t t D 04221 1 o 5731 1521 o R 1 N 1 o 1101 I 1 N N S N 1 I C S K ) A )
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$v sr d
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SNESSAR-P1 But et, a TABLE 6.2.1-10 (COPrr) 1) NTA Pipe1976. Decvaber Break Criteria for the Design of Babcock & Wilcos teaclear Steam Systems
- tbpical
, It eport BAw-10127,
- 2) Break location not cruneidered in B-SAR-205.
- 3) Ereak location not considered in BAW-10127 4)
As stated in BAW-1GI27 the break area is determined by an iteratlee method SME.SSAR will be ctmfirmed by the Utility-Applicant in his . The break area assumed in application for a construction petuit.
- 5) All breaks are of the guillotine tyle.
- 6) Pipe tareak areas are not gieen in RAW-10127.
7) leot analysm1, not 11.miting break due to less severe mass and emergy release. 35 CN (^ 7
~
l' ( -' B&W 2 of 2 C h h at 35 10/6/'#
SWESS AR-P 1 TABLE 6.2.1-13 SUMVARY OF REACTOR CAVITY SUBCOFTARTFENT VENT LOSS COEFFICIENT Vent bend Contraction ExpansAon Priction Area K Input to K K K K THRIED
- 21 1-2 0.08 0.07 1.0 0.05 34.9 0.20 1-6 0.08 0.07 1.0 0.05 36.9 0.20 1-7 ---
0.25 1.0 0.02 22.3 0.27 1-13 --- 0.05 1.0 0.02 34.3 0.07 2-3 0.06 0.07 1.0 0.05 36.9 0.20 2-8 --- 0.25 1.0 0.02 22.3 0.27 2-13 --- 0.05 1.0 0.02 34.3 0.07 3-4 0.08 0.07 1.0 0.05 36.9 0.20 3-9 --- 0.25 1.0 0.02 24.3 0.27 3-13 --- 0.05 1.0 0.02 34.3 0.07 4-5 0.08 0.07 1.0 0.05 34.9 0.20 4-10 --- 0.25 1.0 0.02 22.3 0.27 4-13 --- 0.05 1.0 0.02 34.3 0.07 5-6 0.08 0.07 1.0 0.25 36.9 0.20 5-11 --- 0.25 1.0 0.02 22.3 0.27 5-13 --- 0.05 1.0 0.02 34.3 0.07 6-12 --- 0.25 1.0 0.02 24.3 0.27 6-13 --- 0.05 1.0 0.02 34.3 0.07 7-8 0.02 0.47 1.0 0.01 2.44 0.50 7-12 0.02 0.47 1.0 0.01 2.15 0.50 8-9 0.02 0.47 1.0 0.01 2.15 0.50 9-10 0.02 0.47 1.0 0.01 2.15 0.50 10-11 0.02 0.47 1.0 0.01 2.44 0.50 11-12 0.02 0.47 1.0 0.01 2.15 0.50 Note: HVFM-1 is used for all vents listed above.
- An expansion loss at the vent exit with a coef ficient of 1.0 27 is built into THREED.
C-E 1 of 1 ic .endmen t 27 6/30/76 pr (i( bJ
SWESSAP-P1 TABLE 6.2.1-14 VENT AREAS K-FAC':VRS, AND VENT FLCM MODEIS USED IN THE PRESSL'RIZEk CU BICLE , PRESSURIZER SUPPORT SKIRT,
.AND PRESSURIZER RELIEF TANK COMPAR'IMEh"P ANALYSES Contraction Expansion Priction K Input to Area Flow Vent K K K THFEED* fita) Mcxiel 1-2 .395 1.0 .057 452 36.6 HVFM-1 1-6 .172 1.0 .057 .23 184. HVFM-1 8 2-1 .14 1.0 .024 .164 36.6 HVFM-1 2-3 .43 1.0 .024 454 2.4 NOTE 1 3-2 .45 1.0 .06 .51 2.4 l'VFM- 1 3-4 .43 1.0 .06 .49 12.4 HVFM-1 3-6 .39 1.0 .06 45 30. HVFM-1 4-3 .43 1.0 .019 .45 12.4 HVFM-1 4-5 .07 1.0 .51 .58 234. EVF.- 1 4-6 .373 1.0 .019 .392 47. HVTM-1 5-4 .15 1.0 .50 .65 234. HVFM-1
- An expansion loss at the vent exit with a coef ficient or 1.0 is built in'o THFIED.
N,LO 1 - WrH-1 is used in the pressurizer support skirt and pressurizer relief tank cor::partment analyses. Frictionless Moody flow is used in the pressurizer cubicle analysis for the surge line DER. C-E 1 of 1 Amendnent 18 lof30/75
!i'
,} DU
SWESSAR-P1 TABLE 6.2.1-14A NCDE IENGTH-TO-AREA RATIOS A?O FLOW AREAS FOR 'IEE PFISSURIZER CULICLE ANALYSIS WITH REIAP4
!ktde L/A, ft-1 Plow Area, fta 17 1 .154 298 2 .1 53 3 .11 219 4 .03 276 5 .021 345 6 .003 34,000 C-E 1 of 1 Amendment 17 9/30/75 hhk bs '
7% 272 17
/
a8, et 0007%740409690 0001 2221 177769290 1 1 1 3 3 t3 n3 1 111 1 ./ AJ r >
+9 n
lD e m e E. A df oF 21221211112221 hi t
- - - - - - - - - - - - - - 0 T MMMMMMMNMMMMMM 1 sD I. N w tttPtt t P P t t P tt or VNVYNVVVVVNvVV t A l o tIHHI HHHRHI I HH f f f
OT Fl o K t PN n E e MM 1 i A IP c FA P te W_15476779721 tP 6 4 1 i f rMS nF 0093438984,8483 f e f) E I E0SY f o 1 1 1 m
- V KT 1 RL 1 - DUA a P 1 N1N
- h 1
- AAA T 4 5 t R 2 P 4 6 33232 f A .FEL tC 1587 07 i o
S 6 S N. I w S RPA u% p 0 0 0 3 4 3 8 7. d. 4 0 0 0 4 . 1 E nT t 1 L i h I x S b A NGW M1 AAI J KT o e T FEE
- TI t KSI . n N t 484 0 058 e
,E c 1217377773323I v Sf iK 00030333300600 AM L
r F h e RN t AI 52
. 2 7 3 3 5 tD TD n 0 3 9 63697778 aJf NE e 0 9 9002097379 FS pF sP VU x 0 0 0 sM E
l oT o
. 2 4 4 nt r 2 3 292 9 2
on t nK 0 4. D 4 3 4, 4 ii s o 0 00 Oo 000 nt c el
}i xu eb t
m 2183 s245687778 t ns Ai L V 12122333334$67
- C
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Ct E_
SWESSAR-P1 TABLE 6.2.1-15A NODE LENGTH 'ID-AREA RATIOS AND FLOW AREAS FOR THE STEAM CT. RATOR COMPARTMENT ANALYSIS WITH RELAP 4
- Jode
. Flcm .rea,
- ft2 L/A, ft-8 1 1,200 .0125 2 1,260 . 02 t4 4 g 3 1,260 .019 4 99 .101 5 326 .0307 6 99 .101 7 826 .02 8 3.3x10* .003 C-E 1 of 1 Amendment 17 9/30/75 b y -t bb
SWESSAk-P1 TABLE 6.2.1-16 PEAK CONTAINMENT S1RUCT11stE TEMPERA %W M)RST STEAM PIPE BRFM ACCIDENT T e rature, F Containment Liner 220.0 13 Containment Concrete 195.5 Internal Steel 256.7 Internal Concrete 229.7 C-E 1 of 1 Amendment 13 6/30/75 i ,- , n ,- JV
SWESSAR-P1 TABLE 6.2.1-17 ACCIDENT CHRONOLOGY STEAM PIPE BREAK Time, Sec Event 0 Accident occurs. Ruptured and intact loop steam generators begin blowdown into con-tainment. 7.4 Reactor scram initiated. Main steam iso-lation valves (MSIV) and feedwater isolation valves (FWIV) start to close. l24 One MSIV fails to close. Turbine plan-piping blowdown into containment is initioted. 12.5 MSIVs fully closed. Intact loop steam steam generator isolated. 17.1 Turbine plant piping blowdown complete. 27.4 FIVs fully closed. 31.8 Containment atmosphere recirculation coolers start. 123.0 Containment spray enters containment atmosphere. Peak containment pressure and temperature reached. 170.4 Ruptured loop steam generator blowdown complete. Main feedwater pipe blowdown initiated. 222.7 Main feedwater pipe blowdown complete. 1 of 1 Amendment 24 C-E 4/23/76 oI7 e/$ 'lv 1/ 9 .;, 5
SWESSAR-P1 TABLE 6.2.1-18 Bl<EAFS CONSIDEPED IN TifE SUlKUMPARTMEPFP ANALY.WS CESSAptsa CENPD-16 stat swyssap-pt Subnerpartment Pipe Mea fin.83 h , Area fin.a1 my bgea lin.81 M Blowknen D t m Beactor Cavity Hot leg 2,771 DER 100 IDR Not e 3 Hot leg 1,385 SES Not Specified 1,385 *Sp1118*3 Notes 5 and 6 Pump discharge 1,413 DER 350 IDP 350 LDP Note 7 Pump discharge 707 SES 532 Split 532 Split
- Note 7 Steam Generator Hot leg 2,771 DER 600 IDR Skit e 3 Subcumpa rtment Not leg 1,385 EE.S 812 Split 1,385
- Split Notes 5 and 6 Pune discharge 1,413 DER 400 1DR 900 1DR Notes 5 and 6 Pump discharge 707 SES 532 Split Not e 3 Pump outtien 1,413 DER 430 IDR Note 3 Pump suction 1,413 DER 592 FDR Note 3 Ptunp suction 707 SES 532 Split Note 3 Pressurizer surge 161 DER Not specified 161
- DER Note 8 Suhmespartment Spray Not Specified Not Specified 10.8 DER Note 9
~_
C .' _R 2 CJ L7 C:)
- Denotes limiting break for this subetwpartment .
C-E 1 of 2 Amendment 24 4/23/76
p 4 46 27
/
t3 n2 e/ m4 as r e m A
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- 1. ka ngs f oa i i t e eq of v 1 etii s vk t c o rnss i ia I enr cop 2,b e e e n ter t
, mdh a st ni p 6ct h ceb uo t atA iryn c p mh i e ua nt at g e ss f nh i scn o s pii m ei seort si g i i s' vs rh ih e ili t nuae d t st yPn ps cac cd n e y t i epe ehb a ht lfi n haD Scure t a aosod t esoh a h n sin el Rm ft i ot aneta oha A e r t occ tt n Sahds e s sieu, i S t eay t ne st nrl ea rnF W En Co"od dl e i
r c wt o. dc rl esv t uhno t a i oi ra1 3 E S
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rgavu h e od l n s a opd.C p deo rr z W ei,e vs e t bi a A[ eti t l W & ee6 eb t e dir u & S wd1b s et aedrn i rp h S o - o e n hn t e hti og t avfi vc o I S c s n H ty 2. tot i m ro s rkP t w
.i6 s f f t dirne pa- . l o Rv Aaega yk d
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)
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- 1 e e n si hrnet b cp t ,
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E 2 eenym e e nN er ii ni. e li rl Ct eg a - 2 W t 5 rsil .e ho tf t np pa yees fi I r C b S aat ne 5e oosd v rhre , e E booh 7s gk Mi e1 o at D 'O b c L e"l l t 9( na (t gwPr n t e I B sp yit y e 1 i e aee- p ionl . c A n dr gcliRp mt aa I . T i auaf yg ub ni vAA i cn P. A, pmqri li l il t eS ldii J . , tt us cg rporSn eelF A, W lt asn Jed ii nr s e ph ea Eg rt p papr o W m l neue eWi m aal1d , (t n abSs dAo r o c 1i . s1 i iR e npf ouy f m r nit 1 7 kP sm. gA ,l gD o f il o l1 a- eik nSsl n t n , r phRl - eP tl a ESuiil dn= i o 4 f , sg a e E SadSw. Ai2 rA BS rer nCnwcann sit eili 7
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4 7
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.i r t ei F alt a blUc i i
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n dt e hF st pf r dwal 1 3 6 n n a P e at - uRiseo bAv e msyp d / p 1 at i's p wiR rf d a ah owA Sag riba e 2 e a h en Rtl e d t ", it s l bks CE S eScetn l no os f yt1 t a d 7 t a E bdat ao aE Wr ai rliP d e 3 d Dsuicn Bt eeW hSotct eam- , a t - eovie hrs t t uia pnrR 1 , r oatd r ol pt m glne tb iechl e what c pi aeA s PS 7 6 d 2 9 9 fteppr ib f sh e e pl as S 5 , 2 sht Aa sa oit e rhAp wine - 2 6 - a ap ec h cr t - p hoW 4 3 4 ty t ym Di nty noe o b e f h gt ya sTiS 7 - el 7 alf at
,p l
sr i t - C - dl odicl d tni1 s . ce E R b E a 8 p i od nd ilP y8 uh P P a P n cd nih 6 a rf e oenti - l6 rt P R T P wiawtc 1 a p pmiatR a1 t of eoUa - s p)o sr lUA n sn r r R r ditd e Dk m sl a eotu S aDni e e A e wcswh Pa ree rfit aS E P o eNCd t t S t oenocr Ne cdv trl s t t S t l pi l ao Er E n sI CE Bs(bEf Cb Ame on bC pes Eb pcS yW nCaa Is I s
) ) ) ) ) )
- 1) 2) 3( 4) 6 9
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. ) 7 O(~ A 'C L C
SWESSAR-P1 Tables 6.2.1-19 and 20 are deleted 28 ( c ,_ r n'n
'uU 1 of 1 Aniendment 28 8/6/76
SKESSed< -P 1 TA11LE 6.2.2-1 CO?TTAINMENT SPRAY SYSTdM AILM,YSIS DATA W-41 W-3S bf,W C-E Spra y Pump timber 3 2 2 2 Type Vertical Vertical Vertical Vertical Ratna flow, gpm 3,500 3,700 3,700 3,700 Ra ted head , it (prelinina ry) 470 440 470 470 Mechanical Mechanical Mechanical Mechanical Seal Material 304 SS 304 SS 304 SS 304 SS Retueling water Storage Tank -y 1 1 1 1 Number (,k. )- 508,700 (315,100 '34 Volume, gal {l .9 528,800 520,800 Doron concentration, ppm w) V 2,000 Nominal 2,000 Nominal 2,270 Minimum 4,400 Nominal Design pressure, psig (M
- v. s Ifydraulic head Ilydraulic head !!ydraulic head I!ydraulic head Design tesaperatare, F $ $q 4 150 150 150 150 Operating pressure, psig cQ 1,5 -.m Ilydraulic head liydraulic head Ilydraulic head ifydraulic head cc l}
4 <120 <120 <120 <120 Operating temperature, F 2 Diameter (inside) , it [# g 37 37 37 37 Material A2'4 0-T304L A240-T304L A240-T304L A240-T304L Design code AME III, Class 2 ASME III,Clasa 2 Af2(E III Class 2 ASME III, Class 2 -[ 4- Refuelinq Water Chemical Addition Tank 1 1 1 1 Number Vertical Vertical Vertical Vertical Type cylindrical cy12ndrical -- cylindrical cylindrical Cw 9,200 4,600
" usable voltane, gol 10,900 5,000 fu Ilydraulic head -!Iydr a u li c head Ilydraulic head Ilydraulic head Design pressure, psig 1 of 3 Amendment 34 7/22/77
" y 9 d e
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SWESSAR-P1 TABLE 6.2.2-2 FAILURE ANALYSIS FOR CONTAINMENT HEAT REMOVAL SYSTEMS Component Malfunction Comments and Consequences Containment Pump casing The redundant train (s) supply the Spray Pump ruptures. necessary heat removal capacity. Containment Pump fails The redundant train (s) supply the Spray Pump to start. necessary heat removal capacity. Containment Valve fails The redundant train (s) supply the Spray Pump to open. necessary heat removal capacity. Discharge Motor Oper-ated Iso-lation Valve Containment Check valve The redundant train (s) supply the Spray Pump in pump dis- necessary heat removal capacity. Discharge charge line Check Iso- sticks closed. lation Valve Containment Valve fails Motor operated valve in suction line Spray Pump to close when from RWST closes. Suction the recircu-Check Valve lation phase From RWST is initiated. Containment Valve fails The redundant trains supply the Spray Pump to open when necessary heat removal capacity. Suction the recircu-Isolation lation phase Valve is initiated. Containment Valve fails Check valve prevents backflow to Spray Pump to close RHST. Suction when the Motor Oper- recircula-dted Valve tion phase From RWST is initiated. Containment Rupture of The redundant train (s) supply the Spray Pip- containment necessary heat removal capacity. ing spray piping 24
/ o- -
( > L) ' f Ud B&W, C-E, W-3S 1 of 3 Amendment 24 4/23/76
SWESSAR-P1 TABLE 6.2.2-2 (CONT) Component Malfunction Comments and Consequences Containment Fan fails to The redundant train (s) supply the Atmosphere start. necessary heat removal capacity. Recirculation Cooler Residual Tube or shell The redundant train (s) supply the Heat Removal rupture necessary heat removal capacity. Heat Exchan-ger Spray Noz- Spray nozzles The redundant train (s) supply the zles plugged necessary heat removal capacity. Engineered Screens The redundant train (s) supply the Safety Fea- clogged necessary heat removal capacity. tures Sump Automatic Failure of Redundant train actuates redundant Electric and one train equipment. Control Instrmoenta-tiun Trains @~ to Actuute Engineered Safety Features Equipment Motor Valve fails Redundant valve (in parallel) opens. Operated to open. Valves in Chemical Addi-tion Tonk (CAT) Discharge Line Motor Valve fails CAT drains dry. Check valves in g Operated to close eductor lines prevent excess air Valves in flow. Chemical Addi-tion Tank Dis-charge Line Check Valve Check valve Motor operated valves in CAT dis-in CAT Dis- tails to charge line and other check valve charge Line close close, isolating piping. B&W, C-E , W- 3 S 2 of 3 Amendment 24
,4/23/76
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SWESSAR-P1 TABLE 6.2.2-2 (CONT) Component Malfunction Comments and Consequences Low Level Lack of CAT drains dry. Check valves in Transmitter signal to eductor lines prevent backflow. on CAT close motor operated valves in CAT discharge line Low Level Sjurious Redundant parallel valve controlled Transmitter on signal to by redundant transmitter. CAT close motor operated valve 3 Containment Fails to One eductor loses motive flow. Spray Pump start Suf ficient NaOH to provide minimum sump pH is drawn by redundant train spray pump and eductor until CAT contents are discharged and low level isolates CAT. CAT Dis- Piping Even with the single active f ailure charge ruptures that results in the lowest flow rate Piping (after from the CAT, the CAT is isolated Eductors 24 hours) in less than 2 hours. Therefore sufficient NaOH will have been introduced , and the CAT piping can be manually isolated.
^*"" *"" "
B&W, C-E, W-3S 4/23/76
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SWESSAR-P1 TABLE 6.c.2-2 FAILURE ANAINSIS FOR CONTAINMENT HEAT REMOVAL SYSTEMS Component Malfunction Comments and Consequences Containment Pump casing The redundant train (s) supply the Spray Pump ruptures. necessary heat removal capacity. Containment Pump fails The redundant train (s) supply the Spray Pump to start. necessary heat removal capacity. Containment Valve fails The redundant train (s) supply the Spray Pump to open, necessary heat removal capacity. Discharge Motor Oper-ated Iso-lation Valve Containment Check valve The redundant train (s) supply the Spray Pump in pump dis- necessary heat removal capacity. Discharge charge line Check Iso- sticks closed. lation Valve Containment Valve fails Motor operated valve in suction line Spray Pump to close when from RWST closes. Suction the recircu-Check Valve lation phase From RWST is initiated. Containment Valve fails The redundant trains supply the Spray Pump to open when necessary heat removal capacity. Suction the recircu-Isolation lation phase Valve is initiated. Containment Valve fails Check valve prevents backflow to Spray Pump to close RWST. Suction when the Motor Oper- recirculu-ated Valve tion phase From RWST is initiated. Containment Rupture of The redundant train (s) supply the Spray Pip- containment necessary heat removal cupacity. ing spray piping Mot.or Opera- Vol"e fails Redundant valves open. ted Valves to open. @ W 1 of 2 Amendment 24 24 4/23/76
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{ .k [4 SWESSAR-P1 [< TABLE 6.2.2-2 (CONT) s
~
Malfunction Corments and Consequences { Cormonent V
- in CAT Dis-
[ charge Line Containment Fan fails to The redundant train (s) supply the hJ Atmosphere start. necessary heat removal capacity, Recirculation I' Cboler [ Residual 'nabe or shell The redundant train (s) supply the
; Heat Recoval rupture necessary heat removal capacity.
W Heat Exchan-U ger s-Spray Noz- Cpray nozzles The redundant train (s) supply the
.- zles plugged necessary heat removal capacity.
b
- f. Engineered Screcas The redundant train (s) supply the
[ Safety Fea- clogged necessa n heat removal capacity. c tures Sump y [. Automatic Failure of Redundant train actuates redundant F Electric and one train equipment. Control Instrumenta-
% tion Trains to Actuate L.
- f. Engineered
!- Safety Features si Equipment I
i: s Il V. c p I F f., k I~ V b 1; g , n F-' ' . e . . , w 2 of 2 Amendment 24 l24 e 4/2 3/7 f., l h* 't** - ( A. 5
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SWESSAR-P1 TABLE 6.2.2-3 NET POSITIVE SUCTION HEAD (NPSH) FOR ENGINEERED SAFETY FEATURES PDMPS Injection
- Recirculation
- NSSS Vendor Required Phase NPSH, Phase NPSH Pump NPSH, ft Available, ft Available, ft Westinghouse (W-41)
Containment Spray 15 44.9 18.1 High Head Safety Injection 15 42.4 19.0 Iow Head Safety Injection 15 41.8 18.3 Babcock & Wilcox (BSAR-205) Note 1 Note 2 Note 2 Containment Spray 15 44.9 18.1 High Pressure Injection 42 >42 (Note 4) Note 3 Low Pressure Injection 16 >16 18-28 32 Combustion Engineering Note 1 Note 2 Note 2 Containment Spray 15 44.9 18.1 High Pressure Safety Injection 12-17 39-44 16-21 low Pressure Safety Injection 15-25 41-51 18-28 Westinghouse (W-3S) ?;ote 1 Note 2 Note 2 Containment Spray 15 44.9 18.1 Charging Pump 28 >28 Note 3 Safety Injection 25 >25 Note 3 Low Head Safety 15 41.8 18.3 Injection /RHR NOTES:
- 1. The NPSH requirement for the high pressure safety injection pump (C-E) and low pressure safety in jection pump (C-E , W-3S, BSW) has been modified to reflect a deep well type pump.
The values shown are typical.
- 2. The NPSH available to the deep well type pump is a function of the specific pump length chosen. The NPSH available will be greater than that required by some engineering margin.
- 3. Adequate NPSH is available since the low pressure safety injoction pump supplies the suction of the pump.
- 4. HPI pump suction is automatically switched from the RWST to LPI pump discharge at an RWST level that is greater 32 than the NPSH requirements. , ,
G< + i.;,a.l
- Preliminary data 1 of 1 Amendment 32 5/11/77
SWESSAR-P1 TABLE 6.2.2-4 PARAMETERS USED FOR RADIOLOGICAL CONSEQUENCES OF RECIRCULATION LIQUID LEAKAGE
- 1. Before the incident, the reactor was operatir" at its maximum calculated rating, 4,100 MWt (BSW, C-E , W-41) or 3,636 MWt (W-3S) .
- 2. Fif ty percent of the equilibrium radioactive iodine inventory developed from maximum full power operation of the core is immediately released to the containment structure and may be acted upon by the containment spray system.
- 3. Noble gases released to the containment atmosphere are not affected by the containment sprays.
- 4. The iodine removal coefticient by the containment spray system for elemental iodine is 10 hr-2
- 5. The iodine concentration in the containment water volume of dpproximdtely 2 x 10' cc just prior to the start of the recirculation phase with normal ESF operating is:
I131 2.5 E 04 uCi/cc I132 3.0 E 04 uCi/cc I133 5.5 E 04 uCi/cc I134 4.2 E 04 uCi/cc I135 4.9 E 04 uCi/cc No significant addition to the iodine inventory occurs af ter this time.
- 6. The temperature of the water being circulated as a function of time is given in Fig. 6.2.2-2.
- 7. The potential recirculation loop leakage external to the containment begins at the initiation of recirculation and continues for the 30 day period.
The total leakage based on the discussion in Section 6.2.2.3 is given below: ECCS Containment Total Leakage Spray Leakage Leakage (cc/hr) (cc/hr) (cc/hr) Babcock-Wilcox 2,443 1,915 4,358 32 Combustion-Engineering 3,810 635 4,445 Westinghouse 41 10,830 953 11,783
~" Amendment 32 1 of 2 (', .,
5/11/77
SWESSAR-P1 TABLE 6.2.2-4 (CONT) Westinghouse 3S 7,820 635 8,455 17
- 8. The iodine partition factor for the leakage is 1.0.
- 9. Noble gases resulting from iodine decay in the sump have a partition factor of 1.
- 10. The annulus building exhaust is collected in the supplementary leak collection and release system (Section 6.2. 3.1) and filtered through high efficiency particulate air (HEPA) filters / charcoal adsorbers with an overall efficiency for iodine of 95 percent before release to the environment.
- 11. The atmospheric diffusion factor (CHI /Q) including the volumetric building wake correction factors is based on values given in Section 2.3.
- 12. No credit is taken for depletion of the effluent plume of radioactive isotopes due to deposition on the ground or for the radioactive decay of iodine in transit.
2 of 2 Amendment 17 9/30/75 i 0t is ,
SWESSAR-P1 TABLE 6.2.3.1-1 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM COMPONENT DESIGN AND PERFORMid'CE CHARACTERISTICS Design and oerforr.ance Component Characteristics
- 1. Filter Bank Capacity 15,000 cfm l8
- 2. Exhaust Fun Capacity 15,000 cfm l8 Head 15 in. W.G.
Motor 75 hp \8
- 3. Charcoal Filter Decay Heat Eemoval Fan capacity 200 cfm Head 3 in. W.G.
Motor 1/2 hp bb$ I t!'i c'O 1 of 1 Amendment 8 3/28/75
9 SWESSAR-P1 TABLE 5.2.1.1-2 CONTAIte'ENT LEAT PATE TEST DATA Tyre A Testing As W st ed Ret Unit No. Pass /Pai1 % Limit * % La Conments Dresden 1 ( 1) P 70 53 Failed initially due with >2,000% leakage due to or* n valve. litunboldt Bay 3 (1) P 10 8 Millstone 1 (1) P 37 28 Monticello (1) P 71 53 4 of 8 main st eam line isolation valves required repair to meet reactor operating criteria. Monticello (2) P 48 36 Tested July 1970. Pilgrim 1 (1) P 32 24 Fort Calhoun 1 (1) P 50 37 Initial test of containment. 8 Calvert Clif f s (3) P 31 23 Connecticut Yankee (4) P 23 17 1967 test. P 11 8 1970 test. Maine Yankee (5) P 29 22 Dresden 2 (6) P 8 10 James A. F it r Pat rick (7) P 47 35 11 . B. Robinson No. 2 (8) P 11 9 Pa t e st . P 39 29 Pt test rerun after finding two 6 in. valves not tight. H. B. Robinson No. 2 (8) Oyster Creek (9) P 64 48 Vermont Yankee (10) P 57 43 1/2= pipe plug required, excess leakage detected within first 1/2 hour. Connecticut Yankee (11) P 45 34 Average 29% C.
- Limit for Tyie A Test is 0.75 La 1 of 2 Amndment 8 3/28/75
SWESS AR-P 1 TABLE 6.2.3.1-2 (CIUT) (1) "Nuclea r Power Plant Opera t ing Experience During 1973, Containment Leak Rate and Surveillance Tests," Document No. 00E-ES-004, U.S. Atomic Energy Commission, Of fice of Operations Evaluations, Decemter 1974. (2) "Integr at ed Primary Containwnt Leak Rate Test Perf ormed July 1970" Northern States Power Company, Decemler 1970. (3) "Feactor Containment Buildinq, In t egrat e*1 I4=a k Ra t e Test ," Ca lver t Cliffs Nuclear Power Plant Unit No. 1, January 11, 1974 (4) "Conne c ti cut Yankee Containment O pera t ion and Testing May 1967 through April 1970," Edmund C. Tarnuzzer, Novmte r 30, 1970. (5) "Knine Yankee Class A Integrated Onntairner:t Leak Rat e Test ," Pichard P. LaPhett e, David P. Boynton, September 8, 1972. 8 (6) "Sumary Technical Petort of Dresden Nuclear Power Station Unit 2 Primary Containment Leak Pate Test," Comn>nwealth Edison Cmpany, Decvamler 1970. (7) "Peact or Cont ainment Building Integrated Leak Rat e Test Final Report, August 1974," James A. FitzPatrick Nuclear Power Plant, Augu st 1974. (8) "Addendtsu I to Preogwrational Integrated Leakage Rate and Sensitive Leakage Data Test of Peact or Containment Building," H. B. Robinson Unit tb. 2, July 1970. (9) " Primary Cont ainment Ie=akage Rate Test, Kirch 1969," Oyster Creek Nuclear Generating Station, D. A. Fors, R. J. Toole, January 1971. (10) " Primary Cont ainment Leakage Fate Testing, 1972-1974," Vermont Yankee, R . W. Burke, C. T. Tarnuzzer. (11) "Connec ticut Yankee Atanic Power Company Containment Test nq June 1970 Through June 1972," Stanley Flming, Sept. 7, 1972. CN r ,x r3 2 of 2 Arendment 8 3/?R/75
SWESSAR-P1 TABLE 6.2.3.1-3 CONTAI?P'E?TP TYPES B C C TEST DATA Ref Wst* As Tested As Cor rect ed Unit t h Type Pass / Fail 1 Limit ** Pass / Fail 1 Limit ** % La Corynent utg Fock Point (1) L P 70 - - n2 Greatest leakace from supply vent valve. Oyater Creek 1 (1) L F 1,000 P 1 1 Main steam line isola-tion valves and air-lock exceeded limits. Dresden 3 (1) L F - P 16 10 - Vermont Yankee (1) L F - P 45 27 3 main steam line iso- . lation valves exceeded limits. ' (h.ta d-Cit ies 1 (1) L P 82 - - 49 - Quad-Cities 2 (1) L F 540 P 35 21 - San Onofre 1 (1) L F 138 P 10 6 Containment isolation valves exceeded limits. 3 Palisades (1) L P 66 - - 40 - Connecticut Yankee (2) L P 48 - - 29 - James A. Fitzpatrick (3) L F 94 - - 56 - Averace 28% La
- L represents a total of B E C t est data.
** Limit for TyIe D & C test is 0.6 La.
(1) " Nuclear Power Plant Ot erating Ex}erience During 1973, Containment Leak Rat e and Surveillance Tests," Document No. OOE-ES-004, U.S. Atomic Diergy Commissim , Office of m Operations Evaluations, December 1974. (2) " Connecticut Yankee Cont ainwnt Operation and Testing May 1967 through April 1970," Edmund C. Tarnuzzer, ?bvember 30, 1970. (3) " Reactor Containwnt Building Integrated Leak Rate Test Fi nal R ep>rt , Augu st 1974," James ._ A. Fitzpatzick Nuclear Power Plant, August 1974. - J _J1 1 of 1 Amendnent 8 3/28/75
SWESSAR-P1 TABLE 6.2.3.1-4 DESIGN PARAMETERS FOR AREAS SERVED BY SLCRS Estimated Free Volume Leakage Fan Exhaust x106 ft3 Rates * (cfm) Rates (cfm) Annulus Bldg 3.30 6,000 9,000 16 Fuel Bldg 0.93 2,000 3,000 Main Steam and Feedwater 0.16 1,000 1,500 Valve Enclosure Electrical Tunnels 0.27 1,000 1.500 in The Annulus Bldg Totals 4.66 10,000 15,000
- At 0.25 in. H O differential pressure 2
bUk . U 1 of 1 Amendment 16 8/29/75
O SWESSAR-P1 TABLE 6.2.3.1-5 POTENTI AL 11AK AGF THAT CAN BYPASS AREAS SERVED BY SIfRS Fluid Normally Approx Valve Type /ATME Sect ori XI IWV i SAR Safety in Line Irak Speci t i( at ion System Section Bypass Line Cla s:1 Syst em S i ?.e, [i n .1 Inside out ::1de Reactor Plant Com- 9.2.2 NNS supply UNS Water 6 Check /45 SCF/ day AOV but t et t ly/6 SCF/da) ponent Cooling Water System NNS return NNS Water 6 AUV buttertly/45 AOV bett er f ly/45 SCF/ day SCF/ day Dmineralized Water 4.2.3 Supply NNS Water 2 Check /15 SCF/ day AOV qate/15 SCF/ day Primary Grade Water 9.2.7 Supply NNS Water 4 Check /30 SCF/ day AOV qate/30 SCF/ day Instrument and 9.3.1 Instrument air NNS Gas 2 3 top check /15 AOV qat e/15 SCP/ day Service Air supply SCF/ day Service air NNS Gus 2 Stop check /15 L.C. gat e/15 SCF/ day supply SCF/ day Water Fire Pro- 9.5.1 Supply NNS Nater 4 Check /30 SCF/ day ADV qate/30 SCF/ day tection Supply NNS Water 4 Check /30 SCF/ day AOV oute/10 SCF/ day Reactor Plant Gas 9.5.8 Inw pressure N2 NNS Gas 1 Check /7.5 SCF/ day AOV qat e/ 7.5 SCP/da y Supply supply 20 liigh pressure N2 NNS Gas 1 Caeck/7.5 SCF/ day AOV qate/7.5 SCF/ day supply containment Atmos- 12.2.4.1 Suction NNS Gas 1 MOV gate /7.5 SCF/ MOV qate/7.5 SCF/ day phere Monitoring day Discharge NNS Gas 1 Check /7.5 SCF/ day MOV qat e/7.5 SCF / day Containment Leakage 6.2.6 Open pressure NNS Gas 3/8 SOV Globe /2.8 SCF SOV globe /2.8 SCF/ day O Monitoring taps / day J1. Scaled pressure NNS Gas 3/8 SOV Glote/2.8 SCP SOV olote/2.H SCF taps / day / day C Total Max Leakage 261 SCF/ day N 1 of 2 Amenchnen t 20 1/23/76
SWLaaAF-P1 TABLL 6.2.3.1-5 (Corn ) Sntilest contaanment voluaw- (RESAL-4S) = 3.06 x 106 cu tt Test pressure (1< E SAR- E ) = '10.1 psi ; (sti .R psia) Owntainment total volume = (3.06 y 106) 54.8 = 1.14 x 10F 1;C}' N 14.7 Total l eak aq e = 261 SCP/ day = .0023 1 of t ot a l volurw /etty NNS = Non-Stu c le.t r saf et y ALN = Air operat ed valve NW = Notor operated valve SCP = Standard cubic toot NOTE: The main steam, f e<M! water, and st eam generator blowdown systems are not considered p.trt of bypass leakago because these systems are not e xtosed t o the con t.t i nmen t atmosphere. Following a DHA, the steam generator semndary pressure will be higher than the steam generator primary pressure. The pressure ditterence will decrease with time. As pressur e equilibrium is approached, t he .nuxiliary feedwater 70 system is available to atintai , the steam generator tu1* s in a i loodesi condition, thus precluding the twissibility of bypass leakage via t hese- systems. r~. C C 'I 2 ot 2 Arrenitnen t 20 CD 1/2 V76
SWESS AR-P1 TABLE 6. 2. 3. 2-1 NOMENCLATURE USED FOR EQUATIONS IN SECTION 6. 2. 3. 2 Symbol Nomenclature Ad Surf ce area per drop, cm2 Cd Amount of iodine initial in containment atmosphere, curies ce Amount of iodine at time t in containment atmosphere curies d Mean surf ace diameter, cm DL Diffusivity of iodine in water, cm2/sec Dy Diffusivity of iodine in steam-air mixture, cm2/sec F Spray flow rate, cm3/sec h Header height, em H Liquid-to-gas iodine partition factor Kg Liquid film mas s transfer coef ficient, cm/sec T Iodine removal coef ficient, se c- 1 1 conservative total iodine removal coef ficient, hr-1 s gv viscosity of steam-air mixture, gm/cm/ sac Re Reynolds number 'L Density of water at temperature T, gm/cm3 C v Density of steam-air mixture, g m/cm3 S Schmidt number c S d Total surf ace area of spray drops in containment atmosphere, cm2 t Time, see tf Average residence time of drop in atmosphere, sec 1 of 2 (- c d. VI : I
SWESSAR-P1 TABLE 6. 2. 3. 2- 1 (CONT) 1 Symbol Nomenclature-T Lowest temperature of steam-air mixture #during time period, degrees C V Terminal velocity of drop, cm/s ec Vc Containment structure volume, cm3 Vd Volume per drop, cm3 VD Overall. deposition velocity, cm/sec Vg Gas film deposition velocity, cm/sec Vs Volume of water in the containment, cm3 2 or 2 () ' , ' ovJ
SWESSAR-P1 TABLE 6.2.3.2-2 IODINE REMOVAL CDEFFICIEtrr5 FOR CDNTAlleENT SPRAY First Second Third Fourth Perio_d Period Period Perl(wi Ohantity Unite A* E b* M M P* M R.* Tims Sec 70-600 600-1200 1200-2400 2400-3600 Teperature C 118.3 110.6 96.7 92.2 Spa-y pH 9.5 9.5 9.5 9.5 Cases I and II u t g/ W s 0.2375 0.2554 0.2940 0.3087
#, g/m/s 0.000174 0.000176 0.000180 0.000181 D s 0.04276 0.04785 0.05726 0.06023 ca /s S 1.7352 1.7368 1.7368 1.7363
,L g/cz
- 0.002339 0.002112 0.001806 0.001729
,V' g/cm8 0.9449 0.9510 0.9612 0.9643 d micr m 1000 1000 1000 1000 1000 1000 1000 1000 F an*/s 1.11E05 3.32EOS 1.11E05 3.32E05 1.11E05 3.32E05 1.11E05 3.32E05 h em 6344 5671 63e4 5671 6344 5671 6344 5671 V, can/s 6.950 6.950 7.535 7.535 8.567 8.567 8.881 8.881 V cm/s 291. 291. 303. 303. 322. 322. 328. 328.
vc ans 9.49E10 9.49E10 9.49E10 9.49E10 9.49E10 9.49E10 9.49E10 9.49510 f_ase I 22 A (header) sec-8 0.0106 0.0284 0.0110 0.0296 0.0118 0.0316 0.0120 0.0323 h (ccanbined) sec-a 0.0390 0.0406 0.0434 0.0443 A (corabined) hr-s 140.40 146.16 156.24 159.48 Case II Do m8/sec 0.6429E-04 0.5861E-04 0.4905E-04 0.4616E-04 C,, Kg on/sec 0.00423 0.00423 0.00386 0.00386 0.00323 0.00323 0.00304 0.00304 cp
~
H 8000. 8000 8000. 8000. 8000. 8000. 8000 8000. V cm/sec 5.766 5.766 6.056 6.056 6.433 6.433 6.504 6.504
?> A (header) sec-8 0.0088 0.0236 0.0089 0.0238 0.0089 0.0238 0.0008 0.0236 A (ccanbined) see-a 0.0324 0.0327 0.0326 0.0324 h (combined) hr-8 116.64 117.72 117.36 116.64 g7 Case III (3 A tembined) see-8 0.009818 0.009909 0.009879 0.009818
_. A (corribined) hr-a 35.345 35.673 35.564 35.345
*A = Small diameter headet B = Large diamet er header W 1 of 1 Amendment 22 3/17/76
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SWESSAR-P1 TABLE 6.2.3.2-3 CONTAINMENT SPRAY SYSTEM FLOW RATES DURING AN ACCIDENT Flow .qrral from ECCS from RWST fapm) CAT FWST CSS Min ECCS 9,000 Min CSS 252 6,748 7,000 Max CSS 307 10,193 10,500 Max ECCS 13,500 22 Min CSS 323 6,677 7,000 Max CSS 378 10,122 10,500 RWST = Refueling Water Storage Tank CAT = Chemical Addition Tank CSS = Cbntainment Spray System W 1 of 1 Amendment 22 3/17/76 [,, [. d
. { 10 -lOJ
SWESSAR-P1 TABLE 6.2.3.2-4 CHEMICAL COMPOSITION OF THE CON'1,d!OENT SPEAY AND SUMP SOLUTIONS WITH 13.4 WEIGHT PERCENT NaOH IN THE CAT NaOH Boron Iniection Phase (Wt-Percent) (ptwn) pH (1) Minimun ECCS Minimum CSS 0.546 2,398 9.6 Maximum CSS 0.445 2,417 9.3 Maximun ECCS 22 Minimum CSS 0.700 2,369 10.0 Maximum CSS 0.546 2,398 9.6 Recirculation Phase Sump 0.220 2,350 8.5 (1) pH at 77 F W 1 of 1 Amendment 22 3/17/76 f! D . . ()U l t UVU
SWESSAR-P1 TABLE 6.2.3.2-5 CHIMICAL COMPOSITION OF THE CONTAINMENT SPRAY A2 SUMP SOLUTIONS WITH 14.4 WEIGHT PEPCENT NaOH IN TID: CAT NaOH Doron Iniection Phase (Wt Percent) jpy_ra1, pHfa> ttinimum ECCS Minimum CSS 0.588 2398 9.7 Maximum CSS 0.478 2417 9.4 Maximum ECCS 22 Minimum CSS 0.751 2369 10.I Maximum CSS 0.588 2398 9.7 Recirculation Phase Sump 0.238 2350 8.6 (8)pH at 77 F W l of 1 Amendment 22 3/17/76
, , , r 6 -'
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ShESSAR-P1 TABLE 6.2.3.2-3 CCNTAINMENT SPRAY SYSTEM FLOW RATES DURING AN ACCIDENT Flow (grn) from ECCS from RWST forn) CAT RWST CSS Min ECCS 5,268 24 Min CSS 40 3,460 3,500 Max CSS 80 6,920 7,000 Max ECCS 11,256 Min CSS 40 3,460 3,500 Max CSS 80 6,920 7,000 RWST = Refueling Water Storage Tank CAT = Chemical Addition Tank CSS = Contairenent Spray System W-3S 1 of 1 Amendment 24 4/23/76 e onO Obt bdd
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SNESSAR-P1 TABLE 6.2.3.2-3 CDtTIAI184EFF SPRAY SYSTEM FIDW RATES DURING AN ACTID1DTP Flow (qvm) from ECCS frca RMST (qm) CAT RMST CSS Min ECCS 6,%0 Min CSS 60 3,440 3,50 0 Max CSS 120 6,880 7,000 3 Max ECCS 13,120 Min CSS 60 3,440 3,50 0 Max CSS 120 6,880 7,00' RWrT = Refueling water Storage Tank CAT = Chemical Addition Tank CSS = Contain -_t Spray System B&W 1 of 1 Amatviment 34 7/22/77 {3 ( } ,
SWESSAR-P1 TABLE 6.2.3.2-4 QtDtICAL CtmPuGITION Of THE CDefTAINMENT SPRAY AND STMP 501M"IOII Phnee EaQIfwle/11tegl Mrtm otole/11t erl DH888 h Injectlam 0.134 0.209 9.6 nacir m iation tsump water) Kiniman ECCS Minimon CSS 0.04C 0.205 8.1 At switetat ser 0.052 0.204 8.5 16 minutes after switdaower ja Maximum (23 0.059 0.204 8.6 At switctiover Maximann ECCS Minisman CSS 0.024 0.206 7.7 0.053 At switetsover 0.206 8.5 38 minutma after suitetarver Maximua GS 0.040 0.205 8.1 At. witdaover 0.052 0.204 8.5 0 minutes after switelnover 888pH at 72 F e
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SWESSAR-P1 TABLE 6.2.3.2-3 CONTAINMENT SPRAY SYSTEM FIDW RATES DURING JJ ACCIDE?TT Flow (qte) from ECCS from RWST fcmm) CAT RKST CSS Min ECCS 6,100 Min CSS 115 3,385 24 Max CSS 3,500 230 6,770 7,000 Max ECCS 12,200 Min CSS 115 3,385 Max CSS 3,500 230 6,770 7,000 RWST = Refueling Water Storage Tank CAT = Chemical Addition Tank CSS = Containment Spray System C-E 1 of 1 Amendment 24 4/23/76 I 3 {T t- Ul ((f () :'I
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SWESSAR-P1 TABLE 6.2.4-1 RAL DESIGN CRITERIA 10. 54 (SECTION 3154) Valve Closure Tenp Penetration Isolation Valves Actuation Time (Cold Closed Type Inside Outside Signal (See) Fluid < 250 F) System (Fig. 6.2.4-1) Refer to NSSS SAR Globe (A) Globe (A) CIA 10 Gas Hot Yes A Globe (A) Globe (A) CIA 10 Liquid Hot Yes A Globe (A) Globe (A) CIA 10 Liquid Hot Yes A ' Globc(A) Globe (A) CIA 10 Liquid Hot Yes A 35 Refer to ICSS SAR Refer to NSSS SAR None Butterfly (M) Recire. NA Liquid Cold No B Check Butterfly (M) CDA NA Liquid Cold No E Gate (M) Gate (M CIA NA Gas Cold No F Check Gate (M CIA NA Gas Cold No G Globe (S) Globe (S) CIA 10 Gas Cold Yes O Globe (S) Globe (S) CIA 10 Gas Cold No 0 35 Amendment 35 10/6/77 of 4 [u V ,i t, , r n /, V j 'I
~,
PFING PEI.MPATICIG PER G: No. of Penetrations Per GDC No. 55 Penetrations Line Status (Section 3.1.55) W-3S W-41 B&W C-E Nomal Shutdown DBA Chemical and Volune Control f stem (Section 9 3.4) Sampling System (Section 9 3 2) From Pressurizer Vapor Space 1 1 1 1 Either Closed Closed From Pressurizer Liquid Space 1 1 1 1 Either Closed Closed From RCS 1 1 1 1 Either Closed Closed From RHR 0 1 0 0 Either Open Closed Residual Heat Removal System (Section 6.3) Penetrations Per GDC No. 56 (Section 3.1 56) Emergency Core Cooling System (Section 6.3 1) Containment Spray System (Section6.2.2) Suction Line (from ESF Sump) 2 3 2 2 Closed Closed Open Discharge Line 2 3 2 2 Closed Closed Open Combustible Gas Control System (Section6.2.5) Recombiner Suction Line 2 2 2 2 Closed Closed Closed Recombiner Discharge Line 2 2 2 2 Closed Closed Closed Containment Leakage Mordtoring System (Section 6.2.6) Sealed Pressure Taps .L 1 1 1 Either Either Closed Open Pressure Taps 1 1 1 1 Either Either Closed r, r. ' 7 U Uit U/J
SSAR-P1 !.4-1 (COITI'D) Valve Closure Temp Penetration Isolation Valves Actuation Time (Cold Closed Type Inside Out side Signal (Sec) Fluid t 250 F) System (Fig. 6.2.4-1) Gate Gate L.C. NA Liquid Cold No H Gate Gate L.C. NA Liquid Cold No H Gate Gate L.C. NA Liquid Cold No H Gat e Gate L.C. NA Liquid Cold No H Check Butterfly (A) CIA 10 Liquid Cold Yes(l) K 35 Butterfly (A) Butterfly (A) CIA 10 Liquid Cold Yes(l) J Gate Gat e L.C. NA Liquid Cold Yes(l) H Check Gate (A) CIA 10 Liquid Cold Yes(l) K Check Gate (A) CIB 10 Liquid Cold Yes(1) M Gate (A) Gate (A) CIB 10 Liquid Cold Yes(l) L Check Butterfly (A) CIA 10 Liquid Cold Yes(l) K 35 Butterfly (A) Butterfly (A) CIA 10 Liquid Cold Yes(l) J ' of 4
, n r,
/*f #
Amendment 35 10/6/77
1 ST TABLE 6 No. of Penetrations Line Status W-3S W-41 B&W C-E Nomal Shutdown DBA Fuel Pool Cooling and Purifice. tion System (Section 91.3) Refueling Cavity Purification Supply Line 1 1 1 1 Closed Open Closed Refuelin6 Cavity Purification Return Line 1 1 1 1 Closed Open Closed In-containment Fuel Storage Cooling Supply Line 0 1 0 0 Closed Open Closed In-contain=ent Fuel Storage Cooling Return Line 0 1 0 0 Closed Open Closed Reactcr Plant Component Cooling Water System (Section 9 2.2) N1;S Supply Line 0 0 0 1 Open Open Closed I:';S Peturn Line 0 0 0 1 Open Open Closed Demineralized Water Makeup Systen (Section 9 2.3) Supply Line to Containment 1 1 1 1 Closed Open Closed Primary Grade Water System (Sectier. 9 2.7) Supply Line to Containment 1 1 1 1 Open Open Closed Chilled Water System (Section 9 2.0) Supply 1 1 1 1 Open Open Closed Return 1 1 1 1 Open Open Closed CRIti Motor Cooling System (Section 9.?.10) Supply 0 0 1 0 Open Open Closed Return 0 0 1 O Open Open Closed ((; Ly .,t ['u,I D ~'
SSAR-P1 4-1 (COIE 'D) Valve Closure Tenp Penet ration Isolation Valves Actuation Time (Cold Closed Type Inside Outside Signal (Sec) Fluid <250 F) Syst em (Fig. 6.2.4-1) Stop Check Gate (A) CIA 10 Gas Cold Yes(l) K Stop Check Gate L.C. HA Gas Cold Yes(l) N Gate (A) Gate (A) CIA 10 Liquid Cold Yes(l) J Cate(A) Gate (A) CIA 10 Liquid Cold Yes(1) J Gate (A) Gate (A) CIA 10 Gas Cold Yes(l) J Gat e(A) Gate (A) CIA 10 Gas Cold Yes(l) J Butterfly (A) Butterfly (A) CIA NA Gas Cold No C Butterfly (A) Butterfly (A) CIA IR Gas Cold No C Check Gate (A) CIA 10 Liquid Cold No K 35 CIA 10 Gas Cold Yes(l) K Check Gate (A) Gate (M) Gate (M) CIA 10 Gas Cold No I 35 Check Gate (M) CIA 10 Gas Cold No D 3 of 4 [' [, ;' '
" ' ."' Attachment 35 10/6/7/
S', T/GLE 6 11o. of Penetrations Line Status W-3S W-41 B&W C-E I;omal Shutdown DBA Compressed Air System (Section 9 3.1) Instrument Air Supply Line 1 1 1 1 Closed Open Closed Service Air Supply Line 1 1 1 1 Closed Open Closed Vent and Drain System (Section 9 3.3) Containment Su=p Pump Discharge 1 1 1 1 Either Either Closed Reactor Coolant Drain Tank Pu=p Discharge 1 1 1 1 Either Either Closed Reactor Coolant Drain Tank Vent Header (C-E, B&') - - 1 1 Closed Closed Closed Reactor Coolant Drain Tank Vent Header (W-41 and W-33) 1 1 - - Open Open Closed Containment Purge Air Syste= (Section 9.4.5 2) Supply 1 1 1 1 Closed Open Closed Exhaust 1 1 1 1 Closed Open Closed Water Fire Protection Syst n (Section 9 5.1) Supply 3 3 3 3 Open Open Closed Re. actor Plant Gas Supply System (Section 9 5.8) N2 Supply to Containment 2 2 2 2 Either Either Closed Containment Atmosphere Monitoring System (Section 12.2.4) Blower Suction Line 1 1 1 1 Open Open Closed Blower Discharge Line 1 1 1 1 Open Open Closed
, r:c,,
U v, <- V, /
AR-P1 -1 (COIff'D) Valve Closure Tc:np Penet rat ion Isolation Valves Actuation T ir.e (Cold Closed Type Inside Out side Signal (Sec) Fluid < 250 F) System (Fig. 6.2.4-1) I;SSS SAR None Butterfly (M) None NA Liquid Cold Yes P None Butterfly (M) None NA Liquid Cold Yes P 35 None Globe (A) CIA 10 Liquid Cold Yes Q None Globe (n) CIA 10 Liquid Hot Yes Q Uone Ball, Gate, SLI 5 Gas Hot Yes R or Globe None Gate (A) WI 5 W) Liquid Hot Yes Q 15 B&W) 20 C-E) None Globe (A) AFAS 10 Liquid Hot Yes Q None Gate (M) None NA Liquid Cold Yes P
, c
=n eu. v Of k Arendment 35 10/6/77
SWE TABLE 6.2 No. of Penetrations Line Status W-3S W-41 B&W C-E :;omal Shutdown DBA Penetrations Per GDC No. 57 (Section 3 1.57) Emergency Core Cooling Systcc (Section 6.3.1) Refer Reactor Plant Component Cooling Water System (Section 9.2.2) SC3 Supply Line 2 3 2 2 Open Open Open SC3 Return Line 2 3 2 2 Open Open Open Sampling System (Section 9.3 2) Pron Accumu]ators 1 1 O O Either Either Closed From Stea Generators 0 0 1 O Either Either Closed Main Steam System (Section 10.3) Main Steam Lines 4 4 4 4 Open Closed Closed Feedwater Syste= (Section 10.4.7) Feedwater Lines 4 4 2 4 Open Closed Closed Steam Generator Blowdown System (Section 10.4.8) Stem: Generator Blowdown 4 4 0 2 Open Closed Closed Auxiliary Feedwater System (Section10.4.10) Auxiliary Feedwater Lines (W) 4 h 2 2 Open Open Open (1) This is a closed, NNS, nonseismic system inside the containment structure. It is assuned that it becoces an open systen following a DBA, tsking GDC56 applicable. A - Air operated M - Motor operated S - Solenoid operated LC - Locked closed D 0l
cerEC AR-F1 TA312 6.2.4-P I!CTIQ'1:CATID:3 I E2.TTPATICIC I O PJE"'A?Ff GUHE 1.11 (C EfI IC ! M .1- 1.11 ) Va.lve No. of Clcsure l'one t ra- Line Status Isolation Valvas A et wition Tiw Temp Closed tions Normal S h u t.tm LBA Ins 11e O'It s i ' . Cignal (Sac) Fluit { Cold P50 F) Srt es Conttinaant Pressure Monitoring System TS'ection 7. 3.3 9) Open Pressure Taps (Isolation valve downstreu 4 4-Open 4-Open 49en Nane 31ote L.O. NA G as Coli No of Pressure Transmitters) i L.O. - la ked c;*n UN C._3 I'd I of 1 Amerkhnen t 17 9/30/75
SWESSAR-P1 TABLE 6.2.5-1 COMBUSTIBLE GAS CONTROL SYSTEM DESIGN PARAMETERS Hydrogen Recombiners Number 2 Minimum capacity 50 scfm Hydrogen recombination efficiency 99% Design internal pressure 24 psig }3 Hydrogen Analyzers Number 2 Required accuracy 10.1 volume % H2 Design internal pressure 24 psig 13 Dilution Air Subsystem Minimum capacity Dilution air supply blower 50 scfm Dilution air exhaust blower 50 scfm Design internal pressure 24 psig 13
<, 4 n --
(i v r i JJ 1 of 1 Amendment 13 6/30/75
SWESSAR-P1 TABLE 6.2.5-3 PARAMETERS USED FOR POST-DBA HYDROGEN PUPGE ANALYSIS
- 1. Before the incident, the reactor was operating at its maximum calculated rating, 4,100 MWt (W-41, C-E), 3,876 MWt (B &W) , or 3,636 MWt (W-3S) .
- 2. Twenty-five percent of the equilibrium radioactive iodine inventory developed trom maximum full power operation of the core is immediately available for leakage from the containment structure. Ninety-one percent of this 25 percent is in the form of elemental iodine, 5 percent in the form of particulate iodine, and 4 percent jn the form of organic iodides.
- 3. One-hundred percent of the equilibrium radioactive noble gas inventory developed f rom maximum full power operation of the core is immediately available for leakage trom the containment structure.
- 4. The fission product inventory available for leakage from the reactor containment structure af ter the accident is shown in Table 15.1.13-1.
- 5. The iodine removal coetticient by the containment spray system for elemental iodine is 10 hr-1 until elemental iodine has been reduced to 1 percent of original inventory, then no credit is taken for adcitional reduction in elemental iodine.
No credit is taken ter reduction in particulate and organic iocine by the containment spray system.
- 6. The containment structure leaks at a rate of 0.2 percent per aay for the first 24 hr and 0.1 percent per day until the start of purge.
7A RtSAR-41 NSSS Assumption 26 The purge intervals and correspondina purge rates are as follows: l ime/ Hours Purae Pate (scfm) 0- 1008 (No Purge) 1008 - 1016 18.9 1016 - 1032 18.9 1032 - 1104 18.6 1104 - 1728 15.8 1 of 2 Amendment 26 bO: } ]i; 6/2/76
SWESSAR-P1 TABLE 6.2.5-3 (CONT) 28 7B C-E and BSW Assumption As the limit ng case, purge is initiated after 24 days at a constant rate c i 50 scim. This purge rate is more than the minimum purge rate required to maintain the hydrogen levels below the m&ximum allowable concentrations, Fig. 6.2.5-3. 7C RESAR-3S Assumption Purge is initiated after 35 days at a constant rate oi 28 2 2. 5 scim. This purge rate is sufficient to maintain the hydrogen levels below the maximum allowable concentrations, Fig. 6.2.5-3.
- 8. The purge is discharged through the process vent portion ot the radioactive gaseous waste system and filtered through high efficiency particulate air (HEPA) filters / charcoal adsorbers with an overall e f f iciency of 95 percent before release to the environment.
- 9. No credit is taken for depletion of the effluent plume of radioactive isotopes due to deposition on the ground or for the radioactive decay of iodine in transit.
- 10. The atmospheric diffusion factor (CHI /Q) including the volumetric building wake correction factors is based on values given in Section 2.3.
2 of 2 Amendment 28 8/6/76 b6n - 'O., tsi
5 7
/
0 3
/
\f R 6
L _ O 3 0l 8
- I l gDl Ll g ! l i l i l i 80 1
T C 1 T N 1 A E E S M e 3 5 T T R AL D D F - _ F N 2 E 4 FI M 2 8 ' 8 8 O N T A 8 1 4
* 'L S E 5
3l 1 lI l I" I I D Ls I l l I ! i i l
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CIRCUM FEREN TI AL - f s BOUNDARIES 6 ~*
' s.:
. 4(8) _
i s 1(5) N c .e
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NOTE THE FIRST NUMBER REPRESENTS THE NODE BELOW T HE NOZZLE CENTERLINE. NODE NUMBERS FIG 6*2
- 1-1 A ABOVE THE CENTERLINE ARE SHOWN IN PARENTHESIS. REACTOR CAVITY NODALIZATION PLAN VIEW THROUGH CENTERLINE OF NOZZLES PWR REFERENCE PLANT SAFETY ANALYSIS REPORT SWESSAR-P1 3 q7 7 ,
dv i i U/ AMENDMENT 26 6/2/76
NODE AXIAL
^
RPV FLAAGE
~
^
P ,' . f ,
/ f
/ lt:
/ \
/ : NODES 5-8 7
i RDV NODE AXtAL j/ f BOUNDARY j [ - T. RPV NOZZLE
/
i hSUPPORTS A- o'
/ NODES 1-4 _,-
RPV g /h f
/
'/(([//// ,- NODE AXIAL BOUNDARY
/ / /
/ / RV SUPPORT /
/ / TANK y
/ ,
/ *> " .
/ 2 d.
INSULATION (TYP) NOTES
- 1. THE VENT AREAS ARE AS SHOWN ON TABLE 6.2.1-13
- 2. THE VOLUMES ARE SHOWN ON FIG. 6.2.1-1
- 3. SUPPORTED LEG SHOWN. NONSUPPORTED LEG SIMILAR.
FIG. 6.2.1 - 1B REACTOR CAVITY NODALIZATION SECTION VIEW THROUGH CENTERLINE OF NOZZLE PWR REFERENCE PLANT SAFETY AN ALYSIS REPORT SWESSAR-P1
;nO W i' b ' ! V AMENDMENT 27 6/30/76
NODE 4 740 CF 10 , NODE 3 NODE 7 2,900 CF LOOP 3 OPER ATIN G ,g 4 7,000 CF FLOM - - - -- - - - - 10 *
<r NODE 2 :
\
e- 17,300 C F NODE 3 C10 d b _ _ _ '_ _ _ _ 9* TATE LOOP 2 5 8,000 C F
- 10 NODEI # : : ,,
10 16,500 CF a, D 5 16,0 00 CF NODE 9 LOOP l [ l 56,400 CF NOOE 6 6,700 CF - 10 to M NODE RECEIVES 8 LOWDOWN NOOf Me DE SCRIPT 80M i SROEEN LOOP STE AM G ENERATOR CCMPARTM ENT BELOW OPERAtlMG FLOOR ISELOW SR ATEI E SROKEN LOOP STEAM SENERATOR COMPAR T MENT BELOW OPER ATING FLOOR t ABOVE OR ATE) 3 BROKEN LOOP STE AM GE NER ATOR SNIELD W ALL ASOV E OPER ATING F LOOR TO E L. 41*-9* 4 BEOR EN LOOP ST E A M W ENE R ATOR SNIEL D W ALL ABOVE OPER Af f MG FLOOR FR OM EL.48'- 9"TO E L 4S'- 3*1 S IM-CORE sNSTRUMENTATION Dn eve ROOM F I G. 6 2.1 - 2
. a v0LuMe sELOW sRouEN LOCA STE AM eENER ATOR SUPPORT PEOESTAL iN TNE eNoutM Loop PAR AMETERS FOR AN ALYSIS OF STE AM GENER ATOR STE AM eENERATOR COMPARTMENT COMPARTMENT PRESSURE DIFFERENTI ALS T INTACT STE AM OE'sEM e ATOR COMPARTMENT ILOOP 3) PWR REFERt- CE PL ANT
$ INTACT STE AM GENER ATO8t COMPARTMENT (LOOP f l S A F E T V AN A .Y SIS R E PORT
,. . 9 INTACT STE AM $ENER ATOR COMPARTM ENT tLOOP f l SWESSAR-Pi 10 g utu CONTAI NM ENT t v0LuME 3,197,T SO FT31 E AMENOMENT IE 8 / te / TS
NODE 5# ' 5380 FT 3 N SPR AY LINE DER POSSIBLE IN THIS NODE W K SURGE LINE DER POSSIBLE
. r; ODE 4 IN THIS NODE 5605 FT 3 NODE 6 NODE 3 3,38 3,752 F T 3 8490 FT3
/"3 NODE M 523 FT 3 NODE I" 16,250 FT 3 NODE No D ESCRIPTION I PRESSURIZER RELIEF TANK COMPARTMENT 2 PRESSURIZER SUPPORT SKIRT 3 PRESSURIZER C U BICLE (LOWER P O RT IO N )
4 PRESSURIZER CUBICLE (U PPER PORTION VO LUME BELOW OR ATING) 5 PRESSURIZER CUBICLE (UPPER PORT:ON VOLUME ABOVE GR ATING) 6 BULK CONT AINMENT F I G . 6. 2.1 - 3 PAR AMETERS FOR AN ALYSIS OF PRESSURIZER CUBICLE PRESSURE DIFFERENTI ALS PW R REFERENCE PLANT S AFETY AN ALYSIS REPORT , , m SWESS AR- Pl s s , it U E AMENDMENT 96 8/29 / 75
5 7
/
6 1
/
T 6 E R 2 0 L E T O 1 i 1 CS NPE T N I A B N L R E M UO D
- C P RP S SEE I S E N
REDC Y V 9 ER EEN L! E A 0 4 Z ENR AP I N-
- R RI E AR
. UU L F
EYA 2 S SY RTS 6, SE SA ES E RR FE G RRPWAW 8 I
' - F P P S P SS 0
7 0
)
S D N O 6 C E 0 S (
)
5 ) ) 3 T E 4 N D E E D E O D 5 D N O O ( N 0 I N ( C E ( C R E E U R R A S U U S S S R E S S E E E 4 R R T P R P F P 0 R L L A C O L A C A C - E M O O I L L L T f 3 7 0
- 2 0
C' /, 1 0 f l 0 94 9m 9* ) u~ o ~ - 9O %
, ES j:z$Im D E= CwCCrL
o ~ t
- i 0 8' "T .
- o M f UNIFORM PRESSURE (NODE 2) d - 5$ --
- = =
%.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 08 0.9 10 TIME AFTER ACCIDENT (SECONDS) 2
- o E
a M, CAL PRESSURE (NODE 3)
$o~
~ /
/
$ - e 5
b o 5 E / w m o. 7iUNIFORM PRESSURE (NODES 3,4 AND 5) i i i i i y [ e LOCAL ' PRESS'URE (NODES 4 AN D 5)
- a. /
o - ~ i .0 01 0.2 0.3 0.4 0.5 0.6 07 0.8 0.9 1.0 TIME AFTER ACCIDENT (SECONDS) FIG. 6.2.1 - 5 PRESSURIZE 9 CUBICLE AND SUPPORT ShlRT PRESSURElES PONSE FU RTNIN t. - ER IN
.juPPORT SKI PWR REFERENCE PLANT SAFETY AN ALYSIS RE PORT SW ES S AR- Pt W_
AMENDMENT l3 6/30/75
- l. 1 It
5 7
/
6 I
/
6 T 0 N ? l E ! T 1 M N T E M R O
- A N N -
9 0 P M O C E M A
- K N
A T T R 8 E F T O E 0 S E N P A E I - N L L R OE _ P S RR PS I EES E R DC Y 7 RZ E EENAL I P 0 6 EI NR N
) - R RIL EF A-R UU 1
S
. SSE E YA 2
D R TS N 6 SSG ES O , E E R RFE 6 C G RRUWAW I 1 I E F PPSPSS 0 S E D ( O N T ( N E E R 5 D U I S 0 C S C E R R P R M E R 4 T O 0 F F R I N U E M ( 3 I T 0 7 2 0 1 0 t 0
". " O f )
t t N D*O" =r - m N O
) .
Cm _ JT HZWMWts. o wMD UmMQ -
- - 's
O d N N w LOCAL PRESSURE (NODE 3)3 a m _7 w UNIFORM PRESSURE (NODES 3 AND 4 g COMBINED) m - 7 LOCAL PRESSURE (NODE 4)
%.0 0.1 0'. 2 0'. 3 0'. 4 0.5 TIME AFTER GCCIDENT (SECONDS) 8 T.
~ e
~
LOCAL PRESSURE (NODE 1)g d n ~ N - h~ UNIFORM PRESSURE (NODES I AND 2 COM31NED)
* \ LOCAL PRESSURE (NODE 2)
U 8 0
- E 3 '. 0 0.1 0.2 0.3 0.4 0.5 TIME AFTER ACCIDENT (SECONDS)
F I G. 6. 2.1 - 7 PRESSURE RESPONSE STE AM GENERATOR COMPARTMENT PUMP DISCHARGE SER PWR REFERENCE PLANT SAFETY AN ALYSIS REPORT SWESSAR-Pi .- 1 U C, '.1t i i -- AMENDMENT 12 6/16/75
N g= 0. w
]
T N z N UNIFORM PRESSURE w / (NODE 5 I I UNIFORM PRESSURE o (NODE 61
. s N
3 u) W x 1 n_ w
%.0 01 0'. 2 0.3 0.4 05 TIME AFTER ACCIDENT (SECDNDS)
F IG. 6. 2.1 - 7A CM PRESSURE RESPONSE T' ~~ VOLUMES ADJACENT TO STEAM GEN ER ATOR COMPARTM ENT PWR REFERENCE PL ANT S AFETY AN ALYSIS RE PORT
.. SWESSAR - Pl Un t_y A NE NDMENT 13 6/30/75
7
/
6 1
/
6 T N 2 E 1 M T N T E R M D A N P E. M I t T A O R 5 EC TO S R NP _ 0 N O A E W OT L R P A PE IS SR C S E E _ R N EEASN LY t P 8 EE
- RG S RE NA-t U GF R 4
2 S MEE YA 0 6.SEETRA LRFE TS ES G RTOWAW I
)
F P S H P SS S D N O C _ E 3 S ( 0
)
I T N E E D D O I N ( C E C R A U 2 S R S E E 0 T R F P A L A E C M O I L T 1 0 0 o e- D m < 3
^ m a. ~ J. C zwxwu...O L
L . E omswMa. u 9c:' . _ _ _ C
5 7
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5 1
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)
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5 0 EC C E D E D R EEMEAP NNAIP O 9 N O N 0 4 R RURA-R E ( . 1 UEP YA S D* FETS ( ) E E 6 '0 R R 2. S O N R E S D U S U N 6.E G NOI RFE S E S S A 2 5 3
) I FP1 RS WAW 6I PSS E S R R S P P E ) 0 D D 7 N L L O A A O ) D C C N L N C O (
E A ) 0 E O L L-E R D O 3 b 8 3 S U M S N '0 ( T S E A S E D N E D 4 O O 5 E R S
- P E
Ii N 1 ( N ( b E O 2
'0 D
I C R E E N C U R R ( A S U U E S S S R 0 E S S U 2 R R E E S E P P R R S '0 T L P E F e' A M R R L P C R M-O L F I O N A C O L A C L O 5 1
'0 E
M I U L T 0 _ 1 N \ 0
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f f [ 5 0 0 f 0
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f / 0 S- gm O o& 8 O o om o" 3 [tQo J_$ GwQ" BOW $w
- n ^
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SWESSAR-P1 The pump discharge LDR is more severe than the hot leg LDR for a Westinghouse NSSS, as shown in Fig. 6.2.1-9 (W-3S) and 6.2.1-10 (W-3S) , since the mass al.' energy releases from the pump discharge break are greater. The Westinghouse SATAN code used to calculate the blowdown yields a higher flow for breaks in the cold leg piping because of the higher density and pressure of the cold leg fluid. The reactor cavity geometry for RESAR-41 and RESAR-3S plants is identical U (compare Fig. 6.2.1-1(W) with Fig. 6.2.1-1(W-3Sk the break sizes are identical for both plants, and the fluid conditions are nearly identical. Based on the above similarities, the cold leg break is considered limiting for RESAR-41 as has been shown for RESAR-3S. Thus, this figure is not provided for the Westinghouse RESAR-41 NSSS. Fig. 6.2.1-10 Pressure Response Reactor Cavity 144 in.2 Hot Leg LDR H W Amendment 17 9/30/75 bU- ll]
zw w w w w w Iso I?V 180 190* 200* 280* I S 0*
- MS + N
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-ll.,' % N'.Q NODES ABOVE THE (, OF RV NOZZLES ["* ,A f'# "[ '-
y h' iY '
& + 'w e . + +
n%p++ r.g3&,r *n.g p r,- . f v?"> ' </ j:, f 7, -
,x sw a '%, ' % ~~ ' & .
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% 'yNODES BELOW THE { OF RV NOZZLES d' >,,A"
% QRhi m
$ y g." g 3BP ] #
sw sw 3se
- m. sw 20~ ,seo-w w o w w w
iiill iijiiiiilli tilijliii liijiiiitTITI[11lllliijiiiijiiiijiiiij FI G.6.2.1 -l i --- ONE NODE RE ACTOR C AVITY MODEL CALCUL ATEv PR.'SSURE DIFFERENTIAL SIXTEEN NODE REACTOR CAVITY MODEL VS ANGLE FROM BRE AK(BREAK ATO') 150lN 2PUMPDISCHARGE LDR w s~ s-v s. - ,..y PWR REFERENCE PL ANT THE STRUCTURE IS DESIGNED TO SAFETY AN ALYSIS REPORT WITHSTAND THIS PRESSURE PROFILE ' 1 * ') SWESSAR-Pl , l APPLIED AT ANY LOCATION AROUND [ i (; lt 1 I / THE REACTOR CAVITY W L.s s_. -s . % -s , - . . - AMENDMENT 28 876/76
SWESSAR-P1 The pump discharce LDR is more severe than the hot leg LDR for a Westinghouse NSSS, as shown in Fig. 6.2.1-9 (W-3S) and 6.2.1-10 (W-3S) since the mass and energy releases from the pump discharge break are greater. The Westinghouse SATAN code used to calculate the blowdown yields a higher flow for breaks in the cold leg piping becatu.e of the higher density and pressure of the cold leg fluid. The reactor cavity geometry for RESAR-41 and RESAR-3 S g plants is identical (canpare Vig. 6.2.1-1 (W) with Fig. 6.2.1-1 (W-35) , the break sizes are identical for both pl ats, and the fluid conditions are nearly identical. Based on the above siinilarities, the cold leg break is considered luniting for RESAR-41 as has been shown f or RESAR-3S, thus, this figure is not provided for the Westinghouse RESAR-41 NSSS. Fig. 6.2.1-12 Calculated Pressure Differential vs Angle tram Break (Break at 00) 144 in.2 Hot Leg LDR g 3^ , t <_ U W Amendment 17 9/30/75
5 7
/
0 3
/
M 6 AS 3 EK 1 TA T N SE E E R M O R B T N U E TOR E M S SIPNP A EP AELR R R P w P S E E I S TT C Y NA NL I 3 E WE A P 1
- MD R E N -
1 NE F A R A EF EY A I 2 6 T ND R RT'S E F F
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4(8) I 2(6) 4 ~ '3 ;. kNODE/ ,, de ,t J. CIPCUMFERENTIAL . g. :', . s dOUNDARIES A ; :. 6
':c,}
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% q i .,
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s., X 2(6) \ .,
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/ '36' 1(( . S, s.*{l{>o'<*
NOTE THE FIRST NUM3ER REPRESENTS THE NODE BELOW THE NOZZLE CENTERLINE. NODE NUMBERS ABOVE THE CENTERLINE ARE FIG
- 6* 2.1-1 A SHOWN IN PARENTHESES. REACTOR CAVITY NODAllZATION PLAN VIEW THROUGH CENTERLINE OF N0ZZLES PWR REFERENCE PLANT SAFETY ANALYSIS REPORT SWESSAR-P1 -
1q, W-3S ((l\ LO AMENDMENT 26 6/2/ 76
NODE AXIAL RPV FLANGE 9.' i . I
/ -
/ .?-
/ l L NODES 5-8
/ , '. .
f
/ ' *
.o.s j . . .
RPV
% \ '
NODE AXIAL
/
BOUNDARY
/ _ _ [- T. RPV NOZZLE
/
i 6d66636dd66d hSUPPORTS A*- A*
/ NODES 1-4 ..
N EA AL RPV g / [j f, ,,,{ B
/ / /
/ / kRV SUPPORT /
TANK
/ / /
/
, , ,. /
/ o. c: 3 INSULATION (TYP)
NOTES
- 1. THE VENT AREAS ARE AS SHOWN ON TABLE 6.2.1 -13.
- 2. THE VOLUMES ARE SHOWN ON FIG. 6.2.1-1
- 3. SUPPORTED LEG SHOWN. NONSUPPORTED LEG SIMILAR.
FsG. 6.2.1 - 1B REACTOR CAVITY NODALIZATION SECTION VIEW THROUGH CENTERLINE OF NOZZLE PWR REFERENCE PLANT SAFETY ANALYSIS REPORT SWESSAR-P1
, 4 t n li w-3S i I,7 J iL7 AMENDMENT 26 6/2/76
NODE 7 - 1280 FT* h NODE 8 9 152,000 FT' NODE 6 319 0 FT" OPERATING FLOOR < L NODE 5
; 14.170 FT'
=
9 d' GPATE
. 9
=
NODE 4
- _ _
19,680 FT*
- ~
NODE 3 + l g NODE 2 p l 4,8 60 FT' 2400 FT* -
=
9 NODEI
- 4360 FT' 9
- NO3E RECEIVES BLOWDOWN N ODE No. DESCRIPTION I VOLUME BELOW BROKEN LOOP STE AM GENER ATOR SUPPORT PEDESTAL 2 VOLUME BETWEEN BROKEN LOCP STE AM GENER ATOR SUPPORT PEDESTAL AND BIOLOGICAL ShlELD WALL 3 VOLUME SVE IN-CORE INSTRUM EN TATION DRIV E ROOM 4 BRO K E N LO O P S T E A M GEN E R ATOR CO M PART MEN T B E LOW OPER ATIN G FLOOR
( BELOW G R ATE ) 5 BROKEN LOOP STE AM GENER ATOR COMPARTMENT BELOW OPERATING FLOOR ( ABOVE GR ATE) 6 BROKEN LOOP STEAM GENER ATOR SHIELD WALL ABOVE OPERATINC FLOOR TO EL.40'-4" 7 BROKEN LOOP STEAM GENER ATOR SHIELD WALL ABOVE OPER ATING FLOOR E L. 4 0 '- 4" T O 4 5 '- 9 " B THREE IN TACT ST E AM OENERATOR COMPARTMENTS 9 BULK CONTAINMENT (VOLUM E 2. 8 5 X 10 ' F T. 8 ) F I G . 6. 2.1 - 2 PARAMETERS FOR ANALYSIS OF STEAM GENERATOR COMPARTMENT PRESSURE DIFFERENTIALS PWR REFERENCE PL ANT S AFETY AN ALYSIS REPORT . SWESSAR-Pl W-?S bul l U AMENDMENT 18 10/30/75
6 r
/
3 2
/
T 4 N 4 E 2 M T T N R E W A D P T N M R E W
"2 O TO -
/ = C NP AE S
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D N _ p b O D O D O D O E D O w.. 7 ( CO N N N N V N
NODE 5 NODE 4* 3.OlxlO6FT 3 7,920 FT 3
-- -G R AT E NODE 3 9,280 FT 3 NODE 2 II500 FT 3 NODE l**
19,600 FT 3
- SPRAY LINE DER POSSIBLE IN THIS NODE
** SURGE LINE DER POSSIBLE IN THIS NODE NODE NO. DESCRIPTION I - PRESSURIZER RELIEF TANK COMPARTMENT 2- PRESSURIZER CUBICLE (LOWER PORTION) 3- PRESSURIZER CUBICLE (UPPER PORTION BELOW GRATING) 4- PRESSURIZER CUBICLE (UPPER PORTION ABOVE GRATING) 5- BULK CONTAINMENT F IG. 6.2.1-3 PARAMETERS FOR AN ALYSIS OF PRESSURIZER CUBICLE PRESSURE DIFFERENTI ALS PWR REFERENCE PLANT SAFETY ANALYSIS REPORT SWESSAR-PI 7
W-3S bO AMENDMENT 17 9/30/75
1 EL. G7 '- 0" ' n
$y/t.9@?MSM.M.? -
D G h1 NODE 4 %
?G
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Y 9 EL.49 - 4,, ',$ h n :92 VD
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" E L. (-) S I '- 0" MfM?.si!4@i. '.'f@y.g.i?M[S'ft FIG. 6. 2.1 - 3 A PRESSURIZER CUBICLE NODALIZATION PWR REFERENCE PL ANT S A FET Y AN ALYSIS REPORT SWESS AR - Pi
/ <. 1mo W-3S U u 'r icu AMENDMENT 24 4 /23/76
5 7
/
0 3
/
E T 9 E L TR S C I NO T NB APE I T L OURP PC R N E S E S M ER D EICS 3 O N REZ E ENYA1 L 3_ E W 4 A 8 - El NRNP , 1 R RIL EF A-R
)
1 2U U
. S SY ER'TS rA D 6 SSA ES E G. EERR FE N I RRPWAW I
B 6 FPPSPSS I M 1 O EC R 4 U E 4 E) SD R . l R3 SN U 1 U EA S SD R , S
)
SN P 3, E S EA 2 R) D R 4 M P2 2 N P i' RS O S OE LE 1 C LE F D AD E AD I O CO S C OO N ON ( L (N U (N L( 0
- T N
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o~ m , N Y.
- a. _ . ;$ t wb 5 Jm$lma.
4 3
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x - ccO - rC
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R s E D ts E T KN NI t w AL e T E R T o m e FGTO n EERNP a SIUAE NL SL R s OE 3_ PRPI S TES S w ERNCY REE NL EAI P EIZM RN 6 I RRTEA-R 8 2UURF SSA EYA 6 $SPRTS ES
.EEMRFE G
I FPPCPSS RROWA* i i 7 E R U S 1 S S E R i P
- 6. ID t
0 M) C R1 E OE FD S IO N ( i U (M 1 5 T N E D I i 1 C
- 4. C A
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- 3. AF E
N I T r i 2 i i 1 O. e ," =~ N" m e, oc eo_w mta :.szwmw1' c . 4 E wE3oowMa-tt
- I CC4 2
- LO
so C LOCAL PRESSURE 2 (NODE 6) $ s h \ UNIFORM PRESSURE g (NODES 6 AND 7 COMBINED) O U
- a m
W LOCAL PRESSURE o- [(NODE 7 )
%.0 0.1 0.2 0'. 3 0.4 0.5 0.6 0.7 TIME RFTER RCCIDENT ISECONDS) 2 O LOCAL PRESSURE D #"(NODE 4) 0 m N 5
5 t E W E S E
%.0 0.1 02 03 04 0.5 06 0.7 TIME RFTER RCCIDENT (SECS)
FIG. 6. 2. I- 7 PRESSURE RESPONSE STEAM GENERATOR COMPARTMENT PUMP DISCHARGE SER PWR REFERENCE PL ANT SAFETY ANALYSIS REPORT SWESS AR - PI , , ,, Jl W-S S L. u 't AWENDMENT18 0/30/75
CD [ LOCAL PRESSURE u) /(NODE 4) c. ( ^m C z W M b . a W u> m 7 u) W Q. 3.0 0.1 0.2 0 '. 3 0.4 0.5 TIME AFTER ACCIDENT (SECONOS) FIG. 6. 2.1 - 8 PRESSURE RESPONSE STEAM GENERATOR COMPARTMENT HOT LEG SES PWR REFERENCE PLANT SAFETY ANALYSIS REPORT SWESS AR - PI ,,, W-SS [,., ; a l. AMENDMENT 18 10/30/ 75
5 7
/
0 3
/
9 Y T I 7 V 1 A T C N E 0 R M D 5 OR N 0 TD C E M AL A EE T 5 RG R 4 R TO S 0 EA S NPE 3 NH A R W OC L 0 PIS PS I SD ESCY 4 E P R NL I 0 M EAP E U RN - 9- R EAR
) 5 1 UP2 F
EYA S
- 2. SNRTS 1
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- 6. E0l R EFE D'
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)
8 0 SDGI FP1 R 5 PSS WAW N ) D L D N ( S E D E E N D N N 0 O A A 3 CE R D O A U O 2 '3 4 '0 S S N M S S ( S S ( E E E E E T E D R ~ R - D O D O D O O 5 N P U S N N N N 2 E L { D A S ( E I ( S '0 I E ( S E C R E E C O R R C P U R ~R U 0 R L S L - U U S 2
- A C
S E R S 'S S S E E S E '0 R R E O P T L R R P F L P P L 5 A
\ - N C A
O M R dC L A 1 C '0 E
~ _O O M 3
g L F I N O L L 0 T I L w~U - N 1,
\ 0 5
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[ 4- LOCAL PRESSURE (NODES 2 AND3)
/
N w f o
- u. in %
N \ O o
\ '
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= a w
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% .00 0.05 0.10 0 15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 TIME RFTER RCCIDENT(SECONDS)
Cs FIG.G.2.1-lO O' PRESSURE RESPONSE REACTORCAVITY 144 IN2 HOT LEG LDR PWR REFERENCE PLANT
- SAFETY ANALYSIS REPORT ta SWESSAR-PI
~** W -3S AM ENDMENT 17 9/30/75
r w w ~ w ,a ,w 15e use six sso. , , o. 20 , , , , 2 - '; ' a g A++
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W NODES ABOVE THE { OF RV NOZZLES N , 'y 4
'? -.; . y, -a: ,ei)
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.U NODES BELOW THE { OF RV NOZZLES en$ggpx sw 3n
/ 3sr
. , a w w w o ix m.
,x 3x n n
{ . iTJ i l l i j i lTITrrTTy Trri i i i i i i j i i i i i TT r[TTrrjTrri l i t i i iiir[Tq
--- ONE NODE REACTOR CAVITY MODEL FIG. 6.2.!-11 SIXTEEN NODE REACTOft CAVITY MODEL CALCULATED PRESSURE DIFFERENTIAL VS ANGLE FROM BREAK (BREAK AT O*)
THE STRUCTURE IS DESIGNED TO WITHSTAND THIS PMESSURE PROFILE PWR REFERENCE PLANT APPLIED AT ANY LOCATION AROUND SAFETY ANALYSIS REPORT THE REACTOR CAVITY SWESSAR-PI , f, 1 ,)
)
/ .1-3S ) y 't 1 AWENDMENT 28 8/6/76
IE ?& l% 180* I5 % % ii>; l1
! l 1' ,
lr IE N II EE W+# n' Nr' l
#~
?% IG 15 15 ?? 15 u G 2. =k . n - ti (MENr "e e 1-:::= *ni ' 4 .W ~ h :::: W-h ,lxg~ D S :; N - Nhp? l:$l?an 5 n
- ' y; '% qq a
3',J : .
.q.; uz, 3Q' 2g, gy--
ich..n / , n
,. % .' ~ i. %p ,w% .4 y 5 \\
y# P j ,y tr ' u. ,o x',p " , sg l}g" "* ' '
,g
; ?,'pj' [,$(NODES[ ABOVE THE { OF RV y .
I ' ,,; / @*w
' b . m .. "._ _ d b ~ <8 g \, p*Ne~ $ z '
- , gp,,7 N u , ODES BELOW THE { OF RV NOZ7LES - s ,
#[n>~
/(a n w y . ,, 5 '~
_y w n Q ,
?l , $4 ; y *, f
(' .r Q 's y n *e -y
/ v
' W4 ++ w $ ,' h* ; 4x" ig lKj,e' / : . @ +; ,
g g$$ gs - ,g
+,.
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+
Hei
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iiiijiiiiiiijiliijiiiijtiil iiiijiiii iisijiiii lisilitti illijiiii FIG.6. 2.1 -12 SIXTEEN NODE RE ACTOR CALCULATED PRESSURE DIFFERENTIAL CAVITY MODEL VS AN 144. INGLE HOT LEGFROM LOR BRE AK(BREAK ATO PWR REFERENCE PLANT SAFETY ANALYSIS REPORT , 1". SWESSAR-Pl ouil 1>d W-3S AMENDMENT 17 9/ 30/ 75
HOURS g,,,,,,,40-', ,,,,,,40-', , , , ,,,,1 0-' ,,,,,,40-* ,
,,,,,40",
O .4 F T2 L DR AT g 30% POWER % f
- 3 m SPR AY PUMP FAILUREN/\
$ CO 1.4 F T 2 D E R A T O C W PUMP A LURE \
Y m_:_ d, \\ j q
!o / Am&V \\
i : // / N \\ _
!~
8 - gu
// / N\ -
0- '"d ' ' " 1 0' 11' ' ' " 1 1 ' ' ' ' ' ' ' 1 0' TIME AFTER ACCIDENT (SECONDS) cs FI G. 6. 2. l - 13 E' ~^~ CONTAINMENT ATMOSPHERE PRESSURE STEAM AND FEEDWATER PIPE BREAKS PWR REFERENCE PLANT SAFETY ANALYSIS REPORT t 2 SWESSAR-Pl
'J w.3s AMENDMENT 28 8/ 6 / 76
16-NODE REACTOR CAVITY MODEL 180' 225 270* 315 O' 45* 90' 135' 17 17 17 18 0 l g g g 17 17 17 17 17 l , , , g I I I l l l l I l l 10 I 9 1 8 l l l l 3 I 5 l 6 1 7 l l l i l l l i l 3 ' l 217.0 FT 217.0 FT3 2OO.0 FT3 200.0 FT3 217.0 FT 3 217.0 FT 3 200.0 FT3
; , 2OO.0 FT3 -t-l l l l l l 1 I I l- 1 I I I I I l 1 l 1 I I I I I I DL ICF IDL lHL IDL CF IDL lHL IDL
_._ _4 _ __. __ __g__ __g__ _ _ g_ _ _. _ y _ _ __g__ _ _ g_ _ __ _ _f. _ -. I l l l l 1 I l l 1 l l l l l 13 I 12 l l l 11 1 2 l 4 l 16 I I5 l 14 l ; I l i i l 1 il 2.0 FT 31 3 3 3 ll 2.0 FT 90.0FT I 90.0 FT I Il 2.0 FT 3 I Il 2.0 FT 3 90.0 FT 3 I 90.0 FT 3 I I I l l l 1 I I I I I 1 I I i 1 I I I I I I I I I - l i I I l l l I l , l I I I I I I I I i 180* 225* 270' 315* O 45 90* 135 180 NOTES:
- 1. NODES 1,2,3, AND 4 RECEIVE 25 /o OF THE BLOWDOWN FOLLOWING A PUMP DISCHARGE LDR.
NODES 1,2,8, AND 11 EACH RECElVE 25 /o OF THE F I G. 6. 2.1-i BLOWOOWN FOLLOWING A HOT LEG LDR. PARAMETERS FOR ANALYSIS OF REACTOR 3 2.THE VOLUME OF NODE 17 IS 3.44 a 10' FT DL-DISCHARGE LEG PWR REFERENCE PLANT [C[' H L- HOT LE G S AFETY ANALYSIS REPORT CF-CORE FLOOD LINE $W F.SS A R- P1 8SW AMENDMENT 30 1/28/77
HOT LEG TYP. 2 PLACES) J . Jl *
/
COLD LEC l (TYP. 4 PLACES) g i 4 1
' I /
\
1 b / N ;*n -
,\
\
\
14 i 15(6) l l ,., c - s (7), \
/
REACTOR VESSEL SUPPORT o
- CORE FLOOD N0ZZLES REACTOR (TYP. 2 PLACES)
' i EL -5'O" 13 (10)
'\ -
Ig g(5)
/ i g p i w
!2(b i REACTOR VESSEL
/ \ ~
' SUPPORT SHIELD TAN K I 11 ( 8 ) 2(l) g, ,
, g,.
. ;,*T D
/ / i+ \ \
f / 4 l \/\
/
I 1* ---IN SUL ATION (TYP) 1 -. I _).$' SEE dOTEI [ SEE NOTE 1 l ll
.I h
PL AN EL. -6'-6" A +i 0 5 10 13 1.... . .,,l,,,, SCALE-FEET NOTES:
- 1. PENETRATIONS NOT SCALE.
- 2. THE FIRST NUMBER REPRESENTS THE NODE BELOW THE NOZZLE CENTERLINE.
NODE NUMBERS ABOVE THE CENTERLINE FIGURE 6.2.1 - 1 A ARE SHOWN IN PARENTHESIS. REACTOR CAVITY NODALIZATION PLAN VIEW THROUGH CENTERLINE OF NOZZLES PWR REFERENCE PLANT SAFETY AN ALYSIS REPORT SWESSAR -PI , ,, 86W h I' ' ? io/ AMENDMENT 30 1/28/77
- 13'0 " e
- 9'o 5" -
*--- 4" T Y P. NODE A X1AL PV FLANGE P.
j
/ o
/ .?-
I
/
/
/ NODES I,3,5 - 10
/ '$.
, / .,;
,/
RPV / I kl NODE AXIAL j/ BOUNDARY I
/ -
/ T RAV NOZZLE- T-
, 3' 6 "
SUPPORTS AI o'
/ '
NODES 2,4,11-16 ,. ,, RPV g / p 1r NOOE AXlAL
/ //(([/ ///- BOUNDARY 7
u u u
/ / RV SUPPORT /
3 a
/ /; TANK /
/ .
,A, g -a .
d
/ f INSUL ATION
*b- __
jTYP) g.7 1 4 12'8" L 4 4" NOTES
- 1. THE VENT AREAS ARE AS SHOWN ON TABLE 6.2.1 -13.
- 2. THE VOLUMES ARE SHOWN ON FIG. 6.2.1-1
- 3. SUPPORTED LEG SHOWN. NONSUPPORTED LEG SIMILAR.
- 4. DIMENSIONS ARE PRELIMINARY FIG. 6.2.1 - 1B REACTOR CAVITY NODALIZAT!ON SECTION VIEW THROUGH CENTERLINE OF NOZZLE PWR REFERENCE PLANT SAFETY ANALYSIS REPORT ,
SWESSAR-Pt p.; j _, y B&W AMENDMENT 35 10/ 6 / 77
STEAM GENERATOR CUSIC'_E VOLUMES & YENT ARE AS 16 NODE WOCEL Nong 5 NODE NO. DESCRIPYlON
- 1 STE AW GEhERATOR CuqiCLE 12,000 FT5 (PORTION BELOW PEDESTALI
,, 2, 10,11,12 STEAM GENERATOR CUSICLE 13,14,15,16 (LOWER POR TION SE TWEEN PF3ESTAL AND TOWERI 3 STE=W GENER ATOR CuesCLE NODE 4 luPPER PORTION SETWEEN PEDESTAL AND TOWERI 4 STEtu GENERATOR Cut #CLE 5930FT3 (pon tiog OF TOWER SELOW UPPER S G SUPPOR T $ I S STEAM GENERATOR CUTICLE (PORTicN OF TOWER ABOVE UPFER S G SUPPORTSI 6 CON TA t h WE N T (LESS NOttS NODE 3 1 1. 2. 5,4 S. T,5,9,9,ll,12,15,14,l%16)
T OPPOSITE STE Am GENERATOR 32,900 FT 3 CU8ict E (LOWER PORT 104 BELOW
<> 6 h '
TOWEMI 1 S FAR CONTAlkWENT AIR NOOE S FILTRAYiON ROOM ( Siggine FT3 9 NE AR CONTAINNENT AIR FILTR ATION ROOM
@ i NOCE 12 NODE 11 NODE 2 NODE 10 490E 14 h00E 13 NODE IS NODE 16
'{
5450 1410 1880 2170 2170 1980 h FT3 FT3 le tO SESO FT3 FT3 FT3 F 3 FT3 FT3
- O l
i, < q < NOOE T NODE I 35,000 FTI ; 32,3OO FT3 _s, C' FIG. S.2.1 - 2 PAftAMETERS FOR ANALYSIS NOOE e NODE 9 OF STEAM GENERATOR CUBICLE
~ ~ ^
13.SOO FT3 13.SOO FT3 PRESSURE DIFFERENTIALS i - PWR REFERENCE PL A AT SAFETY AN ALYSIS REPORT S *E SS AR - PI 8LW Auf mChimT 54 F/f 2 / F F
m N N h NM M
---i NODE r- l
. U- ,
O* i @ . 8 , l i .
. . i
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16-NODE MODEL NODES NODE FIG. 6.2.I - 2 A STEAM GENERATOR CUBICLE ARRANGEMENT PWR REFERENCE PLANT S AFETY AN ALYSIS REPORT SWESSAR - Pl
~ -
bOl 7i BSU AMENOMENT 34 7/22/ 77
i 1
...m _ _______
l
/
/ UPPER ST E AM
[ { - - - -
,, / GENERATJR SUPPORTS f HOT LEG
/
u
, / i
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) -
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' ,N x q/
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\
GENERATOR }G R ATIN;
)
(
\
COLD LEG / A Q ._. _ _ . . /- f,
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l n,___,,
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L__a___ _._a PEDESTAL Gr U ( 6u1o , .
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- - PRESSURE RESPONSE REACTOR CAVITY-PUMP DISCHARGE LDR PWR REFERENCE DL ANT SAFETY ANALYSIS REPORT
( i SWESS AR - Pl t, BGW AVENDMENT 35 10 / 6 / 7 7
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l l l l I I I I l l l l l l 300 0 60* 12 0 18 0 240 300 NOTES-THE VOLUME OF NODE 13 (CONTAINMENT) IS 3 42 x 10" FT' THE CONTAINMENT RECEfVES THE FOLLOWING PERCENTAGES OF THE BLOWDOWN: HOT LIG SES - 59 */o m 532IN2 PUMP DISCHARGE SPLIT - 50 /o FIG. 6.2.1 - 1 2 m 350" in PUMP DISCHAq E LDR - 0% PARAMETERS FOR ANALYSIS OF REACTOR A' lO$0 WIN NE HO LEb SES an EACH NODE RECEIVES 12.5% OF THE BLOWDOWN PWR REFERENCE PLANT FOLLOWING A 532 IN 2 PUMP DISCHARGE SPLIT AND SAFETY ANALYSIS REPORT 25 /o OF THE BLOWDOWN FOLLOWING A 350 IN2 PUMP DISCHARGE LDR SWESSAR-P1 DL -DISCHARGE LEG HL- HOT LEG C-E S - SUPPORT LEG AMENDMENT 19 12 /12 / 7 5
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Out C-E AMENDMENT 26 6/2/76
I NODE AXI AL . RPV FLANGE eOUNDARY
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INSUL ATION (TYP) NOTES
- 1. THE VENT AREAS ARE AS SHOWN ON TABL E 6.2.1 -13
- 2. THE VOLUMES ARE SHOWN ON FIG. 6.2.1-1.
- 3. SUPPORTED LEG SHOWN. NONSUPPORTED LEG SIMILAR.
FIG. 6.2.1 - 1B REACTOR CAVITY NODALIZATION SECTION VIEW THROUGH CENTERLINE OF NOZZLE PWR REFERENCE PLANT SAFiTY ANALYSIS REPORT , 4 / ~'. SWESSAR-P1 i VJ () b ."t C-E AMENDMENT 26 6/2/76
NODE 1 18,500 FT 3 NODE 2 41,600 FT 3 x NODE 3 NCDE 8 31,000 FT3 3.31 x 106 FT3 NODE 4 NODE 5 NODE 6 990 FT 3 3260 FT3 990 FT3 NODE 7 13,700 FT 3 NODE NO. DESCRIPTION
- 1. STEAM GENERATOR COMPARTMENT ABOVE OPERATING FLOOR
- 2. STEAM GENERATOR COMPARTMENT BELOW OPERATING FLOOR ( ABOVE GRATE)
- 3. STE AM GENERATOR COMPARTMENT ABOVE PEDESTAL (BELOW GR ATE)
- 4. VOLUME ADJACENT TO STE AM GENERATOR SUPPORT PEDESTAL
- 5. VOLUME ADJACENT TO STE AM GENERATOR SUPPORT PEDESTAL
- 6. VOLUME ADJACENT TO STEAM GENERATOR SUPPORT PEDESTAL
- 7. STE AM GENERATOR NODE BELOW PEDESTAL
- 8. BULK CONTAINMENT FIG. 6.2.1- 2 PARAMETERS FOR ANALYSIS OF STEAM GENERATOR
- NODE 3 RECEIVES BLOWDOWN COMPARTMENT PRESSURE DIFFERENTI ALS PWR REFERENCE PLANT SAFETY ANALYSIS REPORT SWESSAR-P1 6 g [7 ) ((.
C-E AMENDMENT 17 9/30/75
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