ML19309C567
| ML19309C567 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 05/01/1967 |
| From: | JERSEY CENTRAL POWER & LIGHT CO., METROPOLITAN EDISON CO. |
| To: | |
| References | |
| NUDOCS 8004080800 | |
| Download: ML19309C567 (333) | |
Text
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.O TABLE OF CONTENTS Section P_g;3 1 INTRODUCTION AND
SUMMARY
. . Volume L . . . Tab 1 . . . 1-1
1.1 INTRODUCTION
. . . . . . . 1-1 1.2 DESIGN HIGHLIGHTS . . . . . . . . . . . . . . . 1-2 1.2.1 SITE CHARACTERISTICS . . . . . . . . . . . . . 1-2 1.2.2 PCWER LEVEL . . . . . . . . . . . . . . . 1-2 1.2.3 PEAK SPECIFIC PCWER LEVEL . . . . . . . . . . . 1-2 1.2.4 REACTOR BUILDING SYSTEM . . . . . . . . . . . . 1-2 1.2.5 ENGINEERED SAFEGUARDS . . . . . . . . . . . . . 1-2 1.2.6 ELECTRICAL SYSTEMS AND EMERGENCY PCWER . . . . . . . 1-3 1.2.7 ONCE-THROUGH STEAM GENERATORS . . . . . . . . . . 1h 13 TABULAR CHARACTERISTICS . . . . . . . . . . . . . 1h 1.h PRINCIPAL DESIGN CRITERIA . . . . . . . . . . . . 1-7 1.4.1 CRITERION 1 . . . . . . . . . . . . . . . 1-7 1.4.2 CRITERION 2 . . . . . . . . . . . . . . . 1-9 1.h.3 CRITERION 3 . . . . . . . . . . . . . . . 1-9 1.k.h CRITERION 4 . . . . . . . . . . . . . . . 1-lC 1.4.5 CRITERION 5 . . . . . . . . . . . . . . . 1-lC 1.k.6 CRITERION 6 . . . . . . . . . . . . . . . 1-12 1.4.7 CRITERION 7 . . . . . . . . . . . . . . . 1-12 1.h.8 CRITERION 8 . . . . . . . . . . . . . . . 1-12 s 1.h.9 CRITERION 9 . . . . . . . . . . . . . . . 1-1.:
1.h.10 CRITERION 10 . . . . . . . . . . . . . . . 1-11 1.4.11 CRITERICN 11 . . . . . . . . . . . . . . . 1-ll 1.h.12 CRITERION 12 . . . . . . . . . . . . . . . 1-l! 1.k.13 CRITERICN 13 . . . . . . . . . . . . . . . 1-1( l.4.14 CRITERION 14 . . . . . . . . . . . . . . . 1- l', 1.h.15 CRITERION 15 . . . . . . . . . . . . . . . 1-1 1.4.16 CRITERION 16 . . . . . . . . . . . . . . . 1-lf 1.h.17 CRITERICN 17 . . . . . . . . . . . . . . . 1 - 15 1.k.18 CRITERICN 18 . . . . . . . . . . . . . . . 1-2( l.4.19 CRITERICN 19 . . . . . . . . . . . . . . . 1-2: 1.h.20 CRITERICN 20 . . . . . . . . . . . . . . . 1 - 2.* 1.h.21 CRITERION 21 . . . . . . . . . . . . . . . 1-2; 1.k.22 CRITERICN 22 . . . . . . . . . . . . . . . 1-22 1.h.23 CRITERION 23 . . . . . . . . . . . . . . . 1-2: 1.h.24 CRITERION 2h . . . . . . . . . . . . . . . 1-2: 1.h.25 CRITERION 25 . . . . . . . . . . . . . . . 1-21 1.4.26 CRITERION 26 . . . . . . . . . . . . . . . 1-21 1.k.27 CRITERICN 27 . . . . . . . . . . . . . . 1-2! 1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS . . . . . . . . 1-2! 1.5.1 CNCE-TERCUGH STEAM GENERATCR TEST . . . . . . . . . 1-2! 1.5.2 CONTROL ROD DRIVE LINE TEST . . . . . . . . . . . 1-2t 1.5 3 SELF-PCWERED DETECTOR TESTS . . . . . . . . . . . 1-2t 1.5.k TIiERMAL AND HYDRAULIC PROGRAMS . . . . . . . . . . 1-2( l 1.6 IDENTIFICATION OF AGENTS AND CCNTRACTORS. . . . . . . . 1-2' l 1.7 CCNCLUSIONS . . . . . . . . . . . . . . . . 1-2( 0003 028 i < +
Section Pm 2 SITE AND ENVIRONENT . . . Volume 1 .
. Tab 2 . . .
2-1 2.1 GENERAL DESCRIPTION . . . . . . . . . . . . . . 2-1 2.2 LOCATICN, POPULATICN,AND LAND USE . . . . . . . . . 2-1 2.2.1 LOCATION . . . . . . . . . . . . . . . . . 2-1 2.2.2 POPULATION . . . . . . . . . . . . . . . . 2-2 2.2 3 LAND USE . . . . . . . . . . . . . . . . 2-2 23 METEOROLOGY . . . . . . . . . . . . . . . . 2-3 231
SUMMARY
. . . . . . . . . . . . . . . . 2-3 2.3.2 SEVERE WEATER . . . . . . . . . . . . . . . 2h 2 3.3 AVERAGE ATMOSPHERIC DISPERSION . . . . . . . . . . 2h 2 3.h ATMOSPHERIC DIFWSION FOR ASSESSING ACCIDENTS . . . . . 2-7 2.4 HYDROLOGY AND GROUNDWATER . . . . . . . . . . . . 2-8 2.h.1 CHARACTERISTICS OF STREAMS IN VICINITY . . . . . . . 2-8 2.h.2 OTHER PCWER PROJECTS IJ VICINITY . . . . . . . . . 2-9 2.h.3 LcW FLcW STUDIES . . . . . . . . . . . . . . 2-10 2.h.k FLOOD FLOW STUDIES . . . . . . . . . . . . . . 2-11 2.h.5 DESIGN OF PROPCSED DAMS AND SPILLWAYS . . . . . . . 2-12 2.h.6 GROUNDWATER . . . . . . . . . . . . . . . . 2-lh 2.5 GEOLOGY , . . . . . . . . . . . . . . . . 2-lh 2.6 SEISMICITY . . . . . . . . . . . . . . . . . 2-15 2.6.1 SEISMICITY . . . . . . . . . . . . . . . . 2 ' ',
2.6.2 RESPONSE SPECTRA . . . . . . . . . . . . . . 2.5 2.7 Ru r.naCES . . . . . . . . . . . . . . . . . 2-16 b ') 3 REACTOR . . . . . . . . Volume 1 . . Tab 3 . . . 3-1 31 DESIGN BASES . . . . . . . . . . . . . . . . 3-1 3.1.1 PERFORMANCE OBJECTIVES . . . . . . . . . . . . 3-1 3.1.2 LIMITS . . . . . . . . . . . . . . . . . 3-1 3.2 REACTOR DESIGN . . . . . . . . . . . . . . . . 3-6 3 2.1 GENERAL
SUMMARY
. . . . . . . . . . . . . . . 3-6 3.2.2 NUCLEAR DESIGN AND EVALUATICN . . . . . . . . . . 3-7 3.2 3 THERMAL AND HYDRAULIC DESIGN AND EVALUATION . . . . . 3-26 3.2.h ECHANICAL DESIGN LAYOUT . . . . . . . . . . . . 3-51 33 TESTS AND INSPECTIONS . . . . . . . . . . . . . 3-82 331 NUCLEAR TESTS AND INSPECTION . . . . . . . . . . 3-82 3 3.2 THERMAL AND HYDRAULIC TESTS AND INSPECTICN . . . . . . 3-82 333 NEL ASSEMBLY, CONTROL ROD ASSEMBLY, AND CCNTROL RCD DRIVE MECHANICAL TESTS AND INSPECTION . . . . . . 3-84 3 3.h INTERNALS TESTS AND INSPECTIONS . . . . . . . . . 3-90 3.k REFERENCES . . . . . . . . . . . . . . . . 3-91 h REAC*CR CCOLANT SYSTEM , . Volume 1 . . . Tab k . . . . h-1 k.1 DESIGN BASES . . . . . . . . . . . . . . . . k-1 h.l.1 PERFORMANCE OBJECTIVES . . . . . . . . . . . . k-1 h.1.2 DESIGN CHARACTERISTICS . . . . . . . . . . . . h-1 h.1 3 EXPECTED OPERATING CCNDITICNS . . . . . . . . . . k-2 h l.h SERVICE LIFE .
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Section k Page k REACTOR COOLANT SYSTEM (CONT UUED)
. . Volume 1 . . Tab 4 h.2 SYSTEM DESCRIP!ICN AND OPERATICN . . . . . . . . . . h-6 h.2.1 GENERE DESCRIPIION . . . . . . . . . . . . . h-6 4.2.2 MAJOR COMPONENTS . . . . . . . . . . . . . . 4-6 h.2 3 PRESSURE-RELIEVING DEVICES . . . . . . . . . . . k-12 h.2.h ENVIRONMENTE PROTECTION . . . . . . . . . . . . 4-12 h.2.5 MATERIES OF CONSTRUCTICN . . . . . . . . . . . k-12 h.2.6 MAXINUM HEAMNG AND COOLING RATES . . . . . . . . . h-lb 4.2.7 LEAK DETECU CN . . . . . . . . . . . . . . . k-ll h.3 SYSTEM DESIGN EVALUATION . . . . . . . . . . . . k-16 h.3.1 SAFETY FACTORS . . . . . . . . . . . . . . . h-16 k.3.2 RELIANCE ON INTERCCNNE N SYSTEMS . . . . . . . . k-23 4.3 3 SYSTEM INTEGRITY . . . . . . . . . . . . . h-23 h.3.4 PRESSURE FELTEF . . . . . . . . . . . . . . . k-23 k.3.5 REDUNDANCY . . . . . . . . . . . . . . . . k-2h h.3.6 SAFETY ANEYSIS . . . . . . . . . . . . . . h-2h h.3.7 OPERATIONE LIMITS . . . . . . . . . . . . . . h-2L h.h TESTS AND INSPECTICNS . . . . . . . . . . . . . h-25 k.k.1 CCMPONENT IN-SEEVICE INSPECTION . . . . . . . . . h-25 h.k.2 REACTOR COOLANT SYSTEM TESIS AND INSPECTIONS . . . . . 4-25 h.4.3 MATERIE IRRADIAUCN SURVEILLANCE . . . . . . . . . k-26 h.5 REFERENCES . . . . . . . . . . . . . . . . . h-26 p 5 CONTAINMENT SYSTEM . . . . Volume 1 . . . Tab 5 . . 5-1 K_)
5.1 REACTOR BUILDING . . . . . . . . . . . . . . . 5-1 5.1.1 DESIGN BASES . . . . . . . . . . . . . . . 5-1 5.1.2 STRUCTURE DESIGN . . . . . . . . . . . . . . 5-2 5.2 ISOLAHON SYSTEM . . . . . . . . . . . . . . . 5-lC 5.2.1 DESIGN BASES . . . . . . . . . . . . . . . 5-lC l 5.2.2 SYSTEM DESIGN . . . . . . . . . . . . . . . 5-1C 5.3 VENTILATION SYSTEM . . . . . . . . . . . . . . 5-12 531 DESIGN BASES . . . . . . . . . . . . . . . 5-12 532 SYSTEM DESIGN . . . . . . . . . . . . . . . 5-13 1 5.4 LEAKAGE MONITORING SYSTEM . . . . . . . . . . . 5-ll l 5.5 SYSTEM DESIGN EVALUAM ON . . . . . . . . . . . . 5-16 5.6 TESTS AND INSPEC* ION . . . . . . . . . . . . . . 5-16 5.6.1 PREOPERAHONE TESTING AND INSPECTION . . . . . . . 5-16 5.6.2 POSTOPERATIONAL LEAK MONITORING . . . . . . . . . 5-l'i 6 ENGDEERED SAFEGUARDS . . . Volume 1 . . . Tab 6 . . . 6-1 6.1 EMERGENCY INJECTION . . . . . . . . . . . . . . 6-1 6.1.1 DESIGN BASES . . . . . . . . . . . . . . . 6-1 6.
1.2 DESCRIPTION
. . . . . . . . . . . . . . . . 6-2 6.1 3 DESIGN EVAWATION . . . . . . . . . . . . . . 6-3 6.1.4 TESTS AND DSPECTIONS . . . . . . . . . . . . . 6-6 6.2 REACTCR BUILDING ATMCSPHERE COOLING AND WASEDG . . . . 6-13 6.2.1 DESIGN BASES . . . . . . . . . . . . . . . . . 6-13 6.2.2 DESCRIP~' ION . . . . . . . . . . . . . . . . 6-13 l
1 0003 030 iii
I l Section Pg l l 6 ENGINEERED SAFEGUARDS (CONTINUED) . . Volume 1 . . Tab 6 6.2 3 DESIGN EVAW AU CN . . . . . . . . . . . . . . 6-lh 6.2.4 TESTS AND INSPECTIONS . . . . . . . . . . . . . 6-19 6.3 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS . . . . . . . . . . . . . . . 6-2C 6.3.1 INTRODUCU CN . . . . . .. . . . . . . . . . 6-2C 6.3.2
SUMMARY
OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERAU CNS . . . . . . . . . . . . 6-2C 6.3.3 LEAKAGE ASSUMPTIONS . . . . . . . . . . . . . 6-21 6.3.h DESIGN BASIS LEAKAGE . . . . . . . . . . . . . 6-22 6.3 5 LEAKAGE ANALYSIS CONCLUSIONS . . . . . . . . . . 6-22 7 INSTRUMENTATION AND CONTROL . Volume.2 . . .. Tab 7 . . . 7-1 7.1 PROTECTION SYSTEMS . .. . . . . . . . . . . . . 7-1 7.1.1 DESIGN BASES . . . . . . . . . . . . . . . 7-1 7.1.2 SYSTEM DESIGN . . . . . . . . . . . . . . . 7h 7.1.3 SYSTEMS EVALUATION . . . . . . . . . . . . . . 7-11 7.2 REGULATING SYSTEMS . . . . . . . . . . . . . . 7-15 7.2.1 DESIGN BASES . . . . . . . . . . . . . . . . 7-15 7.2.2 SYSTEM DESIGN . . . . . . . . . . . . . . . 7-17 7.2 3 SYSTEM EVAWATION . . . . . . . . . . . . . . 7-22 73 INSTRUMENTATICN . . . . . . . . . . . . . . 7-25 7.3.1 NUCLEAR INSTRUMENTATION . 7-25 LO . . 732 NONNUCLEAR PROCESS INSTRUMENTATION . . . . . . . . 7-27 7 3.3 INCORE MONITORING SYSTEM . . . . . . . . . . . . 7-29 7.h OPERATING CONTROL STATIONS . . . . . . . . . . . . 7-31 7.k.1 GENERAL LAYOUT . . . . . . . . . . . . . . . 7-31 7.h.2 INFORMATION DISPLAY AND CONTROL WNCTION . . . . . . 7-31 7.4.3
SUMMARY
OF ALARMS . . . . . . . . . . . . . . 7-32 7.h.h CCMMUNICATICN . . . . . . . . . . . . . . . 7-32 7.k.5 OCCUPANCY . . . . . . . . . . . . . . . . 7-32 7.h.6 AUXILIARY CONTROL STAU CNS . . . . . . . . . . . 7-33 7.k.7 SAFETY FEATURES . . . . . . . . . . . . . . . 7-3h 8 ELEC3 ICAL SYSTEMS . . . . Volume 2 . . . Tab 8 . . . 8-1 8.1 DESIGN BASES .. . . . . . . . . . . . . . . . 8-1 8.2 ELECTRICAL SYSTEM DESIGN . . . . . . . . . . . . 8-1 8.2.1 NETWORK INTERCONNECUCNS . . . . . . . . . . . . 8-1 8.2.2 STAUCN DISTRIBUU CN SYSTEM . . . . . . . . . . . 8-2 8.2 3 EMERGENCY PCWER . . . . . . . . . . . . . . 8-5 8.3 TESTS AND INSPECTICNS . . . . . . . . . . . . . 8-8 I
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0003 031 iv l
Sectien P,,,sg 9 AUE LIARY AND EMERGENCY SYSTEMS . .- Volume 2 . . Tab 9 . . 9-1 91 MAKEUP AND PURIFICAU CN SYSTEM . . . . . . . . . . 9-2 9 1.1 DESIGN BASES . . . . . . . . . . . . . . . 9-2 9 1.2 SYSTEM DESCRIPHON AND EVALUATION . . . . . . . . . 9-3 92 CHEMICAL ADDIUCN AND SMPLING SYSTEM . . . . . . . . 9-9 9 2.1 DESIGN BASES . . . . . . . . . . . . . . . 9-9 9 2.2 SYSTEM DESCRIPTION AND EVALUATION . . . . . . . . . 9-10 93 INTERMEDIATE COOLING SYSTEM . . . . . . . . . . . 9-18 9.3.1 DESIGN BASES . . . . . . . . . . . . . . . 9-18 9.3.2 SYSTEM DESCRIDTION AND EVALUATION . . . . . . . . . 9-18 9.h SPENT WEL COOLING SYSTEM . . . . . . . . . . . . 9-22 9.h.1 DESIGN BASES . . . . . . . . . . . . . . 9-22 9.h.2 SYSTEM DESCRIPTION AND EVALUAU CN . . . . . . . . . 9-22 95 DECAY REAT REMOVAL SYSTEM . . . . . . . . . . . . 9-25 9 5.1 DESIGN BASES . . . . . . . . . . . . . . . 9-25 9 5.2 SYSTEM DESCRIPTICN AND EVALUATICN . . . . . . . . . 9-25 9.6 COOLING WATER SYSTEMS . . . . . . . . . . . . . 9-29 9.6.1 DESIGN BASES . . . . . . . . . . . . . . 9-29 9.6.2 SYSTEM DESCRIPH ON AND EVALUAHON . . . . . . . . . 9-30 97' FUEL HANDLING SYSTEM . . . . . . . . . . . . . . 9-35 9.7.1 DESIGN BASES . . . . . . . . . . . . . . . 9-35 9 7.2 SYSTEM DESCRIPTION AND EVALUATION . . . . . . . . . 9-36 9.8 STATION VENTILATION SYSTEMS 9-kl h 9.8.1 DESIGN BASES . . . . . . . . . . . . . . . 9 kl 9.8.2 SYSTEM DESCRIPTION AND EVALUATION . . . . . . . . . 9 kl 10 STEAM AND PCWER CONVEM ION SYSTEM . Volume 2 . . Tab 10 . . 10-1 10.1 DESIGN BASES . . . . . . . . . . . . . . . . 10-1 10.1.1 OPERATING AND PERFCRMANCE REQUIREMENTS . . . . . . . 10-1 10.1.2 ELI:CTRICAL SYSTEM CHARACTERISTICS . . . . . . . . . 10-1 10.1 3 WNCTIONAL LIMITAUCNS . . . . . . . . . . . . 10-1 10.1.k SECONDARY WNCTICNS . . . . . . . . . . . . . 10-1 10.2 SYSTEM DESIGN AND OPERATION . . . . . . . . . . . 10-2 10.2.1 SCHEMATIC FLOW DIAGRAM . . . . . . . . . . . . 10-2 10.2.2 CODES AND STANDARDS . . . . . . . . . . . . . 10-2 10.2 3 DESIGN FEATURES . . . . . . . . . . . . . . 10-3 10.2.h SHIELDING . . . . . . . . . . . . . . . 10-3 10.2.5 CORROSION PROTECTION . . . . . . . . . . . . . 10-3 10.2.6 IMPURITIES CONTROL . . . . . . . . . . . . . . 10-3 10.2.7 RADI0 ACTIVITY . . . . . . . . . . . . . . . 10-3 10 3 SYSTEM ANALYSIS . . . . . . . . . . . . . . . 10-3 1031 TRIPS, AUTCMATIC CONTROL ACTIONS, AND ALARMS . . . . . 10-3 10.3.2 TRANSIENT CCNDITIONS . . . . . . . . . . . . . 10 L 1033 MALWNCMCNS . . . . . . . . . . . . . . . 10 h 10.3.k OVERPRESSURE PROTECU CN . . . . . . . . . . . . 10-5 10 3 5 INTERACU CNS . . . . . . . . . . . . . . . 10-5 10 3.6 OPERATICNAL LIMITS . . . . . . . . . . . . . . 10-5 n 10.h TESTS AND INSPECMONS 10-5 v . . . . . . . . . . . . . 0003 Di? Y
O Section Page 11 R!cI0 ACTIVE WASTES AND RADIATICN PROTECTICN . . . . . . . . Volume 2 . . Tab 11 . . 11-1 11.1 RADICACTIVE WASTES . . . . . . . . . . . . . . 11-1 11.1.1 DESIGN BASES . . . . . . . . . . . . . . . . 11-1 11.1.2 SYSTEM DESIGN . . . . . . . . . . . . . . . 11-3 11.1.3 TESTS AND INSPECTIONS . . . . . . . . . . . . . 11-11 11.2 RADIATICN SHIELDING . . . . . . . . . . . . . . 11-11 11.2.1 PRIMARY, SECONDARY, REACTOR BUILDING, AND AUXILIARY SHIELDING . . . . . . . . . . . . . 11- 11 11.2.2 AREA RADIATION MCNITORING SYSTEM . . . . . . . . . 11-16 11.2 3 HEETH PHYSICS . . . . . . . . . . . . . . . 11-17
11.3 REFERENCES
. . . . . . . . . . . . . . . . . 11-21 12 CONDUCT OF OPERATIONS . . . . . Volume 2 . . Tab 12 . . 12-1 12.1 ORGANIZATION AND RESPONSIBILITY . . . . . . . . . . 12-1 12.1.1 FUNCTICNAL DESCRIPTION . . . . . . . . . . . . 12-1 12.1.2 QUALIFICATIONS . . . . . . . . . . . . . . 12-2 12.1.3 ORGANIZATION DIAGRAM . . . . . . . . . . . . . 12-2 12.2 TRAINING . . . . . . . . . . . . . . . . . 12-2 12.2.1 STATION STAFF 12-2 O
12.2.2 REPLACEMENT PERSONNEL . . . . . . . . . . . . . 12-5 12.2.3 ON-THE-JOB TRAINING . . . . . . . . . . . . . 12-5 12.2.h ENERGENCY DRII*S . . . . . . . . . . . . . . 12-6 12.3 WRITTEN FROCEDURSS . . . . . . . . . . . . . . 12-6 l 12.4 RECORDS . . . . . . . . . . . . . . . . . . 12-6 1 12 5 ADMINISTRATIVE CONTROL ,. . . . . . . . . . . . . 12-6 13 INITIAL TESTS AND OPERATION, . . . Volume.2 . . Tab 13 . . 13-1 1 13.1 TESTS "RIOR TO REACTOR FUELING . . . . . . . . . . 13-1 l 13 2 INITIAL CRITICALITY . . . . . . . . . . . . . . 13-1 l 13 3 POSTCRITICALITY TESTS . . . . . . . . . . . . . 13-1 l ik SAFETY ANALYSIS
. . . . . Volume 2 . . Tab lh . . 14-1 lk.1 CORE AND COOLANT BCUNDARY PROTECTION ANEYSIS . . . . . Ik-1 14.1.1 ABNORMALITIES . . . . . . . . . . . . . .
. Ik-l ik.l.2 ANEYSIS OF EFFECTS AND CONSEQUENCES . . . . . . . . lk-3 14.2 STANDBY SAFEGUARDS ANALYSIS . . . . . . . . . . 14-15 lk.2.1 SITUATICNS ANALYZED AND CAUSES . . . . . . . . . . 1k-15 1k.2.2 ACCIDENT ANALYSES . . . . . . . . . . . . . . Ik-2(
ik.3 REFERENCES . . . . . . . . . . . . . . . . . Ik-Se 15 TECHNICAL SPECIFICATIONS . . . . Volume 2 . . Tab 15 . 15-1 03 033
. . =- . - - -_ - -
1 TABLE OF APPE'IDICES i Apeendix 1A TECENICAL QUALIFICATIONS . . .. . . . Volume 3 . . . Tab 1A 2A ENGINEERING GEOLOGY AND FOUNDATION CONSIDERATIONS . . .. .. .. .. . Volu=e 3 . . . Tab 2A 23 SEISMOLOGY AND METEOROLOGY . . . .. . Volu=e 3 . . . Tab 23 2C GROUND-WATER HYL.kOLOGY . . .. . . . . Volu=e 3 . . . Tab 2C 2D GEOLOGY . . .. ... .... .. . . Volume 3 . . . Tab 2D SA STRUCTUPJLL DESIGN 3ASES. . . .... . Volume 3 . . . Tab 5A 53 DESIGN PROGRAM FOR FIACTOR BUILDING. . Volume 3 . . . Tab 53 5C DESIGN CRITERIA FOR FIACTOR BUILDING . Volume 3 . . . Tab SC j
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5D QUALITY CONTROL. ...... .... . Volu=e 3 . . . Tab SD SE LINER PLATE SPECIFICATICN . . ... . Volume 3 . . . Tab 5E 5F REACTOR SUILDING I:iSTRUMENTATION . . . Volume 3 , . . Tab 5F Su;;1ement
- 1. . . . .. .. .. .. . . .. .. . . Volu=e h . . Supplement No.1
- 2. .. . . . .. .. . .. . . . . . . . Volu=e h . . Supplement No. 2 O
\ 3 . . . . . . . . . . . . . . . . . . . . Volume 5 . . Supplement No. 3 h............ .. .. .. . . Volume 5 . . Supplement No. k 5 . . . . . . . . . . . . . . . . . . . . Volume 5 . . Supplement No . 5 o
G ' 0003 034 v11 (Revised 6-28-63) l _ -
() Docket 50-289 Supple =ent No. I Cetober 2, 1967 1.0 GE'TERAL QUESTICN Discuss briefly the confor=ance of the proposed design to the 1.1 Ccanission's Gene'ral Design Criteria published July 11, 1967, by referencing those portions of the application which discuss the subject of a criterion or group of criteria. If the design does not conform to a criterion or if the subject of the criterion is not treated in the application, the difference should be discussed in detail. ANSWER The following are the ansvers to the TO General Design Criteria proposed in July 1967 applicable to the Three Mile Island Nuclear Station. I. OVERALL PLANT REQUIRDENTS CRITERION 1 -- QUALIT STA TDARDS (Category A) Those syste=s and cesponents of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety
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3 or to sitigation of their consequences shall be identified and then designed, fsbricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or =odified as necessary. Quality assurance programs, test procedures, and inspection acceptance levels to be used shall be identified. A showing of sufficiency and applicability of codes, stSudard quality assurance progra=s, test procedures, and inspection acceptance levels used is required. ANSWER The systems and components of reactor facilities which are essentia to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences are identified in Section 1.k.1 of the PSAR. The over-all quality assurance program is identified in Section 1 of Appendix SD of the PSAR. The following is a tabulation of reference sections of PSAR pertaining to Criterion No. 1: (Inccmplete infor=ation in this tabulation is under development) A) \w
. 1.1-1 J',, "
D003 065
Syste=s and Cc= enents (Sections 1.h.1) Category
- 1. Fuel Asse=blies A. Ccdes or Standards
- 2. Reacter Vessel Internals 3. Quality Assurance
- 3. 3eacter Coolant Syste= Progrs=s
- k. Reacter Instr., Cont., & Prot. Sys. C. Test Procedures
- 5. Eng. Safeguards D. Inspection Acceptanc
- 6. Radicactive Materials Handling Syste= Levels
- 7. Reacter Building (GAI)
- 8. Electric Power Scurces A 3 C D l 1. 3.1.2.4.2 3.3.3
- 2. 3.1.2.u.1; u.1.u 3 3.u l 3.3.k.1
, 3. h.1.h; k.1.k.h; L.h t h.1 5 k.3.1.1.2
- 4. 3.1.2.k.h; 3.2.h.3.2; 3.3.3; 7.1.1.2 3.2.4.3.k; 7.1.1.2 7.1.1.2 l
S. 9 (Page 9-1) 9 (Page 9-1) 9 (P m 9-1)
- 6. 9 (Page 9-1) 9 (Page 9-1) '9 (Page 9-1) '
- 7. 5.1.2.h; App. 5-D, E 5.6; App. 5-D, App. 5-D, E, F App. 5-A, E, F 3,C,D,E
- 8. 1.h.1 13 l
CRITERICN 2 -- PERFORMANCE STANDARDS (Category A) Those syste=s and cc=ponents of reacter facilities which are essential to the prevention of accidents which could affect the public hea1th and safety or ! to =itigation of their consequer. css shall be designed, fabricated, and erected I to perfor=ance standards that , rill enable the facility to withstand, without loss of the capability to protect the public, the additioca forces that =ight be i= posed by natural pn=ac=ena such as earthquakas, tornadoes, flooding condit:.cas, vinds, ice, and other loca site effects. The design tasos so j established shall reflect: (a) appropriate censideration of the = cst severe l l l i 1.1-2 c 0003 036
(% ( ,) of these natural phenc=ena that have been recorded for the site and the sur-rounding area and (b) an appropriate =argin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design. ANSWER The systems and ccaponents identified in Section 1.k.1 have been designed to perfor=ance standards that vill enable the facility to withstand, without loss of the capability to protect the public, tb additional forces that might be imposed by natural phencmena. The designs are based upon the =est severe of the natural phencuena, recorded for the vicinity of the site, with an appropriate margin t account for uncertainties in the histcrical data. The referenced sections of the PSAR are as follows:
- a. Earthquake App. 23,Part 1; Sect. 5.1.2.3.5
- b. Tornado Sect. 5.1.2.3.2
- c. Flood Sect. 2.k.k; 2.k.5; 5 1.2.3.h '
- d. Wind App. 23,Part 2,and Sect. 5.1.2.3.2
- e. Ice Sect. 5.1.2 3.3
- f. Other Local Site Effects App. SC CRITERICN 3 - FIRE PROTECTION (Category A) r~' The reactor facility shall be designed (1) to minimize the probability of events such as fires and explosiens and (2) to mini =ize the potential effects
' of such events to safety. Ncaccatustible and fire resistant =aterials shall l be used whenever practical throughout the facility, particularly in areas 1 containing critical portiens of the facility such as containment, centrol roa and ccaponents of engineered safety features.
ANSUER The reacter facility is designed to =ini=ize the probability of fir' I and explosion. Ucnccmbustibles and fire resistant materials vill be used whenever practical throurhout the facility. Referenced sections of the PSAR .re as follevs: s
- a. Reactor Building Sect. 5.1.'2.ka
- b. Centrol Rcce Sect. 7.k.5 4
- c. Elect. Distribution I Equipment Sect. 8.2.2.10 CRITERICH k - SHARING OF SYSTEMS (Category A)
Reactor facilities shall not share systems or ccaponents unless it is shown safety is not impaired by the sharing. l \ es st Wh.- 1.1-3 0003 037
s. ANS* DER Since the application describes one nucl?ar unit, no sharing of systems is possible. g CRITERIC i 5 - RECORDS RECUIREMENTS (Category A _ Records of the design, fabrication, and construction of er sential components of the plant shall be =aintained by the reacto- 7erator or under its control throughout the life of the reactor. ANSWE:t Section 12.k of the PSAR deals with record keeping and requires the maintenance of records which verify safe and efficient plant operation. In addition, the following records vill be =aintained by Met Ed:
- a. A complete set of as-built facility plans and syste= diagrams.
This vill include arrangement plans, syste= diagra=s, =ajor structural plans , and technical =anuals of =ajor installed equipment.
- b. A set of co=pleted test procedures for all plant testing outlined in Section 13.
- c. A package of all the quality assw ince data generated during fabrication and erection of the essential ec=ponents of the plant, as defined by the quality assurance progra=, within the scope of Section 1.4.1. Further reference to the quality assurance records is found in Appendix SD of the PSAR.
II. PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS e, CRITERION 6 - REACTOR CORE DESIGN (Category A) The reactor core shall be designed to function throughout its design lifeti=e, without exceeding acceptable fuel damage 11=1ts which have been stipulated and justified. The core design, together with reliable process and decay heat re= oval systems, shall provide for this ca.pability under all expected conditions of nor=al operation with appropriate sargins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pu=ps, tripping out of a turbine generator set, isolation of the reactor from its pri=ary heat sink, and loss of all off-site power. l ANSWER The reactor is designed with the necessary =argins to accom=cdate, l vithout fuel da= age, expected transients from steady state operation l including the transients given in the criterion. Fuel clad in-tegrity is insured under all nor=al and abnor=al = odes of anticipated operation by avoiding clad overstressing and overheating. The evaluation of clad stresses includes the effects of internal and l external pressures, te=perature gradients and changes, clad-fuel 1.1 u O 0003 038
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() interactiona, vibrations, and earthquake effects. The free-standing clad design prevents collapse at the end volu=e reglen of the fuel red and provides sufficient radial and end void volu=e to acccm=cdat clad-fuel interactions and internal gas pressures. (Section 3.2.h.2 Clad overheating is prevented by satisfying the folleving core thermal and hydraulic criteria: (Section 3.2.3.1.1)
- a. At the design overpower no fuel =elting vill occur.
- b. A 99 per cent confidence exists that at least 99.5 per cent of the fuel rods in the ccre vill be in no jecpardy of experiencing a DN3 during continuous operation at the design overpower of lik per cent.
The design =argins allow for deviations of temperature, pressure, flow, reactor power, and reactor-turbine power sismatch. The reacter is operated at a constant average coolant temperature above 15 per cent power and has a negative power coefficient to dampen the effects of power transients. The reacter centrol system vill =ainta the reactor operating parameters within preset li=1ts, and the react protection system vill shut down the reactor if nor=al cperating limits are exceeded by preset a=ounts. (Sections 7.1 and 14) Reactor decay heat will be removed through the steam generators until the reactor coolant system is ecoled to 250 F. Steam generate
/~S by decay heat will supply the steam-driven feedvater pump turbine hg and can also be vented to atmosphere and/or bypassed to the condense The steam generators are supplied feedvater frcm either the =ain stems-driven feedvater pu=ps, which can be operated at a reduced flow rate for decay heat removal, er from a stea=-driven e=ergency feed pump sized at 5 per cent of full feedvater flow.
The =ain feedvater pumps supply water centained in the feeavater train and the condensate stcrage tank to the steam generators. The emergency feed pump takes suction frcm the condenser hotvell and the condensate storage tank. These sources provide at least 365,000 gal of water stcrage which is sufficient ter decay heat removal for about two days after reacter shutdown with the pri=ary heat sink (cendenser) isolated. The condenser is nor= ally available so that water inventory is not depleted. (Section 10) The reactor cociant pu=ps are provided with sufficient inertia to maintain adequate flow to prevent fuel danage if pcuer to all pumps is icst. Natural circulatien coolant flow vill provide adquate core ecoling after the pump energy has been dissipated. (Section 1h.1.2.6 and Figure 9-10) l l l l 1 ' (:) -' I
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CRITERICN 7 -- SUPPRESSION OF POWER CSCILLATIONS (Category 3) The core design- terether with reliable centrols, shall ensure that power oscillations which could cause da= age in excess et acceptable fuel da= age limits are not possible or can be readily suppressed. ANSWER Power oscillations resulting frem variation of ecolant temperature are minimited by constant average coolant temperature above 15 per cent power. Power oscillations from spatial xenon effects are minis 1:ed by the large negative power coefficient. Reactor trip prevents fuel clad damage resulting frcs DNB. The ability of the reactor centrol and protection syste= to control the oscillations resulting frcm variation of coolant temperature within the control system dead band and frc= spatial xenon oscilla-
'tions has been analyzed. Variations in average coolant ta=perature provide negative feedback and enhance reactor stability during that portion of core life in which the =oderator temperature coefficient is negative. When the coefficient is positive, rod motion vill ecupensate for the positive feedback. The =axi=us power change rate resulting from te=perature oscillations within the control system dead band has been calculated to be less than 1 per cent per minute.
Since the unit has been designed to follow ra=p load changes of 10 per cent per minute, this is well within the capability of the control system. Control flexibility with respect to xencn transients is provided by the ccmbination of control rods and nuclear instrumentation. ! Axial, radial, or azi=uthal neutron flux changes will be detected by the nuclear instrumentation. Individual or groups of centrol reds can be positioned to suppress and/or correct flux changes. (Section 3.2.2.2.3) CRITERION 8 -- OVERALL PCWER CCEFFICIENT (Category B) The reactor shall b'e designed so that the everall power coefficient in the power operating range shall not be positive. ANSWER The overall power coefficient is negative in the operating range. (Table 3-8) CF ^.TERION 9 -- REACTCd COOLANT PRESSURE SOUNDARY (Category A) The reactor coclant pressure boundary shall be designed and constructed so as i to have an excee;dingly lcw probability of gross rupture or significant ! leakage throughout its design lifetime. l 1 I n, . t.
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4 ANSWER hhe reactor coolant pressure boundary =eets this criterien by the following:
- a. Material selection, design, fabrication, inspection, testing, ar certification in keeping with the ASME (Section III) and USASI (331.1) Codes.
- b. Quality manufacture including veld qualification test platec, permanent identificatica of materials, velder qualification tests, and extensive production undestructive testing.
- c. Service life based on instability of the =aterial, effects of mechanical shock or vibratory leading, and radiation effects with special consideration to increase in the brittle fracture transition temperature due to neutron irradiation. (Section k.1 CRITERION 10 - CONTAINMENT (Categerv A)
Containment shall be provided. The centainment structure shall be designed t to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situatien requires the functional capability to protect the public. ANSWER The containment structure is designed to provide maximum protection and safety to the public over all normal and accident conditions. Considerations which are used in the design of the contain=ent to assure compliance with the above criterion can be fcund in Sections 5.1, 6.2.1,and ik.2.2.3c of the PSAR. III. NUCLEAR AND RADIATICN CONTROLS CRITERION 11 -- CCNTROL ROCM (Category B) The facility shall be provided with a control roca frcm which actions to
=aintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control reca or other areas as necessary to shut down and maintain safe control.of the facility without radiation ex-posures of personnel in excess of 10 CFR 20 limits. It shall be possible to shut the reacter down and =aintain it in a safe condition if access to the control roan is lost due to fire or other cause.
ANSWER Sections 9.8.2 and 11.2.1.2.k of the PSAR describe the ventilation and radiation shielding design criteria for the control reem; the design vill provide fer adequate protectica of personnel folleving the Maxi =u= Hypothetical Accident. All centrol actions for safe shutdown of the plant vill be provided frcm the control recm. tRefer to Criterica 13) m (i. 0003 041
CRITERION 12 -- INSTRUMENTATICN AND CONTROL SYSTE4S (Categerv 3) Instru=entation and controls shall be provided as required to =enitor and
=aintain variables within prescribed operating ranges.
ANSWER Reactor regulatica is based upon the use of =ovable control rods and che=ical neutron absorber (boric acid) dissolved in the reactor coolant. Input signals to the reactor controls include reactor coolant average te=perature, =egawatt de=and, and reactor power. The reactor controls are designed to =aintain a constant average reactor coolant te=perature over the lead range frc= 15 to 100 per cent of rated power. The steam syste= operates on constant pressure at all leads. Adequate instru=entatica and controls are provided to '
=aintain operating variables within their prescribed ranges.
(Section T.2) The acn-nuclear instru=entation =easures te=peratures, pressures, flove, and levels in the reactor coolant syste=, stea= syste=, and auxiliary reactor systems and =aintains these variables within l prescribed li=1ts. (Section 7.3.2) CRITERION 13 -- FISSION PROCESS MONITORS AND CONTROL 3 (Category 3) Means shall be provided for =enitoring and =aintaining centrol over the fission process throughout core life and for all conditions that can reason.bly be anticipated to cause variations in reactivity of the core, such as indica-tion of positica of centrol reds and concentration of soluble reactivity centrol poisons. ' ANSWER This criterien is =et by reactivity control =eans and control rec = display. Reactivity control is by =ovable control rods and by ct.e=ical nautron absorber (boric acid) dissolved in the reactor coolant. The pcsition of each control rod vill be displayed in the control roc =. The reactivity status of soluble boren vill be indicated by the position of the control rods. Periedic boren concentration in the reactor coolant is determined by the sampling syste= and is reported to the reactor operator. (Section 7.2) l CRITERION IL -- CORE pROTECTICH SYS"T4S (Category 3) Core protection syst.e=s, together with asscciated equip =ent, shall be designed to act autc=atically to prevent or to suppress conditions that could result in exceeding acceptable fuel da= age li=its. 1 ANSWER i The reactor design meets this criterion by reactor trip provisions and engineered safeguards. The reacter protection syste= is de-signed to li=it reactor power which =ight result frc= unexpected reactivity changes and provides an autc=atic reactor trip to prevent exceeding acceptable fuel da= age limits. In a less-of-coolant 1.1-8 9 I h;' le.s:b
l 1 r l 0 : accident, the engineered safeguards protection syste= aute=atically actuates the high pressure and low pressure injection systems. The core ficoding tanks are self-actuating. Certain 1cng-ter:: operation in the e=ergency core cooling syste=s which do not require i==ediate actuation are perfor=ed =anually by the operator, such as remote switching of the low pressure injectica pu=ps to the recirculation
=cde and sa=pling of the recirculated coolant. These operations are perfor=ed =anually to i= prove the overall reliability of the system.
(Section 7.1) CRITERICN 15 - EiGimMED SAFETY FEATURES PROTECTICN SYSTE4S (Category 3) Protection syste=s shall be provided for sensing accident situations and initiating the opere, tion of necessary engineered safety features. ANS*4ER The Reacter Protection Syste= senses abncz=al reactor power level, reacter cutput te=perature, reactor ecclant pressure, and reactor startup rate, and trips the reactor for each abnormal conditicn. The Engineered Safeguards Protection System senses reactor coolant pressure which initiates core coolant injection, and senses reactor building pressure which initiates emergency building cooling. (Section 7.1) CRITERICN 16 - MONITORING REACTOR CCOLANT PRESSURE 3OUNDARY (Catezery 3) Means shall be provided f.or monitoring the reactor coolant pressure boundary k to detect leakage. ANS*4ER Instru=entation is provided to meet this criterien by measuring fluid volu=e changes (pressurizer and reactor building su=p) and radicactivity levels in the reactor building. An increase in net
=akeup to the cc=bined reactor coolant syste= and connected high pressure injectica and purification syste= will indicate leakage.
(Section k.2.7) CRITERION 17 -- MONITORING RADICACTIVI"'Y RMMSE (Category 3) Means shall be provided for =cnitoring the contain=ent at=osphere, the facility effluent discharge paths, and the facility environs for radicactivity that could be released frc= normal operations, frc= anticipated transients, and frc= accident conditions. ANS'4ER The "'hree Mile Island Nuclear Statica is in full ce=pliance with this criteria as indicated by the following references:
- a. PSAR, Section 11, Para. 11.2.2 pp 11-16 through 11-21
- b. PSAR, Section 1, Para. l.h.27, p 1-25
- c. Answers to AEC questiens nu=bered: 10.h and 10 5 O n U. ,
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CRI"'IRICN 18 -- MONITORING FUEL AND WASTE STCRAGE (Cateacry B) Monitoring and alar = instrumentation shall be provided for fuel and vaste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposures. ANSWER The Three Mile Island Nuclear 'tation is in full compliance with this criteria as indicated by the following references:
- a. PSAR, Section 11, Para. 11.2.2 pp 11-16 through 11-21 vith specific emphasis upon sub paras. 11.2.2.1; 11.2.2.2 and 11.2.2.3.
- b. PSAR, Section 1, para. 1.k.25, p. 1-24
- c. Answer to AEC questica numbered: 10.4 IV. RELIABILITY AND TESTABILITY OF PROTECTICN SYS"T4S CRITERION 19 -- PROTECTION SYSTE4S RELIABILITY (Category B)
Protection systems shall be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed. ANSWER The protection systems design meets this criterion by specific location, ample design capacity, component redundancy, and in-service testing. The =ajor design criteria stated below have been applied to the design of the instrumentation.
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- a. No single component failure shall prevent the protection systems from fulfilling their protective function when action is required.
- b. No single component failure shall initiate unnecessary protec-tion system action, provided i=plementation does not conflict with the criterion above.
Manual testing facilities are built into the protection systems to provide for:
- a. Pre-operational testing to give assurance that the protection systems can fulfill their required functions.
- b. On-line testing to provide operability and to demonstrate reliability (Section 7.1.1).
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O - CRITERION 20 -- PROTECTION SYSTE4S REDU'IDANCY AND INDEPENDENCE (Category 3) Redundancy a.d independence designed into protection systems shall be sufficie to assure that no single failure or re= oval frem service of any ecmponent or channel of a system vill result in loss of the protection function. The redundancy provided shall include, as a minimum, two channels of protection for each protection function to be served. Different principles shall be used where necessary to achieve true independence of redundant instrumentation ecmponents. ANSWER Reactor protectica is by fou- channels with 2/k coincidence and engineered safeguards is by three channels with 2/3 coincidence. All protection system functions are i=plemented by redundant sensors instrument strings, logic, and action devices that ccebine to form the protectica channels. Redundant protection channels and their associated elements are electrically independent and packaged to provide physical separation. The reactor protection system initiate a trip of the channel involved when =edules, equipment, or sub-assemblies are removed. (Section 7 1.1) CRITERICN 21 -- SINGLE FAILURE DEFINITION (Category 3) ' Multiple failures resulting frca a single event shall be treated as a single failure.
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ANSWER The protection systems =eet this criterion in that the instrumenta-
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tion is designed so that a single event cannot result in multiple failures that would prevent the required protective action. (Section 7 1.3) CRITERION 22 - SEPARATION CF PROTEC"' ION AND CONTROL INSTRUMENTATION SYSTD4S (Category 3) Protection systems shall be separated frc= control instrumentation systems to the extent that failure or removal from service of any control instrumenta-tien system ccmponent or channel, or of those common to centrol instrumentatio and protection circuitry, leaves intact a system satisfying all requirenents for the protection channels. ANSWER The protection systems instrument strings are electrically and physically- independent . Shared instrumentation for protection and control functions satisfies the single failure criteria by the employ =ent of isolation techniques to the =ultiple outputs of various instrument strings. (Section 7.1 3) i
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CRITERION 23 - PROTECTION AGAINST MULTIPLE DISABILITY FCR PROTECTION SYSTEMS (Categerv 3) The effects of adverse conditions to which redundant channels or protection systems might be exposed in cc= mon, either under nor=al ec2?itions or those of an accident, shall not result in loss of the protection function. ANSWER The protection systems are designed to extre=e ambient conditions and to redundancy. Protection systems instru=entation vill operate frcm k0 - Ik0 F and sustain (except for neutren detectors) the loss of coolant conditions of 55 psig and 281 F and still be operable. Out-of-core neutron detectors are designed for 175 F and 150 psig. Protectica system instru=entation vill be subject to environ = ental (qualification) tests as required by the proposed IEEE Standard for Nuclear Power Plant Protection Systems. (Section 7.1.1) CRITERION 2k - E4ERGENCY PCWER FOR PROTECTION SYSTEMS (Category 3) In the event of loss of all offsite power, sufficient alternate sources of power shall be provided to per=it the required functiening of the protection systems. ANSWER The design of this station vill confor= to this criterion as discussed in detail in Sections 8.2.2.6 and 8.2.2.7 on page 8 k and Section 8.2.3.1 (d) on page 8-6. CRITERION 25 - DE4CNSTRATICN OF FUNCTIONAL OPERABILITY OF PROTECTION SYSTE4S (Category B) Means shall be included for testing protaction systems while the reactor is in operation to demonstrate that no failure or loss of redundancy has accurred.
- ANSWER Test circuits are supplied which utilize the protection systems redundancy, independence, and coincidence features. This makes it possible to =anually initiate en-line trip signals in any single protection channel without affecting the other channels. (Section 7.1.3)
CRITERION 26 - PROTECTION SYSTE4S FAIL-SAFE DESIGN (Category 3) The protection systems shall be designed to fail into a safe state or into a state established as tolerable on a defined basis if conditions such as l disconnection of the system, loss of energy (e.g. , electric power, instrument ( air), or adverse environ =ents (e.g., extreme heat or cold, fire, steam, er l vater) are experienced. l l ANSWER The reactor protection system vill trip the reactor on loss of power. The engineered safeguards protection system is supplied with =ul-tiple sources of electric power for control and valve action. A 1.1-12 0003 046 O p s,c.s,.
total loss of electrical powcr to the engineered safeguards' protecti system vill cause its instrumentation to assu=e a tripped position v the exception of the final control relays. These relays require power to trip. However, since the engineered safeguards equipment also requires power to operate, this relay need not assume the tripped position upon a total loss of power. The =ultip a power supplies for the control relays are also battery-backed and therefor more reliable than the power supply for the equipment. The system is designed for continuous operation under adveyse enviro-ments. The reactor protection system instru=entation within the reactor building is designed for continuous operation in an environ-ment of 140 F, 60 psig, and 100% relative humidity. The neutron detectors are designed for continuous operation in an environment of 175 F, 90% relative hu=idity, and 150 psig. Engineered safeguard equipment and vital instru=entation inside the reactor building are designed for conditions (60 psig, 286 F, and 100% RH) which are in excess of the requirements of the loss of coolant accident. (Sectio: 7.1.1) V. REACTIVITY CONTROL CRITERION 27 -- REDUNDANCY OF REACTIVITY CONTROL (Category A) At least two independent reactivity control systems, preferably of different O, principles, shall be provided.
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ANSWER This c-iterion is met by control rota e ad soluble boren addition to, or removal from, the reactor coolant. (Section 7.2.2.1) CRITTRION 28 -- REACTIVITY HOT SHUTDOWN CAPABILITY (Category A) At least two of the reactivity control syster s provided shall independently b- capable of making and holding the core subscritical from any hot standby or t operating condition, including those resulting from power changes, suffi-ciently fast to prevent exceeding acceptable fuel damage li=its. ANSWER Cne highly redundant reactivity control system consisting of 69 control rods is provided to rapidly make the core suberitical upon a trip signal and protect the core from damage due to the effects of any operating transient. The soluble absorber reactivity control system can make the reactor suberitical even from ultimate power. However, its action is slow and the ability to protect the core from damage which might result from rspid load changes,such as an ulti= ate load turbine trip,is not a design criteria for this system. The high degree of redundancy in the control rod system is considered sufficient to meet the intent of this criterion. (Section 3.2.2.1)
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CRITERICN 29 - REACTIVITY SHUTDOWN CAPABILITY (Categorf Ai
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At least one of the reactivity control syste=s provided shall be capable of ' naking the core suberitical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage 11=its. Shutdevn =argins greater than the maxi =u= vorth of the =ost ; effective control red when fully withdrawn shall be provided. I ANSWER The reactor design =eets this criterion under nor=al operating conditions, and under the accident conditions set forth in Section l 14.1. The reactor is designed with the capability of providing a . shutdown =argin of at least 1 per cent s k/k with the single most I reaesive control red fully withdrawn at any point in core life with the reactor at a hot ::ero power condition. The =ini=u= hot shut-down hargin of 2.1 per cent a k/k occurs at the end of life. (Section 3.2.2.1) CRITERION 30 - REACTIVI""I HOLCDOWN CAPABILITY (Categorv 3) At least one of the reactivity centrol syste=s provided shal: be capable of making and holding the core suberitical under any conditions with appropriate
=argins for centingencies.
ANSWER The reactor =eets this criterica with control rods for hot shutdown under nor=al operating conditions and accident conditions as set forth in Section 14 Reactor suberitical margin is =aintained during cooldown by changes in soluble boren concentration. The rate of reactivity co=pensation frc= bcron addition is Feater than s the reactivity change associated with the =axi=u= allevable reacter cooldown rate of 100 F per hour. Thus suberiticality is assured during cooldown with the = cst reactive control red totally unavail-able. (Sectica 3.2.2.1) CRI"'ERICN 31 -- REACTIVITY CONTROL SYSTEMS MALFUNCTICN (Categorv 3) The reactivity control syste=s shall be capable of sustaining any single
=alfunction, such as unplanned continuous withdrawal (not ejection) of a centrol rod, without causing a reactivity transient which could result in exceeding acceptable fuel da= age limits.
ANSWER The reactor design meets this criterion. A resctor t, rip will protect against continuous withdrawal of one red. CRITERION 32 -- MAXIMLM REACTIVI'"Y WORTH OF CONTROL RODS (Categorf A) Li=its, which include censiderable =argin, shall be placed en the =aximum reactivity worth of control rods or ele =ents and on rates at which reactivity can be increased to ensure that the potential cffects of a sudden er large change of reactivity cannot (a) rupture the reatter coolant pressure 1.1_1s 0003 048 g
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l boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency , core cooling. ANSWER The reactor design meets this criterien by safety features which limit the =ax1=um reactivity insertion rate. These include red group withdrawal interlocks, soluble boren concentratien reduction interlock, maximum rate of dilution water addition, and dilution time cutoff. (Secticn 14.1.2.k) In addition, the rod drives and their controls have an inherent feature to limit everspeed in the event of malfunctions. (section 3.2.h.3) Ejection of the maximum worth control rod will not lead to coolant boundary rupture or in-ternals damage which interfere with emergency core cooling. (Section ik.2.2.2) VI. REACTCR CCOLANT PRESSURE BOUNDARY CRITERION 33 - REACTOR COOLANT PRESSURE BCUNDARY CAPABILITY (Category A) The reactor coolant pressure boundary shall be capable of acccamodating with-out rupture, and with caly limited allowance for energy absorption through plastic deformation, the static and dynamic loads imposed on any boundary ccmponent as a result of an inadvertent and sudden release of energy to the coolant. As a design reference, this sudden release shall be taken as that which would result from a sudden reactivity insertion such as red ejection (unless prevented by positive mechanical means), red dropeut, or cold water addition. ANSWER The reactor design meets this criterien. There are no credible mechanisms whereby damaging energy releases are libertated to the reactor coolant. Ejection of the maximum worth control red vill cat lead to coolant boundary rupture. (Section 14.2.2.2) CRITERION 3h -- REACTOR COOLANT PRESSURE BOUNDARY RAPID PROPAGATION FAILURE PREVENTION (Categerf A) The reactor coolant pressure boundary shall be designed to minimize the prob-ability of rapidly propagating type failures. Consideration shall be given (a) to the notch-tcughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loadings, (c) to the quality cont 'l specified for materials and component fabrication to limit flav sizes, and (d) to the provisions for control over service temperature and irradiation effects which may require operational restrictions. ANSWER The reactor coolant pressure boundary design meets this criterica by the following:
- a. S'lection e of reactor vessel plate material opposite the core to a specified Charpy-V-notch test result of 30 ft-lb or greater at a corresponding NDTT cf 10 F cr lens.
O m . (l.,;l o, t 1-15 0003 049
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- b. Deter =ination of the fatigue usage factor resulting frc= expected I states and transient leading during detailed design and stress analysis.
- c. Quality centr:1 procedures, including pennanent identification of
=aterials and condestructive testing for flaw identificatien.
- d. Operating restrictions to prevent failure resulting frc= increase in brittle fracture transition ta=perature due to neutron irradia-tica, including a =aterial irradiatica surveillance progra=.
(Section k.1.k) CRITERION 35 -- REACTOR COOLANT PRESSURE SOUNDARY BRIT"TI FRAC"'URE PREVENTION (Category A) Under conditions where reactor coolant pressure boundary syste= cc=penents constructed of ferritic =aterials =ay be subjected to potential leadings, such as a reactivity-induced loading, service te=peratures shall be at least 120 F above the nil ductility transition (NDT) te=perature of the cc=ponent
=aterial if the resulting energy release is expected to be absorbed by plastic defor=ation or 60 F above the NDT te=perature of the cc=penent =aterial if the resulting energy release is expected to be absorbed within the elastic strain energy range.
l l ANSWER The reacter vessel is the only reactor coolant syste= cn=ponent exposed to a significant level of neutron irradiation and is there-fore the only ec=ponent subfect to =aterial irradiation da= age. ,The end-of-unit-life NDTT value of the vessel opposite the core vill be g, W ' not more than 260 F. Unit operating procedures vill be established to limit the operating pressure to 20 per cent of the design pressure when the reactor ecolant syste= te=perature is below NDTT +60 F throughout unit life. (Section k.1.k) CRITERION 36 -- REACTOR C00LA3T PRESSURE BOUNDARY SURVEILLANCE (Category A) Reactor coolant pressure boundary ec=penents shall have provisions for inspec-tion, testing, and surveillance by appropriate =eans to assess the structural and leaktight integrity of the boundary cc=ponents during their service lifeti=e. For the reactor vessel, a =aterial surveillance progran conforming with ASTM-E-185-66 shall be provided. ANSWER The reactor coolant pressure boundary ec=penents =eet this criterion in the sense that space is provided for nondestructive testing
=ethods during plant shutdown A reactor pressure vessel =aterial surveillance program confor=ing to ASTM-E-185-66 vill be established.
(Section k.k.3) i 1 0 0003 050 .
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VII. ENGINEERED SAFETY FEATURES CRITERION 37 -- ENGINEERED SAFETY FEATURES BASIS FCR DESIGN (Categorf A) Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure bcundarf, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundarf assuming unobstructed discharge frem both ends. ANSWER The reactor design meets this criterion. The emergency core cooling systems can protect the reactor for any size leak up to and includin the circumferential rupture of the largest reactor coolant pipe. (Section 1h.2.2) CRITERION 38 - RELIABILITY AND TESTABILITY OF ENGINEEREE SAFETY FEATURES (Categorf A) All engineered safety features shall be designed to provide high functional reliability and ready testability. In deter =ining the suitability of a facility for a proposed site, the degree of reliance upon and acceptance of the inherent and engineered safety afforded by the systems, including engineered safety festures, vill te influenced by the known and the demon-3 strated perfomance capability and reliability of the systems, and by the , extent to which the operability of such systems can be tested and inspected '
\ where appropriate during the life of the plant.
ANSWER This criterion is met by the single failure criteria, by periodic testing and by the normal sertice use of protection system active components. The performance testing of the engineered safety features consists of alternate use of equipment in operation and by periodic operation of active ecmponents. (Section 6.2.h and Table 6-3) CRITERICN 39 'iMERGENCY POWER FOR ENGINEERED SAIT"Y FEATLTES (Categerf A) Alternate power systems shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioning required of the engineered safety features. As a minimum, the ensite power system and the offsite power system shall each, independently, provide this capacity assuming a failure of a single active ecmponent in each power system. ANSWER The electrical systems vill confo m to this criterion, as dis-cussed in Sections 8.2.3 and 6.3 of the PSAR. O m . t
..,(r 1.1-17 0003 051
CRITERION ho -- MISSILE PROTECTION (Category A) Protection for engineered safety features shall be providec against dyna =ic effects and missiles that =ight result frc= plant equi;=ent failures. ANSWER Engineered safety features are redundant and either physically separated or shielded to provide against dynamic er m ts and missiles hypothesised by plant equip =ent failure. Refer tc, 5.1.2.7 CRI"ERICU kl -- ENGINEERED SAFETY FEATURES PERFCRMANCE CAPABILITY (Category A) Engineered safety features such as emergency core cooling and contai nent heat removal systems shall provide sufficient perfor=ance capability to accen=cdate partial less of installed capacity and still fulfill that required safety func-tien. As a mini =u=, each engineered safety feature shall provide this required safety function assu=ing a failure of a single active cc=ponent. ANSWER The reactor design =eets this criterien. A single failure analysis of the emergency core ecoling system (Section 6.1.3.1) and Reactor Building heat re= oval syste=s (Section 6.2.3.1) de=custrate that these syste=s have sufficient redundancy to perfor= their design function. The core fleeding tanks contain check valves which operate to pemit flow of e=ergency coolant frc= the tanks to the reactor vessel. These valves are self-actuating and need no external signal or external supplied energy to ake the= operate. Accordingly, it is not considered credible that they would fail to operate when needed. CRITERION h2 -- ENGINEERED SAFE *Y FEATURES CCMPONENTS CAPA3ILITY (Category A) Engineered safety features shall be designed so that the capability of each cc=ponent and syste= to perfor= its required function is not i= paired by the effects of a loss-of scolant accident. ANSWER The reactor design meets this criterion. Piping restraints and equipment location prevent damage to the engineered safeguards equipment frc= pipe whipping 0: =issiles. (Section 6.1.3) CRI"'ERICN h3 - ACCIDENT AGGRAVATION PREVENTION (Category A) Engineered safety features shall be designed so that any action of the engineered safety features which might accentuate the adverse after-effects of the loss of nor=al cooling is avoided. ANSWER The engineered safety features are designed to =eet this criterien. The water injected to assure core cooling is sufficiently berated to insure core suberiticality. Ncnessential scurces of water inside the reacter building are autc=atically isolated to prevent dilution of the berated coolant. Essential scurces of post- , accident eccling vater are =cnitored to detect leakage which =ay 1 1.1-13 0003 057
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O lead to dilution of boron content. An analysis has been made to demonstrate that the injection of cold water en the hot reactor coolant system surfaces will not lead to further failure. The desi6 of the equipment and its actuating system insures that vater injec-tion vill occur in a s' 'ficiently short ti=e period to preclude significant metal-water reactions and subsequent energy releases to the reactor building. (Section 14.2) gTERION kh--EMERGENCY CORE COOLING SYSTEMS CAPABILITY (Category A) At least two emergency core cooling systems, preferably of different design principles, each with a capability for acccmplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core anall be designed to prevent fuel and clad damage that would interfere with ti emergency core cooling function and to limit the clad metal-water reaction to negligible muounts for all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe. The perfor-mance of each emergency core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shall not share active components and shall not share other features or components unless it can be demonstratec that (a) the capability of the shared feature or component to perform its required function can be readily ascertained during reactor operation, (b) failure of the shared feature or component does not initiate a loss-of-coolant accident, and (c) capability of the shared feature or ccaponent to perform its required function is not impaired by the effects of a loss-of-coolant accident and is not lost during the entire period this function is required \( ) following the accident. ANSWER Emergency core cooling is provided by the makeup and purification (high pressure injection) system, the decay heat removal (low pressure injection)~ system, and the core flooding system. These systems prevent clad melting for the entire spectrum of reactor coolant system failures ranging from the smallest leak to the complete severance of the largest reactor coolant pipe. These systems are provided with redundant ecmponents and meet the intent of this criterion. The single failure analysis (Table 6-2) and the dynamic post-accident performance analysis (Section ik) demonstrates that there is sufficient redundancy to preclude the possibility of any single credible failure leading to core melting. As a result of these analyses, the design of the emergency core cooling systems meets the intent of this criterion. (Section ik.2.2 3) CRITERION k5 -- INSPECTION OF IMERGENCY CCRE COOLING SYST3MS (Category A) Design provisions shall be made to facilitate physical inspection of all critical parts of the emergency core cooling systems, including reactor vessel internals and water injection nozzles
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V ? ,l" 'wu i 0003 053 1.1-19
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l s. ANSWEh All critical parts of the e=ergency core cooling system including the reactor vessel internals and water injection nozzles can be inspected during station shutdown. (Section h.k) l , CRITERION L6 -- TESTING OF EMERGENCY CCRE CCOLING SYSTE4S CCMPCNENTS (Category A). Design provisions shall be made so that active co=ponents of the emergency core cooling systems, such as pu=ps and valves, can be tested periodically for operability and required functional perfor=ance. ANSWER The emergency core cooling systems design =eets this criterion by rotating the active ecmponents which are normally in service and also serve as safeguards equip =ent. In addition, periodic tests are performed on ecmponents not nor= ally in service. (PSAR Table 6-3) CRITERION k7 - TESTING OF DERGENCY CCRE COOLING SYSTDfS (Categorf A) A capability shall be provided to test periodically the delivery capability of the emergency core cooling syste=s at a location as close to the core as
- is practical.
t l ANSWER The high pressure and low pressure injection systems are included l as part of normal service systems. Consequently the active components
, can be tested periodically for delivery capability. The core flooding system delivery capability can be tested during shutdown or refueling. In addition, all valves vill be periodically cycled to insure operability. With these provisions, the delivery >
capability of the emergency core cooling systems can be periodically demonstrated. (PSAR Table 6-3) CRITERICN k8 -- TESTING OF OPERATIONAL SEQUENCE OF D4ERGENCY CORE COOLING SYSTDfS (Category A) A capability shall be provided to test under conditions as close to design as practical the full operational sequence that vould bring the emergency core cooling systems into action, including the transfer to alternate power sources. ANSWER The emergency care cooling systems are designed so that insofar as practical th~ full operational sequence can be tested. (Sections 6.1.h and 7.1.3) ! CRITERICN k9 - CONTAINMENT DESIGN BASIS (Categorf A) The contain=ent structure, including access openings and penetrations, and any necessary containment heat rs=cval systems shall be designed so that i the contain=ent structure can acec=modate without exceeding the design leakage rate the pressures and te=peratures resulting frem the largest
- .'s O
-.._aC 0003 054 I
l
s credible energy release following a loss-of-ccolant accident, including a considerable margin for effects frcs =etal-water er other chemical reactions that could occur as a ccusequence of failure of e=ergency core cooling systems AUSWER The contai=ent structure, access openings, p?netrations and necessary containment heat removal systems will be designed to acccmodate the loads specified in Appendix 53, Section 1.2, " Design Loads." The design vill be based upon the factored 1 cads and load ccmbinations as specified in Appendix 53, Section 1.3, " Design Stres Criteria." CRITERICN 50 -- NDT REQUIREENT FCR CONTAINME?C MATERIAL (Categery A) Principal load carrying ecmponents of ferritic =aterials exposed to the external envirc:=ent shall be selected so that their te=peratures under normal operating and testing conditions are not less than 30 F above nil ductility transition (NDT) temperature. ANSWEh Consideration of NDT requirements for ferritic materials is described under the reply to former Criteria No. 11. Further elaboration is contained in the answer to Question T.10.h. CRITIRION 51 -- REACTOR COOLAIC PRESSURE BOUNDABY OUTSIDE CONTAIN!ENT (Category A) If part of the reactor coolant pressure boundary is outside the ccatainment, l appropriate features as necessar-/ shall be provided to protect th". health and safety of the public in case of an accidental rupture in that part. Determination of the appropriateness of features such as isolatien valves and additional contain=ent shall include consideration of the environmental and population conditions surrounding the site. l ANSWER The reactor design =eets this criterion. The reactor coolant pressure boundary is defined as these piping systems or ccmpenents l which contain reacter coolant at design pressure and temperature. ' With the exceptien of the reactor ecolant sa=pling line, the reactor coolant pressure boundary as defined above, is located entirely within the reactor building. The sampling line is provided with remotely operated valves for isolation in the event of a failure. This line is used only during actual sampling operations and can be readily isolated to minimize envircemental doses. All other piping and components which =ay contain reactor coolant are at lov temperature such that any leakage vould be collected by the contaminated drain system. No significanc envin nmental dose would exist frcm these scurces. (Sections 9 1 and 9 2) O m s- -
- 1 m.
1.1-2' 0003 055 i l
CRITERION 52 -- CONTAI:: MENT HEAT REMOVAL SYSTQS (Category A) Q
*ihere active heat re= oval syste=s are needed under accident conditions to provent exceeding contai=ent design pressure, at least two syste=s, preferably of different principles, each with full capacity, shan be provided.
ANS*4ER The reactor design =eets this criterion with two syste=s. the Reactor Building Spray System and the Reactor Bu11 din 6 Energency Cooling Syste=, each with a fun capacity of 2h0 x 10c Stu/hr. (Section 6.2) CRITERION 53 I'CUTAINI:ENT ISOLATION VALVES (Category A) . Penetrations that require closure for the contai=ent function shall be protected by redundant valving and associated apparatus. ANSWER The Three Mile Island Nuclear Station is in fun co=pliance with this criteria as indicated by the following references:
- a. PSAR, Section 5, Para. 5.2 pp. 5-10 and 5-n ; Table 5-3 and Fig. 5 4 (latter two a= ended per reference 2, below)
- b. PSAh, Section 1, Para l.h.22, p. 1-22 CRI"'ERION 5h -- CONTAINMENT LEAKAGE RATE TESTING (Category A)
Contai=ent shall be designed so that an integrated leakage rate testing can be conducted at design pressure after co=pletion and installation of au penetrations and the leakage rate =easured over a sufficient period of time j to verify its confor=ance with required perfor=ance. ANSWER Section 5.6.1.3 of the PSAR describes the initial integrated leak rate test which wi n verify that the leakage rate does not exceed 0.2 per cent by weight of contained air per 2k hours at design pressure. The duration of the test vill be a =in1=u= of 24 hours. CRITERICN 55 -- CONTAINMENT PERIODIC LEAKAGE RA"'E 2 STING (Category A) The contain=ent shan be designed so that integrated leakage rate testing can be done periodicany at design pressure during plant lifetime. ANSWER The contai=ent building vin be designed to provide for periodic integrated leak rate tests. This will enable testing to be perfor=ed throughout the life of the plant at design pressure if required. As stated in Section 5.6.2.1 of the PSAR, periodic leak rate tests of the reactor building vill be performed to de=onstrate the leak rate integrity specified in Section 5 1.2.2. 1.1 22 0003 056 O
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O CRITERION 56 -- PROVISIONS FOR TESTING OF PENETRATIONS (Category A) Provisions shall be made for testing penetrations which have resilient seals or expansion believs to permit leaktightness to be demonstrated at design pressure at any ti=e. ANSWER As stated in Section 5. Para. 5.1.2.6.1 of the PSAR "All penetrations are designed to provide a captive air space that can be pressurized to the Reactor Building design pressure for leak testing and accident conditions." The penetrations are designed with double seals so as to pernit pressurization of the interior of the penetratien by the penetration pressurization system. Large penetrations such as the personnel equip =ent access are equipped with double gasket seals with the space in between the gaskets capable of being pressurized also by the penetration pressurization system. CRITERION 57 - PROVISIONS FOR TESTING OF ISOLATION VALVES (Category A) Capability shall be provided for testing functional operability of valves and associated apparatus essential to the containment function for establishing that no failure has occurred and for deter =ining that valve leakage does not exceed acceptable limits. q ANSWER The Three Mile Island Nuclear Statica is in full ccmpliance with p _j this criteria as indicated by the answer to the original criterica 22, PSAR, Section 1, Para, l.k.22 p. 1-22. CRITERION 58 - INSPECTION OF CONTAINMEN" PRESSURE-REDUCING SYS*LNS (Catagoyq, A}, Design provisions shall be =ade to facilitate the periodic physical inspec-tion of all i=portant ecmponents of the cortain=ent pressure-reducing systems, such as, pumps, valves, spray no::les, torus, and sumps. ANSWER Centainment pressure reducing syste=s are the reactor building spray system and reacter building cooling 17 stem. As shown in
, Figures 1-3 and 1-8 the reactor building ecoling units and the spray pu=ps are so located that physical inspection of these units is possible during non=al plant operation. Valves and operators associated with the piping in each of these systems vill be so located to per=10 the physical inspection of these ecmponents during no m al operatien.
The cpray header and no::les of the reactor building spray system are located in the dcme section of the containment building. This portion of the system is provided with an external air test I connectica for testing the spray ac::les. Telltale devices such ; as strea=ers will be attached to the spray nc::les to enable visual n__- p , 5 t 1.1-23 0003 057
inspection of the functicn of each no::le. This test vould not be cenducted during tor =al operatica but only during a shutdevn period. CRITERION 59 -- TESTING OF CONTAINME'C PRESSURE-REDUCING SYSTE45 CCMPONENTS (Category A) The containment pressure-reducing syste=s shall be designed so that active cceponents, such as pumps and valves, can be tested periodically for operability and required functional perfor=ance. ANSWER This criterica is met by scheduled testing of the active components which are normally in service and also serve as safeguards equip =ent. Valving en the emergency cooling coils can be periodically cycled, thus placing the coils into sertice periodically during operation. The spray system will be tested as described in Criterion 60. (Section 6.2.k) CRITERION 60 - TESTING OF CONTAINMENT SPRAY SYSTE4S (Category A) A capability shall be provided to test periodically the delivery capability of the contain=ent spray systes at a position as close to the spray no::les as is practical. ANSWER The criterion is =et by operation of the spray valves, by blowing low pressure air through the no::les, and by pu= ping water through a return line to the berated water storage tank. (Section 6.2.k) CRITERION 61 -- TESTING OF OPERATIONAL SEQUENCE OF CONTAINMENT PRESSURE-e., REDUCING SYSTE43 (Categer1 A) A capability shall be provided to test under conditions as close to the design as practical the full operational sequence that vould bring the con-l tain=ent pressure-reducing systems into action, including the transfer to ! alternate power sources. ANSWER Provisions to test the operational sequence of the containment I pressure reducing systems (reactor building spray system and i reactor building cooling system) are included in the design of Three Mile Island Nuclear Station. As shown in Fig. 9-10 3 provisions are available whereby the emergency cooling coils of the reactor building ecoling system can be subjected to nuclear services cooling water flov. At such time that an operational sequence test was performed en this syste=, nuclear services ecoling water flow would be established within the coils coincident with the transfer of the nuclear sertices closed cooling veter pu=ps to their alternate emergency pcVer scurce. See Section 6.2.k of PSAR. Test connecticas for testing the building spray nos:les and the reactor building spray pu=ps are provided and shown c= Figure 6-5 One test connection provides for the air test of the
< m .
1.1-as 0003 058 6, , 4 , = 4 s
centainment building spray nozsles. The other test connection provides for an operational test of the spray pumps by recirculating back to the borated water storage tank. A description of the action of switching the ecmpenents of these systems to alternate power sources is presented in Sections S.2.3.1 and 8.2.3.2. . CRITERICN 62 L INSPECTION OF AIR CLEANUP SYSTEMS (Categery A) Design provisions shall be =ade to facilitate physical inspection of all critical parts of containment air cleanup systems, such as, ducts, filters, fans, and dampers. ANSWER The centaic=ent air supply and purge system is in service during normal plant operation. Under an accident condition, the air supply and purge syctem is isolated. Containment air cleanup for iodine re= oval it accomplished by use of a chemical system and this criterion, therefore, is not considered applicable for thic station. CRITERION 63 -- TESTING OF AIR CLEANUP SYSTDt3 CCMPONENTS (Category A) Design provisions shall be =ade so that active ecmponents of the air cleanup systems, such as fans and dampers, can be tested periodically for operability p and required functional performance. D
\
ANSWER Refer to Criterion 62 Answer CRITERION 6h -- TESTING OF AIR CLEANUP SYSTDfS (Category A) A capability shall be provided for in situ periodic testing and surveillance of the air cleanup systems to ensure (a) filter bypass paths have not developed and (b) filter and trapping =aterials have not deteriorated beyond acceptable limits. ANSWER Refer to Criterion 62 Answer I l CRITERION 65 -- TESTING CF OPERATIONAL SEQUENCE OF AIR CLEANUP SYSTEMS l (Category A) l A capability shall be provided to test under conditions as close to design as practical the full operational sequence that would bring the air cleanup systems into action, including the transfer to alternate power sources and the design air flev delivery capability. ANSWER Refer to Criterien 62 Answer I * .\ t. 1 .. s , 4 O 0003 039 1.1-25 l 1
VIII. FUEL AND WASTE STORAGE SYSTDIS CRITERION 66 -- PREVE'ITION OF FUEL STORAGE CRITICALITY (Category B) Criticality in new and spent fuel storage shall be prevented by physical systems or processes. Such means as gec=etrically safe configurations shall be e=phasiced over procedural controls. A'TSWER This criterica is met by the design of the new and spent fuel asse=bly storage facilities to =aintain .tn eversafe gec=etric spacing of 21 in, between asse=blies. Fuel asse=blies cannot be placed in other than the prescribed locatiens. (Section 9.7.2.3) CRITERION 67 -- FUEL AND WASTE STORAGE DECAY HEAT (Category B) Reliable decay heat re= oval syste=s shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public envirens. ANSWER This criterion is met by the spent fuel pool cooling system which contains provisions to maintain water cleanliness, temperature, and water level. Two pu=ps and two coolers vill =aintain the pool at less than 130 ? vith 1-1/3 core storage. If only one heat exchanger and one pu=p are available at this time, the te=perature vill be held at less than 150 F. (Section 9.h.2) CRITERICN 68 -- FUEL AND WASTE STORAGE RADIATION SHIELDING (Category B) Shielding for radiation protection shall be provided in the design of spent fuel and vaste storage facilities as required to meet the require =ents of 10 CFR 20. ANSWER The spent fuel handling building and the areas of the auxiliary building which contain the radicactive vaste disposal facility and auxiliary syste=s carrying primary coolant are provided with shielding to =eet the require =ents of 10 CFR 20. Section 11.2 of the PSAR specifies the design criteria for shielding throughout ) the plant and the design dose rates at various locaticas. The l limits and criteria ec= ply with 10 CFR 20. l i 0003 060 ba
, t-e 1.1-26
O CRITERION 69 - PROTICTION AGAINST RADI0 ACTIVITY RELEASE FRC'M SPENT FUEL AND WASTE STORAGE (Category 3) Containment of fuel and waste storage shall be provided if accidents cculd lead to release of undue amounts of radioactivity to the public environs. ANSWER The Three Mile Island Nuclear Station is to be in full compliance with this criteria as indicated by the following references:
- a. PSAR, Section 1, Para. 1.2.24, p. 1-23 l
- b. PSAR, Section 11, Para. 11-2 pp. 11-11 through 11-15 l IX. PLANT EFFLUENTS
)
CRITERION 70 - CONTROL OF RELEASES OF RADICACTIVITY TO THE ENVIRONMENT (Category 3) The facility design shall include those means necessary to maintain control ; over the plant radioactiva effluents, whether gaseous, liquid, or solid. Appropriate holdup capacity shall be provided for retention of gaseous, liquid or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of l radioactive effluents to the environment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10 CFR 20 require ments for normal operations and for any transient situation that might reasonably be anticipated to occur and (b) on the basis of 10 CFR 100 dosage \ level guidelines for potential reactor accidents of exceedingly low probabilit of occurrence except that reduction of the reconmended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents. ANSWER Section 11 of the PSAR discusses the gaseous, liquid and solid vaste storage facilities and demonstrates that substantial holdup capacity is provided. Release of gaseous and liquid vastes vill be monitored at all times and controlled by the radiation monitoring system described in answer to questions 10.1, 10.h and 10.5 Both the gaseous and liquid vaste discharge lines vill possess autcmatic isolation features to prevent releases in excess of 10 CFR 20 limits. 1 O 1,1_ , 0003 061
( Docket 50-289 Supple =ent No. 1 October 2, 1967 QUESTION What is your criterion with respect to the types and timing of operat 1.2 action to be relied on after a design basis accident? Consideration should be given to such things as switching to the recirculation mode and detecting and isolating a broken injection line. Where the actic is vital to accident recovery and is required within a short time aft the accident (%1 hour), we believe that autcmatic action should be provided. ANSWER A general criterion is to rely upon autc=atic devices to acecmplish all actions required i=nediately following an accident to provide prc tection of the reactor core and reactor building integrity. Operator action is considered quite acceptable 15-20 =in following an accident In fulfilling this criterion, activation of all engineered safeguards systems and isolation valves is acccmplished automatically. Switchin to the recirculation mode is accomplished =anually from the control room; and, depending on the number of pumps in operation following an accident, recirculation is not necessary until at least 20 sin after the accident has occurred. Allowance is made in the calculations pertaining to the effects of a broken low pressure injection line for the a=ount of water that would be lost directly to the reactor building through a broken line. Henc vT detection of a broken injection line is not a critical item, and does @s / not require autcmatic actuation for line isolation. With regard to the failure of a high pressure injection line--when th-final analysis is ccmpleted, it may be determined that it would be advantageous to adf the high pressure injection water thrcugh smaller lines in order to assure that there is adequate ti=e available for operator action. A decision on this matter will be made after final calculations are performed which reflect the piping layout inside the reactor building. In arranging the systems according to the above criterion, it is be- i lieved that the engineered safeguards equipment and associated actu- ' ating circuitry is less complicated and hence more reliable. l O d 0003 062 1 1.2-1
Docket 50-289 O Supplement No. 1 Cetober 2, 1967 QUESTION State your policy on continues operation of the reactor if it 1.3 beccmes necessary to valve off an accumulator tank due to excess leakage or other =alfunction. ANS'aER If it becomes necessary to valve off one ccre ficoding tank due to excessive leakage or other malfunction, the reactor plant will be shut down. I l 1.3-1 0003 0^3
Docket 50-289 O Supplement No. 1 October 2, 1967 2.0 SITE QUESTION Submit the additional information on the fotndation soil' characte: 2.1 istics which we understand has been obtained. Indicate the re-lation of the horings to the location of the foundations of the principal structures including reactor, auxi.'.iary and turbine buildings, cooling towers and service water intake structure. ANSWER Forty-eight additienal borings have been made on Three Mile Island: 15 at the cooling towers, 12 at the Reactor Building, and 21 at the other structure locations. The logs of each hole and the boring location plan are included in Appendisc 2A (soil and rock classification sheets h2 through 136, Figure 2A-3). These borings substantiate the concludions drawn from the Phase I drilling program. At the cooling tower locations,the top of rock elevation varies from 270.5 to a maximum elevation of 276.2. At the Reactor Building and other main structures, the top of rock varies from elevation 275 0 to 279.5. The top of dense to very dense granular overburden material (in excess of 30 B.P.F.-Standard Penetration Test) at the plant O- site ranges from elevation 28h.5 to 289.5, while at the cooling towers the elevation of this horizou varies from 285.0 to 29h.0. The top of the granular material with blow counts in excess of 20 B.P.F.--S.P.T. varies from 288.6 to 295.h at the cooling towers and at the plant site from 295.0 to 298.7. The soils and rock are capable of safely supporting the antici-pated design loads at the proposed elevations. ; The following test data was obtained on BX core samples which ; vere sealed in cheesecloth and wax to preserve natural vate concrete until laboratory tests could be performed. i l TEST PROCEDURES l Five representative samples were tested in unconfined compression and one sample in consolidated, undrained triaxial compression. The triaxial compression test was performed with a cell pressure of 50 psi. All compression test samples, having a diameter of approximately 1.65 2 .02 inches, were trimmed into lengths of 3.03! .02 inches by means of a diamond saw prjor to testing. O e> . 2.1-1 0003 044 I l
l "aree unconfined compressica tests were carried to failure using a rate of strain of .025 in/ min. The remaining unconfined com-pression tests and the triaxial compression test were terminated at a compressive load usually in the order of 100 tsf. Unload-reload cycles were carried out in four of the tests with stress excursions below the elastic limit. TEST RESULTS The pertinent physical, strength,and elastic property test data are tabulated on the attacheu summary of laboratory test results l Table - 2.1-1. It is noted that the First Load, Secant Modulus (Ee ) of the test specimens averaged 0.6 x 103 i.51 and that the Second (re) Load, Secant Modulus (E) aiveraged 1.35 x 103 ksi, based on five and four tests, respectively. The average un-confined compressive strength of the three test specimens which were loaded to failure was 87.h tsf. No significant difference was noted between an ucconfined and triaxial test performed on specimens trimmed from the same core sample. This finding is noted to be consistent with the rel-atively small confining pressure utilized during triaxial testing. O l O
, . .s 2.1-2 0003 0
O D O Table 2.1-1 SLM4ARY OF LABORATORY TEST RESULTS Sample Water Dry Elastic Modulus Compressive Boring / Sample Depth Content Density E E Strength Type (Ft) (%) (pst) c(psi) (psi) (taf) Failure HB - 1BX/1 30.0-30.7 0.85 168.3 0.67 x 10 1.6 x 10 90.7 chisel-shattered HB - IBI/2 33.k-34.1 0.82 167 9 0.50 x 10 1.2 x 10 91.6 " l RB - ILX/3 44.6-45.0 0.92 167.0 0.62 x 10 1. 3 x 10 82.9 " HB - IBX/2A 33.h-34.1 0.90 166.8 0 90 x 10 1. 3 x 10 - - RB - IBX/3a 44.6-45.0 0.94 166.1 0.30 x 10 - - _ C O O u O O
Docket 50-289 O' - Supplement No. 1 October 2, 1967 QUESTION Discuss the availability of emergency and shutdown cooling water 2.2 from the intake structure durir.g period: of extreme low flov (for example, the minimum reported flow of l?' O dvs) assuming a failure of the east channel or York Dams. Is there a flov level belov .<hich water vould be diverted from the channel which serves the service water intake structure? Are periodic dredging or other procedures envisioned to ensure water avail-ability at the intake point over the. life of the plant? ANSWER Loss of York Haven Reservoir could involve failure of the main dam, failure of the East Channel dam, or failure of both dams. Failure of a dam is, in this case, considered a complete removal ; of the structure as a barrier to flow. Loss of the East Channel : dam alone vould occasion caly a partial loss of reservoir, since j the lovest elevation in a section across the entrance to the 1 East Channel is Elevation 275. The reservoir would be lowered i three to four feet, but vould still be of = ore than ample capacity l to furnish water for a safe shutdown, and would also furnish , sufficient water for operation. ! Less of the main dam would return the river to normal open channel flow with the depth of the water, at any point, dependent upon the hydraulic continuum of the river channel systems. A detailed study has been made of the river to determine such conditions, s since the loss of the main dam is the critical condition. A hydrographic survey was conducted over the entire reach of the river, involving horizontal and vertical control by normal methods , of surveying procedure, determination of water surface elevations ) by an integrated gaging system, and the use of a fathemeter survey to determine stream bed elevations. These results gave the nec- l essary information regarding actual channel geometry, in order l to enter an analytical process of flow distribution and water surface I t'.et erminations . I The analytical process was carried out by writing a general program of water surface profiles in a multi-divided channel, using FORTRAN II language, for an IBM 1620 ccmputer. With the channel character-istics, as determined by the hydrographic survey, inserted into the program, it became a specific program for the reach of the Susque-hanna River under question. The program was run with the main dam removed and with a total flow of 1700 cfs in the river, which is the minimua flow on record. This represents the most severe restrictive effect possible, under the existing conditions, for water to reach the nuclear service pump intake. The results of the study sbov that, under the most conservative evaluations, approximately h30 cubic feet per second vill be available at the nuclear service pumps intake based upon a minimum river flow O of 1700 cfs. The quantity is about eleven times the amount required f' 2.2-1 0003 0<7
l l for a safe and orderly plant shutdown. The intake structure vill be constructed at an elevation to take water from the bottom of h l the river and to maintain minimum submergence on the intake pumps at all times. Any silting which might occur, of a consequential nature, vould occur toward the end of the annual spring floods. Hydrographic surveys vill be made as soon as the spring floods subside, and any remedial dredging which is required vill be done immediately. The necessity of any annual dredging with the existing river l conditions is not considered likely, however. A recirculation line vill extend from the plant to the intake to provide hot water frcza +,he circulating water system to control any ice problems which might devcop durin6 cold weather. l i l O I O
. r 2.2-2 0003 068
Docket 50-289 Supplement No.1 October 2, 1967 QUESTION '4 hat is the " probable maximum flood" as defined by the U.S. Army 2.3 Corps of Engineers at the site, including run-up effects? Discuss the uncertainties in the calculated water levels at the site during the 1936 flood which are extrapolations of =easured levels at locations other than the site. ANS*4ER a.) Probable Maximum Flood I The " probable rsximum ficod" as defined by the U.S. Army Corps of Engineers is 1,083,000 cubic feet per second at Harrisburg. This flood is considered to be applicable to the Three Mile Island site in magnitude. No water surface elevations or wave run-up effects are available from the Corps at the site. The nearest water surface elevation which the Corps has es-tablished is elevation 308.2 at the Middletown Air Depot (Olmsted) for a flood flow of 1,000,000 efs. The =ethod for determining the desigr flood elevations at the site, shown in the Preli:dnary Safety Analysis Report, produces an elevati of 309.h at Crawford Station when extended upstress. Crawford Station is downstream of Middletown Air Depot. Provisions for wave run-up at the site consist of from 3 feet to 6 feet freeboard above the maxi =um probable flood for the dike pro-tecting the plant site. b.) Consideration of the 1936 Flcod The water surface elevations calculated for the reach of the
/"'s dusquehanna River between York Haven Hydro Station and Crawfor h Generating Station are the result of interpolation, rather the extrapolation. The gaged water surface at the York Haven Hydr Station Intake was Elevation 287.h, while the gaged water surf at Crawford Generating Station was Elevation 303.2. Three Mil Island is located between the two points.
Upstream of Three Mile Island a major control exists in the river between Hill Island and the East Shore. The control is approximately 2300 feet long and encompasses Fall Island and the Middletown Rapids. The downstrcam end of the control section is approximately 1000 feet north of Sand Beach Island. This is an appropriate location from which to compute the division of flow in the river, which consists of flow in the each channel that is discharged over the East Channel Spill-vay, and the flow in the re=ainder of the river which is dis-charged over the mai: dam. The upstream end of the control section is approximately h500 feet downstream of Crawford Generating Station. DEL" FED t I j !O ! l l ' H 2.3-1 (?evised 12-22-67) 0003 '969 { 1 \
DELETED Water surface profiles were computed from the York Haven Hydro Station Intake, up the headrace channel, and into the reservoir. The water surface in the reservoir upstream of Cly was computed to be at an average Elevation 29h.3 Com-putations were continued using an iterative process involving spillway head-discharge capabilities , water surface profiles, and the division of flow between the east channel and the remainder of the river. Floods of lower magnitudes with the same gaging information as the 1936 flood were available for simultaneous solutions. With a flow of 750,000 cfs in the river the computations showed the division of flow to be 97,600 efs flowing down the east channel and 652,h00 cfs in the re=ainder of the river, of which Sh2,100 efs passed over the nain das and 110,300 cfs passed over the headrace vall. The average computed water surface in the river at the lower end of the Middletown Rapid centrol reach was at Ele-vation 298.k. Calculations based on river sections showed a water surface drop across the control section to be 2.7 _ feet. The average water surface at the upstream end of the control section was computed to be at Elevation 301.1. t Water surface profiles were developed in the channel flowing between Three Mile Island and Shelley Island, or the middle channel, and in the vest channel, whcih flows between Shelley Island and the West Shore. The profiles shoved a calculated vater surface at Goldsboro of approximately Elevation 296.1. Existing water marks of the 1936 flood found at Goldsboro vere determined to be at Elevation 296.3 This confirmation of a calculated vater surface elevation at Goldsboro, and the final agreement of calculated water surface elevations at Crawford Station and the York Haven Powerhouse with gaged elevations for a range of flows between 100,000 and 750,000 cfs indicates that method of analysis produced flood profiles that are of sufficient accuracy. Further confirmation was obtained by establishing gaging points along Three Mile Islanc which provided close correlation with the calculated vater surface elevation in the flow range of approximately 120,000 to 160,000 cfs. l 41> 1 ee s x ' s 2 3-2 (Revised 12-22-67) 0003 070 l
s Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Indicate the location of liquid effluent release from the plant. 2.4 Discuss the discharge of liquid vaste from the site with respect to sources of radioactive vaste, concentrations at the discharge point, and dilution downstream from the site. ANSWER A pipeline normally carrying the cooling tower blevdown and nuclear services cooling water discharge vill release liquid effluent into the river immediately adjacent to the plant. The liquid vaste discharge line vill be connected into this pipeline to enable mixing of liquid vastes with the cooling water discharge. A minimum flow of 2000 gpm frem cooling tower blevdown vill be avail-able for dilution of liquid vastes during plant operation. After shutdown, dilution water in excess of 2000 gpm vill be available, as required, by operation of the nuclear services cooling vater pumps. For the purpose of the liquid waste analysis, a continuous dilution flov of 2000 gym is assumed. The sources of radioactive licuid vaste considered in the analysis result from continuous chemical shim bleed operation, collection of miscellaneous vastes, and two cold startups near the end of the core life. The reactor coolant activities are those listed in Section 11, - Table 11-3. The chemical shim bleed calculation uses an average bleed rate of 25 gph; a removal efficiency in the purifica-tion demineralizer of 99 percent for all nuclides except Kr, Xe,and Cs; an evaporator decontamination factor of 10-4; and a removal efficiency in the cation demineralizer of 99 percent for ca. The resulting concentration in the plant effluent, with a dilution flow of 2000 gym and no holdup for decay, is 0.005 MPC. The addi-tional activity due to miscellaneous vastes was evaluated on the conservative basis that all vastes are reactor coolant leakage and are collected at a continuous rate of 25 gpd and processed without holdup through the vaste disposal system. With the same assu=ptions as above, the concentration in the plant effluent is then 0.0002 MFC. Two cold startups at the end of core life vill generate approximately 28,000 ft3 of liquid vaste. Processing this vaste through the vaste disposal system, without holdup, at a rate of 7.5 gpm produces an average annual concentration in the plant effluent of 0.005 MPC. The total average annual concentration in the plant effluent as a result of the above chemical shim bleed, miscellaneous vaste, and cold startup operations is about 0.01 MPC. Liquid vastes discharged into the Susquehanna River vill be carried 3-1/4 miles downstream to the York Haven Dam. The main dam begins at the tip of Three Mile Island and progressively channels the river flow into the headrace of the York Haven Hydro Station. With river flows less than 20,000 cfs, all the water passes through the hydro y units and tailrace and ec=plete mixing occurs. Since the river flov l nc .. 2.s_1 0003 071
,ti ; \.
- U
is greater than 2000 cfs over 99 percent of the time, this flow is selected as a conservative basis for estimating the dilution of g liquid vastes passing through the hydro units. The dilution factor corresponding to a dilution flow of 2000 cfs and an esti=ated plant effluent release rate of 52 gph is 9.6 x 10-7 The total average annual concentration in the river will therefore be reduced frem a maximum of 0.01 MPC entering the headrace of the Hydro Station to about 9.6 x 10-9 of MPC downstream of York Haven. This analysis is conservative since it assumes:
- 1. No mixing in river anead of York Haven Dam The dilution flow of 2000 cfs is a minimum value which hss been
-2.
observed less than 1 percent of the time.
)
I I O i*'* i ,- 2.k-2 0003 072
Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Discuss the accidental dumping of on-site stored vaste with respect 2.5 to the effect on downstream water supplies and reconcentration points Supply a list of these points downstream from the site, the storage capacity of each (including fire reserve) and the length of time water usage could be suspended as compared to the length of time the river would be at concentrations higher then MPC. Supply a similar analysis assuming that the accidental discharge was after the occurre: of a design basis accident. What doses vould result to a person who did not suspend normal usage of water under these conditions? ANSWER Accidental dumping of on-site stored vaste has been analyzed and is reported in Section 11.1.2.5.1 of the revised Section n . In this analysis, the pumping of 10,000 gallons of evaporator condensate at 50 gpm into the cooling tower blevdown effluent of 2000 gpm results in a temporary concentration of 0 5 of MPC in the effluent. The slug of vaste vin be carried downstream at an average flow of 7 feet per minute to the York Haven Hydro Station and is conservatively assumed to mix with a minimum dilution flow of 2000 cfs passing i through the hydro units and tailrace. The dilution factor corres- ; pending to a slug release rate of 50 gpm and dilution flow cf 2000 i cfs is 5.5 x 10->. Downstream of York Haven, the slug concentration v111 therefore be reduced to approximately 2.8 x 10-5 of the MPC. The concentration in the river following this accidental release s O. vill thus always be below MPC either ahead of,or downstream of, York i Haven. An inadvertent release of rav waste from a reactor coolant bleed tank or a miscellaneous vaste tank is not regarded as credible since these tanks cannot be dischuged directly to the environment. The plant is designed such that the only credible release point for accidental discharge of liquid vastes vill be from the evaporator condensate storage tanks. Discharge of vastes from these tanks vill be monitored at au times using redundant monitors which vill auto-matica11y shut off the liquid vaste discharge line in the event that concentrations above 10 CFR 20 limits are being discharged. l Accidental release of MHA fluid directly to the environment is not possible since it is contained within the reactor building or within the recirculation systens which comprise the decay heat removal loop and containment spray system. Any leakage from these systems vill collect in the auxiliary building drain system and is pumped to the auxiliary building sump tank. This fluid may then be processed in i the vaste disposal system. The bulk of the MHA fluid vill be contained within the reactor building and recirculation systems until all consequences of the accident have been controlled. When desired, the fluid conected in the reactor building may be pmped out in batches via the reactor building sump to the miscellaneous vaste tank and processed in the vaste disposal system. O l l
" J 0003 073
\ 2.5-1 l
l Section lb.2.2.k.5 of the PSAR gives the activity of post-MHA fluid in the recirculation systems as .037 equivalent curies of I-131 per cc cf water. Assu=ing the fluid collected in the auxiliary building st=p tank and =iscellaneous vaste tank has this activity level for
-131, it vill be processed througg the vaste disposal sys:em, with a decentamination factor of 10 , and collected in an evaporator cendensate storage tank. Re= oval of 90 percent of the iodine by the cation demineralizers is conser-vatively assumed during this operation. Accidental release of 10,000 ganons of vaste frem the evaporator condensate storage tank at 50 gpm win take place with 18,000 gpm dilution f2cv fr a the naclear services cooling pump discharge. The resulting concentration in the plant effluent is about 3300 MPC for a period of approximately 3 hours. A child with a 2gm thyroid consuming 300 cc of this water vould receive a dose of 3h Rem to the thyroid which is well within the limits of 10 CFR 100.
The list of drinking vater intake points given in the latter part of this discussion indicate that au public water supply intakes are downstream of York Haven Hydro Station. With a dilution factor of 5.5 x 10-5, the concentration of I-131 vould be about 0.17 MPC after passing through the York Haven nydro units and tailrace. A child with a 2gm thyroid consuming 300 cc of this water would receive only 0.002 Rem to the thyroid. Thus the accidental release of liquid vastes will not have an appreciable effect on downstream drinking vater supplies. The following summarizes the available information on water supplies
)
downstream from Three Mile Nuclear Station for a distance to 50 miles. The consumers include public water supplies, industries,and utilities. There are no points devnstream from I'hree Mile Island where significant concentration of river flow occurs. Storage capacity for each supply is indicated; however, it was not possible to determine, in most cases, the amount of water in storage raserved for fire protection.
- 1. The Internatienal Paper Ccmpany evns and operates a paper manu-facturing pls.nt along the vest shore cf the Susquehanna River at York Haven. The plant utilizes two nillien ganons of river water per day in paper making. The river water is filtered before use in the plant. The plant also generates its own electric power fr::a the river, estimated to be 2000 KW, utili-ting 1500 ers of the river flov. The plant obtains its potable water supply frem two ve n s; estimated consumption is 2 million gallons per year.
The paper companf ms.:ufanu:-ers approximately 80 tons of paper per 2h hour day, five days per week. About 75 percent of the paper manufactured is used in food vrapping. The vaste dis-charged to the river censists of 1.2 MGD of white paper and beaten pulp and 800,000 San ons per day cf digester liquor and vet room vaste from its pulp p ecess. The vaste has the follow-ing average characteristics:
.s x c . v. :
0003. 374 l .*
() 1. Suspended solids - 1900 lbs/ day
- 2. BOD population, equivalent - 6500 The plant it located approxi=ately 3-1/h miles downstream from the Nuclear Station.
- 2. Metropolitan Edison Company owns and operates a hydro-electric generating station at York Haven, total installed capacity of 20,000 KW. A dam across the Susquehanna River impounds the water for generating purposes. The capacity of the York Haven reservoir is approximately 8,000 acre-feet. The plant also uses river water for air conditioning, transformer and turbine cooling, and for fire protection.
River vater is not used for potable purposes. Potable water is obtained from two wells approximately 100 feet deep. The York Haven Powerhouse is located about 3-1/4 miles downstream from the Nuclear Statien.
- 3. The Pennsylvania Supply Company takes 2200 gallons per minute per 9 hour day frca the month of the Conevago Creek. The water is used in a sand and gravel processing plant and thereby dis-charged into the River. Prior to dischargs the water passes through settling basins for sedimentation. The plant is located about 3-1/2 miles downstream from the Nuclear Station. ':he O material deposited in the settling basins is inert, suspended solids at an average concentration of 10,370 ppm. The total basin storage capacity is sufficient for 295 months of operation.
- h. The Pennsylvania Power and Light Company owns and operates a steam electric generating station, located on the vest shore of the Susquehanna River on 3 runner's Island, about one mile south of York Eaven. The present plant generating capacity is 665,000 KW, and it utilizes 515 cubic feet per second of river vater for the circulating water cooling system and ash removel.
River water for these purposes is untreated prior to use and returned to the river without reconcentration. Settling basins are used to treat the ash sluice water before it is returned to the river. A third unit is presently under construction and vill be in operation in late 1968. This vill increase the plant generating capacity to 1,h15 M4 and the circulating vater flow to 1,155 cfs. The plant is located about five miles dovuttream from the Nuclear Station. River vater is also used for potable water requirements and cycle makeup demineralizers. This vater is treated by coag-ulatien, filtratien, softening, and chlorinatica prior to use. Demineralizer vaste is discharged to the river. Water for potabl and cycle =akeup requirements is stored in a 200,000 gallon t _ ,- stcrage tank. Fifty percent of the storage tank capacity is
; <a. 2.3-3 0003 075 l i
reserved for fire prctection, and the remainder vUl provide water for 2 - 3 days normal operating requirements. 5 The Borough of Marietta has 3 pipe lines under the Susquehanna River to deliver water to the Borough. The water is taken from two reservoirs high on the vesterly bank of the river and piped to the Borough on the easterly b:.uk; however, no vater is taken from the River. The creasings consist of one-10" and one-12" service line and a spare 5" line under the river. The crossings are located 13 miles downstream from the Nuclear Etation.
- 6. The Wrightsville Water Supply Company has a public vater supply intake in the Susquehanna River located about 16-1/h miles downstream from the Nuclear Station. The source is used as a summer reserve supply, the main source being from springs and wells in the area. Water from the main source is treated by chlorination; vater from the Susquehanna River is subjected to l complete treatment, including coagulation, sedimentation, fil-tration and chlorination. The treatment plant can be operated only in varm veather; therefore, water is not drawn from the river during the vinter.
The water system serves the Borough of Wrightsville, population 3,500. Average consumption is 150,000 gallons per day. Sto-age facilities consist of a 200,000 gallon reservoir which stores runoff Excess water frem the main supply from the springs and wells. is stored in a quarry of unknown capacity, but sufficient to enable the system to operate for several months utilizing spring flow. Both the reservoir and the quarry store water for fire l protection pumps. Intake and treatment facilities for river water have a capacity to provide 300,000 gallons per day, and consist of the following: Screened 6 inch suction line Rav vater pump house - 200 gpm 1 sedimentation tank - 55 h00 gallons 2 Sand filters - 100 gpm l 1 Clear well - 10,000 gallons 1 Chlorinator 1 Wash water tank acd lagoon - 15,800 gallons
- 7. The Borough of Columbia takes an average supply of two million gallons per day from the Susquehanna River for potable water supply to serve a population of 14,000 people. The water is treated before use by coagulation, sedimentation, filtering, and chlorination. The intake is located about 16-3/4 miles downstre..2 from the Nuclear Station.
Intake and treatment facilities consist of the following: 16 inch rav vater intake from rivtr to pump station 3 rav vater pumps to treatment pitat 1 coagalation tank 3 settling basins - 213,000 gal. total h e I:: o 2.5 u 0003 076
i I l l 6 anthrafilt filters Os 8 pressure filters -for emergency use at high river stages 1 clear ven - 50,000 gallons 3 chlorinators 2 high lift pumps System sto' age facilities consist of a reservoir and two elevate ; tanks with booster stations, providing for the storage of ' h,850,000 ganons of treated vater, including fire reserve of - unknown quantity. Under normal conditions the total system I storage vill provide about 2-1/2 days supply to the systam. Effluent returned to the river includes filter backvash from the water treatment plant and effluent from a municipal sewage treat-ment plant providing primary treatment-
- 8. The City of Lancaster has a water supply intake on the Susquehanr River located approximately IT-3/h miles downstream from the Nuclear Station. The eity presently takes about eight million galicas per day from the river. Other sources of supply include a dam over Conestoga Creek, a ven,and a quarry. The average consumption of the system is about 15.75 MGD, serving an estima-ted 90,000 people. The capacity of all sources of supply is 24 MGD, one-third of which comes frem the Susquehanna River.
The water is treated by coagulation, sedimentation, filtration, chlorination,and flouridation before distribution in two treat-ment plants. Treatment facilities consist of the following,
\ providing a total capacity of 2h MGD:
8 Coagulation tanks 5 Settling basins 12 Filters (rapid ssnd) 7 Chlorinators 3 Flouridators Filter backvash water is returned to the rive . Municipal sewage is treated in two activated sludge type sewage treatment plant and the effluent is returned to the river. Storage facilities consist of a ground storage reservoir, stand-pipes and elevated tanks, providing a total quantity of 30 minic ganons, including fire reservr. of unknown quantity. Under norma conditions the storage facilities vi n provide a two-day supply. If the Conestoga supply were utilised and the Susquehcnna supply inoperative, the system storage would provide about h days con-sumptive under normal conditions of demand. 9 A public water supply system serves tne vin age of Safe Harbor and the Safe Harber Hydroelectric Plant located on the Susquehann River. The plant is evned by the Safe Harbor Water and Power l Corporation. The powerhouse is located about 27-1/h miles down-stream of the nuclear station. Safe Harbor reservoir cortains x e 2.5 5 conz v7
about 92,000 acre-feet. Water is taken from the pool behind the dam at an average rate of abou: 25,000 ganons per day. h Treatment consists of coagulation, sedimentation, filtration, and chlorinstion with the following facilities: 1 Coagulation tank 1 Settling basin - 24,500 gallons 2 Rapid sand filters - 158,000 ganons 1 Chlorinator 2 High lift pumps - 250 gym 1 Clearvell - 50,000 ganons Storage facilities include a 100,000 gallon tank, in addition to to the clearven, which vill provide for a supply of about six days under normal conditions. Storage for fire protection purposes is included in the 150,000 ganen total amount.
- 10. A public vater supply system serves the vin age of Holtvood on the east bank of the Susquehanna River, and the Holtvood Hydro-electric Plant. Water is taken from a pool behind the Holtvood Dam at an average rate of 22,000 gallons per day and treated before distribution. Treatment consists of coagulation, sedi-mentation, filtration, and chlorination. The faciliti-s serve an estimated 225 people. The Holtvood Reservoir con' ns about 19,300 acre-feet of storage. The powerhouse is loca about 3h-3/h miles downstream frem the Nuclear Statien.
Treatment facilities consist of: 1 Coagulation tank - 125,000 ganons 2 Settling basins h0,500 ganens each j 2 Filters 1 Chlorinator 2 Pumps - 300 gym Storage facilities include three reservoirs and an elevated tank providing a total quantity of 2h0,000 ganons , including fire protection. Under normal conditions the storage facilities vill provide for ab mt u days consumption.
- 11. The Muddy Run PumpedStorage hydro plant is located about 38 miles downstream from Three Mile Island Nuclear Station. It has a capacity of 800 MW. Muddy Run is a remote control plant and is owned by and operated by Philadelphia Electric Company. The generating plant is located on the east bank of the Suequehanna Riter, about two miles downstream from Holtvood Dam. Water is pumped from the Susquehanna River to an upper reservoir by re-versible pump-turbine. The water is used to generate power at times of peak demand by flowing back throu6h the powerhouse to the river. The capacity of the installation is 800,000 KW.
Potable water is obtained from deep vens and is chierinated prior to use. Water for fire protection can bs pumped from the river.
,. ..n o 2.5-6 0003 078
P 12. The Philadelphia F.lectric Company in conjunction with 52 other utility companies owns and operates a plant as part of the Atomic F.nergy Commission Power Reactor Demonstratica Program. The Peach Bottom nuclear reactor plant is located along the vest bank of the Susquehanna River about 9 miles upstream
-from the Conovingo Dam. The plant is capable of producing a net electrical output of h0 megawatts under a normal thermal rating of 115 megawatts.
The plant draws cooling water from the river at a rate of h0,000 gauens per minute, and discharges it about 300 feet further dovustream. The radiological vastes , after proper treat =ent, are discharged to the condenser cooling vater outfall line. The demineralized vastes, backvash vastes, and sludges frem the water purification plant are properly neutralized, settled, and discharged to a storm sever and thereby to the River. All solid or consolidated vastes including containment oil and chemicals containing high solid concentration (i.e., caustic solutions and decontamination chemicals) are shipped off-site and not handled by the radiological vaste system. Also, vater is purified in a small treat =ent plant for potable, laundry, and other uses in the plant. ' dater for fire protection can be pumped directly from the river. The plant is located about hl miles downstream of the nuclear station.
- 13. The city of Baltimore has a water supply intake on the Sus-quehanna River which draws water from Conovingo Reservoir.
The intake is located about 1/2 mile upstream from Conovingo Dam, and A9 miles downstream from Three Mile Island Nuclear Station. Baltimore is permitted to take up to 250 MGD when the river flow exceeds 5,000 efs, but is limited to 65 MGD if the river flow is less than 5,000 efs. The Susquehanna River supply is an auxiliary source of supply, and is in addition to two other sources of supply, consisting of Lock Raven and Prettyboy Reservoirs on Gunpowder Falls, and Liberty Reservoir on Patapsco Creek. The combined capacity of all three reser-voirs is 86 billion gallons. System consumption averages 215 MGD and reaches a peak of 330 MGD during the su==er. Two vater treatment plants provide , complete treatment at a capacity of 360 MGD. I System storage of tree.ted water totals about 700 million gallons in reservoirs and tanks, and this incaudes vater for fire pro-tection. Under nor=al conditions, the system storage provides for about a 3-day censumptica of treated water. The three reservoirs impounding rav vater provide for LOO days of normal requirements. 1 1k. Conovingo Dam and powerhouse is located sbout 50 miles downstream from Three Mile Island Nuclear Station. The project is a hydro-electric generating station evned and operated by Philadelphia l , F.lectric C q any. "he total installed capacity of the station is 513,000 rl.
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The das creates Conovingo Reservoir, which extends 14 miles upstream, and has a storage capacity of 321,500 acre-feet. Potable water supply is taken from the reservoir to supply the generating station and 16 houses in Conovingo Village. The water undergoes complete treatment, including coagulation, sedimentation, filtration, and chlorination. Average consumption is 12,000 gallons per day. Storage facilities consist of 100,000 gallon storage tank which impounds water for normal requirements and also for fire protection. Under normal conditions the tank vill provide approximately 8 days supply. O. O 0003 080 pi , ' 0:!b 2.5-8
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' Docket 50-289 l Supplement No. 1 ;
October 2, 1967 QUESTION Provide the following: 2.6 2.6.1 An estimate of the population in the area h0 years after plant startup, including the basis for the projection; 2.6.2 A desc'ription of the preoperational environmental =onitoring program to be conducted; 2.6.3 A description of local airport takeoff and landing patterns with respect to the facility. ANSWER 2.6 1 Figure 2.6-1 presents the requested population estimate for the year 2011 (h0 years after plant start-up) for each , sector cut to 50 miles. These esti=ates werc =ade based ! on a study of the area by Metrcpolitan Edison Company , personnel. During this study developers and planning ! board. for local and county political units were contacted l in an effort to cbtain first-hand knowledge of future l trends in population growth and migration. l l Two trends are apparent: p a. The population of the anthractie coal regions is decreas
\s ,/ ing with migration southward to surrounding regions.
- b. The population of nearby city centers and boroughs is decreasing with migration to the outlying districts of these population centers.
The population figures reflect these trends and were obtaine by extrapolating the 1960, 1970,and 1980 population curves. These curves included appropriate gr wth factors and are considered to be conservative esti=ates. The estimates are also expected to reflect other fac crs such as the effects of birth and death rates. Another facter considered is the concentration of populatic: in the vicinity of highways. For exa=ple, the ccepletion of U.S. Feute c0 through the anthracite region could attract industry to :ne area resulting in a sicving of present migra tien cut of the area and a pcssible reverse of the present trend in later years. The possibility of such a nrend has been accounted f:r in the projection study. a
'--ad' ate area Of the site, su==er cottage popula:10:
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0003 081 L
2.6.2 The pre-operational environmental =onitoring prc.Iram for Three Mile Island will be conducted under the direction of, or by, Met-Ed personnel. The program is proposed to start in January, 1968. Atmospheric, terrestrial,and aquatic samples will be analyzed to establish background levels of radioactivity in the samples analyzed. Samples will be analyzed for gress alpha, beta,and gamma activity and if any significant amount of activity is found, the sample will be analyzed for specific radionuclides. The pre-operational environmental samples at the Three Mile Island site will provide a =eans of acco=plishing the follow-l ing objectives:
- 1. Determination of the background radiation in the soil, river water, air, and fish.
- 2. Identification of specific radionuclides as a means l of identifying the sources which contribute to environmental activity.
2.6.3 Olmsted State Airport has only one runway (130*/310*). Instrument landing approaches to 310 would align with the runway direction and the aircraft would pass approximately 7500 ft. NIE of the site. Aircraft on intending to land I on 310 could pass near or over the site prior to turning ) on final apprcach; however, this would not be a standard I VFR approach.
)
l The normal take-off pattern on 130 is away from the site, l 1.e.,the aircraft turn to the left after take-off. l Aircraft take-off and landing patterns in the other respec-tive direction are out of the site area. The missed approach holding pattern for O C ced State Air-port is also out of the site area. The Harrisburg-York Airport has three runways (120*/300*, 80*/260*,20*/200*). Instru=ent landing approaches to 30c would align with the runway direction and the aircraft I would pass approximately 1/2 =ile to the NNE of the site at an elevation of about 2300 ft. Aircraft on VFR intend-ing to land on 300 could pass near,or over,the site prior to turning on final approach. However, 300 is seldom used due to high terrain considerations and short length (h000 ft.). Occassionally strong crosswind effects on cther land-ing approaches require the use of this runway. Aircraft departing on 120 would normally start a right turn approxi+ mately 1-3 miles frem end of runway depending upon type of aircraft. O w w 2.6-2 0003 082
() , Aircraft take-off and landing patterns in the other respec-tive directions are out of the sita area. The Harrisburg-York Airport ha: one missed approach holding pattern. It is located such that aircraft would pass near the site at an altitude of rougnly 3000 ft. However, air-craft in the holding pattern would comprise considerably le:4 than 1% of all aircraft making, instrument approaches. Most aircraft having missed the landing would immediately be vectored by radar to make another approach. i d:) I l 4 O
+c m.t .r" 2.6 0003 083 I
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_-[ Year 2011 Population Estimate (10 Years After Plant Start-Up) 4
.. Miles Radius From Plant Site -~.
o-1 1-2 2-3 314 4-5 5-10 10-20 20-30 30 104 10-50 4 Total 13,878 N 30 108 4,130 8,950 660 25,800 10,800 11,300 4 2,100 4 3.200 70,378 1,381 m:E 118 97 226 470 470 12,600 20,900 5,100 2,100 10,70C- 52,781 1,397 NE llii 1341 1 31: 820 265 9,400 35,800 50,900 5,880 32,500 135,877 917 ENE 55 72 178 370 214 2 1,650 8,850 15,100 4 55,500 195,000 277,317 1,384 E 53 Sli 7 300 970 25,000 21,600 121,000 43,000 37,500 249,184 4 525 ESE 7 52 941 114 7 225 8,150 17,100 4 151,000 32,000 112,000 350,775 li o5 SE o 79 25 li i9 152 7,900 31,800 4 11,800 11,300 4 21,000 4 93,105
." 931 P SSE 125 206' 200 295 105 8,100 4 85,500 14,700 22,000 60,500 192,031 1,280 4
S 17 3 460 1,125 2,675 12,900 121,000 4 22,500 28,600 85,000 277,280 3,713 . SSW 13 6 80 1,180 490 1,920 7,900 78,000 77,000 30,000 57,000 253,613 1,902 SW 57 196 229 780 640 5,700 16,800 25,000 36,500 21,500 107,402 3 , 1 81 WSW 7 373 159 1,620 1,025 2,800 22,190 19,600 19,100 58,000 125,174 2,185 W 10 308 92 395 1,380 18,500 56,500 95,000 31,000 18,900 222,085 796 O NNW 33 36 270 187 270 26,500 157,000 17,500 14,900 1,800 221,496 4 O 2,384 O NW So 103 65 161 2,005 36,1:00 19a,000 1 8,14 00 8,375 14,500 261 ,059 4 U 1,629 4 a WNW 17 6 97 1,970 525 1,990 27,000 62,100 18,700 7,710 27,500 147,639 Co A O O O
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Docket 50-269 Supple =ent No.
- Cetober 2, 1967 QUESTION Describe in detail how the confini g nature of the Susquehanna 2.7 River Valley was factored into the long-ters diffusion estimates listed in Table 2 h.
ANSWER Wind data was available from two locations near the Three Mile Island site: Harrisburg-York State Airport and Olmsted Air Force Base. Harrisburg-York State Airport is located on a plateau above
' the river valley, with only one =ajor topographical obstruction to free air passage; the large ridge south of the field. Olmsted Air Force Base is located within the valley valls virtually at river 3
level. A co=parison of the seasonal and annual average vind roses indicates that some river channelling exists at 01=sted in the vest northwest and east-southeast directions, but this channelling effee is not significantly in excess of what is shown by the Harrisburg data due to the nearby hill. In evaluating the potential effect of the river valley on local atmospheric dispersion factors the horizontal crossvind dispersion was assumed to be li=ited by tne width of the valley walls. The cloud was allowed to disperse nor= ally until the valley valls vere intersected. The width of the valley was assumed to average approximately 1,000 meter and resulted it a =aximum value for r . of 163 meters. Thus, values of O- y were assumed to be x const(nt after intersection with the valley walls in esti=ating the long ter= diffusion as further described in Section 2.3.h of the PSAR. 1 I l l I l l n>
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2.7-1 0003 086
i Docket 50-289 Supple =ent No. 1 cotober 2, 1967 QUESTICN Clarity the design and =axi=us earthquake acceleration values to be 3.1 used. As a result of a recent conference call between us, our seis=ic design consultants and your consultant, Dr. Cornell, a
=odification to the response spectra was tentatively a6 reed on.
Provide the revised design basis which we understand to be the.t the El Centro response spectrus vould be utilized at those frequencies where it is = ore ccuservative than your original proposal. ANSWER Figure 3.1-1 indicates the revised e.cceleration response spectra, reflecting the greater respcase at lower frequencies based upon the 19ho El Centrg spectra. For other than a =cdal analysis the =athematical =odels will be subjected to the ground motion described as acceleration as a functi of ti=e. The input ground =otion shall individually be the 1957 Golden Gate Park, San Francisco earthquake and the 19ho El Centro earthquake both nor=alized to 0.06 g for the design earthquake end 0.12 g for the =axi=u= probable earthquake. The design vill satisfy both acceleregrams. b O 3 1-1 0003 087
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]\- Docket 50-289 Supplement No. 1
- October 2, 1967 QUESTION Provide the following with respect to the seismic design of the 3.2 plant.
3.2.1 The damping values which . rill apply to the maxi =um earthquake. 3.2.2 The criteria to be used for the critical pipi g systems with respect to stress loadings (for the =axt=um and design earth-quakes) and further elaboration on the loading ecmbination; to be used in the piping design. 3.2.3 A brief description of the stack (or vent) and whether failure could i= pair plant safety. 3.2.4 Information on the design of the cranes and in particular the support detail and =eans for ensuring stability during an earthquake. 3.2.5 Provide the following with respect to the seismic design of the plant. ANSVER 3.2.1 The damping values which will be applied to the maximum earthquake vill be as stated in Appendix 5A, Section 5, /
" Damping Factors," of the original PSAR.
( 3.2.2 Critical piping systems (i.e.,all Class I piping) vill be designed in accordance with the following criteria:
- a. Pri=ary steady state stresses, when ecmbined with the seismic stress resulting from the response to a ground acceleration of 0.06 g acting in the horizontal and C.0h g in the vertical planes si=ultaneously, are to be held within the allovable working stress li=its generally accepted as good design and as set forth in the code for POWER PIPING 331.1.0-1967.
- b. Pri=ary steady state stresses,when ecmbined with the seismic stress resulting from the response to a around acceleration of 0.12 g acting in the horizontal and 0.08 g in the vertical planes simultaneously, vill be limited so that no stress value vill exceed the mini =u= specified
=aterial yield stress, so that the functior. of the connected component or system vill not be impaired, and so that a safe and orderly shutdown of the plant can be
=ade.
3.2.3 The Reactor Suilding vent vill have a cross section of apprexi-
=ately 14 ft, vide and 3 ft deap. The vent vill be con-structed of steel and vill be securely anchored to the reactor
() building fcr its entire length. The vent vill be supported by the Reactor Building shell and will be designed as a Class I I l 3.2-1 e ;, , -,
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structure as specified in Appendix 5A, Section 1.1, " Class h l I," and shall =eet the require =ents of Appendix 5A, Section 2 " Class I Lesign 3ases." Failure of the plant vent due to any cause vculd not t= pair plant safety since release of activity through perforations or cracks vill still be governed by the ground level virtual source dispersion =cdel developed in Section 2 cf the PSAR. ANSWER 3.2.'4 Reacter Building Cranes The polar crane is a Class I cc=ponent and vill be designed along with its supncrtinz structure to satisfy the criteria for Class I structures and ec=penents contained in Appendix 5A of the Pre-li=inary Safety Analysis Report. In order to ensure stability during an earthquake the crane trolley vill be tied devn to the bridge and the bridge tied devn to the runway girder at all times during plant operation. Crane 3 rackets The preli=inary design nrovides for a standard WF-section used as a bracket, velded directly to a continueus bent plate ring that is thicker than the liner. (See Figure 3.2-1). The
=o=ent introduced to the bracket vill be resisted by two structural tees e= bedded in the vall. The shear vill be transferred frc= tne bracket web plate, through the liner plate, into the stiffener plates, and resisted by bearing of the structural tee web plates.
hj 3.2.5 Since the turbine hall is to be located on soil and the contain=ent 'essel en rock, the potential for different
=ove=ents during a seis=ic disturbance does exist. This condition could accentuate the deflections of che turbine hall with respect to the contain=ent vessel. The piping and interconnecting ele =ents in general, which includes tne
=ain stea= and boiler feed syste=s, are anchcred at the centainment vall at one end and in t5.e turbine hall structure at the other end. Extre=e deflecticns of one end with respect to the other end could lead to excessive stress in piping. Differential =0tions (settle =ent or rotatien) i between anchor points of piping syste=s located en two I
1 structures founded on different materials vill be deter =ined and the effect on the piping considered in a static solution. Resulting stresses vill be ec=bined with these for the dynadic (vihrstion) analysis. Where these values are not k=cun the piping vill be designed and vall thickness chosen l to obtain a vary conservative design. Basis for piping stresses is tae USAS 331.1.0-1967 PCWIR ?!?ING code. Allevable stress values specified by this code are based tu a safety factor Of four (=ateria' -d '-"-
- ensile strength divided by fcur). Added to this conservatis=, the allevable stresses vi'l be held 1:ver than nor=al. In this way the safety of the syste= vill be assured.
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Locket 50-239 Supple =ent No. 1 October 2, 1967 QtESTICN Discuss the effect which an assu=ption of an earthquake si=ultanect. 3.3 with the forces i=pesed by a less-of-coolant accident and other applicable leads vould have en the design of all syste=s and struct necessary to the preventien or =itigation of fission product reless In particular provide (1) the type and =agnitude of forces i=pesed by the maxi =u= or design earthquakes, accident, and si=ultaneous occurrence thereof, (2) the type of design =cdifications, if any, which would be required to cope with the si=ultanecus occurrence of the earthquake and accident, and (3) criteria with respect to stres levels and the =anner of Icad ec=bination which you vculd censider appropriate for the postulated st=ultaneous events. Of particular concern are the steam generator supports, vessel supports, vessel internals, contain=ent penetrations, and e=ergency coolant injectic systems. ANS'dIR The reacter coolant syste= is designed to withstand the nor=al design leads in addition to the design earthquake without exceeding allowable stresses in such critical ec=penents as piping, reactor internals, reactor vessel, steam generator and reactor coolant supports, e=erEency coolant injection syste=s, and structures internal to the Reactor Building and Reactor Ccolant Pu=p. These components, systa=s, and structures are also designed to sustain nor=al leads plus twice the design earthquake without loss of
\
function. Since the earthquake leading '.111 not produce a loss-of-coolant accident, it is concluded that loss-of-coolant accident loadings vill not occur si=ultaneously with the earthquake conditic The Reactor Building is designed for the st:21 aneous application of the leads due to the earthquake and less-of-coolant accident as defined by the pressure transient included as Figure IL h5 of the PSAR. The analysis of all other structures, cc=ponents, and syste: listed above vill be further develored to provide the infor=ation 1 l requested by this question. l \ 3 3-1 0003 393 l o.
O Docket 50-289 Supplement No. 1 October 2, 1967 i QUESTION Submit a turbine missile analysis which assu=es failure at an 3.h overspeed of 180% or above. 3.h.1 Show that the mass, shape and velocity of the missiles assume-to be produced by failure of the turbine are the vorst missiles fran the standpoint of depth of penetration of the structures er equipment required for reactor safety. 3.k.2 State the criterion for protection of ecmponents vital to a safe shutdown and in particular how the criterien vill be fulfilled for the control recm and the feedvater source. ANS'4ER 3.k.1 Refer to Section 5.1.2.7.2 which states that the turbine-generator supplier has =ade a study of the =ajor =issiles that =ight escape the turbine-generater exciter housing as a result of a hypothetic failure. The last stage wheel is censidered to b re the verst combination of weight, size, and energy. Table 5-1 summarizes the missile properties base < cn s speed of 186 per cent of the rated speed with 60 per cen-of the energy being absorbed before leaving the casing. The depth of the penetration of the missile into the Reacter Building is determined using =ethods presented in Nav. Docks () P-51*. No reductica in energy is taken for the penetration of the Turbine Building or intervening equipment. The depth of the penetration is shown in Table 5-2. 3.k.2 The ecmpenents vital to safe shutdevn are listed belev: control red drives nuclear services cooling water system decay heat re=cval system energency feedvater syste= centrol recm The control red drives are located within the Reactor Building which is designed to withstand the turbine missiles hypothesised in Section 5.1.2.7 of the PSAR. Tha control room will be a ecmpletely enclosed concrete st .ucture designed to withstand the same turbine =issiles. Tha decay heat removal system is ccmpletely embossed within the auxiliary and Reactor Buildings. The Reactor Building and that pcrtion of the auxiliary building housing the engineered safety features is designed to withetand the turbine =issiles. The nuclear services cooling water system is protected in the following =anner:
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Redundant river water pu=ps are located in a re=otely located screer and pu=p house. Redundant underground pipe lines with different reutings connect the river pu=ps with the heat exchangers. Redundant heat exchangers are located in a cenerete annex to the auxiliary building. Underground redundant pipeliner connect heat exchangers with nuclear serviess closed cycle pu=ps which are located in the auxiliary building. All services served by this syste= are located in the auxiliary building or reactor building. The e=ergency feedvater supply systa= is broken down below into two categories by location:
- 4. Inside the turbine hall
- a. Condenser hetvell - In the unlikely event that a tur-bine =issile occurs, the missile vould have to eject at an angle directed into the condenser neck where a
=ultitude of braces and extraction lines are located, pass thru a cc=plete tube bank,and finally penetrate the hetvell area if the energy re=aining is sufficient.
- b. E=ergency feed pu=ps - Two emergency turbine driven l l) feed pumps are located on the base =ent floor physically separated by auiliary condensers; these are redundant.
- c. Feedvater piping - Redundant pipe lines with separated routings connect feedvater scurces with the stea=
generators.
- 3. Cutside the turbine hall
- a. Cendensate stcrage - These tanks are located cutside the building; by utili:ing two (2) tanks, the loss of both tanks si=ultaneously frc= the same turbine missils becc=es extre=ely i= probable.
All of the above are predicated on a turbine =1ssile being credible. A turbine =issile penetrating the turbine casing is not probable. If such should occur, the chances of the
=issile striking one of the abcve cc=ponents becc=es further re=cte and the odds of da= aging a vital c=npenent even less.
~41th redundant pu=ps and feed lines and =ultiple feedvater sources (note no credit has been taken for the river water bcoster pu=p to the hetvell - see answer to L.12), sc=e cc=-
bination of the three vculd re=ain operable. g(, ) ;,! ,; 3.--2
Oceket 50-269 Supplement No. 1 , October 2, 1967 i QUESTION Are the cooling towers vital to safe shutdevn of the plant? Could 3.5 collapse of a cooling tower affect the safe shutdevn of the plant by damaging the switchyard? ANSVER The cooling towers are provided c=1y to cool condenser circulating vater and are not required for safe shutdown of the plant. The water level in the basin vill be approx 1=ately at grade in the cooling tower vicinity which is belev plant grade,thus preventing inundating the plant in the event of a basin rupture. Even in the unlikely event that both 230 kv switchyard buses and the 115 kv safeguards line were damaged by =issiles and subsequent 1; faulted, the plant can be safely shut devn with the on-site diesel engine generators. The safeguards buses which these generators feet veuld be autcentically isolated from the external system by indcor, missile prctected circuit breakers. j O l 1 1
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i l l 1 3.5-1 : 0003 096
(s) Docket 50-289 Supple =ent No. 1 Cetober 2, 1967 QUESTIOD Please revise the PSAR to incorporate the results of your more detaile h.1 reactivity calculations, which you su -ized in our =eeting of August 8, 1967. h.1.1 For each of the reactivity worths of Table 3-h of the PSAR, give the expected range of "ariation due to uncertainties in measurements or calculations. Discuss the basis for the ranges given. k.1.2 What is the effect of initial reactor operation with the great-est expected value of positive moderator coefficient en the reactivity control distributica listed in Table 3 k. h.1.3 What is the basis for the specification of the excess centrol red worth of 1.6% ak/k over the holddown require =ents of 5.k% ak/k. ANSWER The folleving reactivity control distribution revisions reflect the change of the 16 fixed shim rods to =ovable centrol rod assemblies, and the results of more detailed 2-D calculations. h.l.1 Each of the four basic groups of Table 3-k are discussed below.
- a. Controlled by Soluble Berog The items in this group are controlled by soluble bcron, the total holddcvn varying with operating conditions and core life. The basic safety parameters, i.e., control red worths and the moderator temperature coefficient, as listed in the PSAR have been generated at boren levels in excess of that expected. These evaluations were =ade at hot, rated pcuer conditions as listed in Table 3-6 of the PSAR, i.e. ,1,860 ppm boren. This level, when ecmpared with Figure 3-1, is approxi=ately 2h0 ppm higher than ex-pected for maxi =um boren at the start of the first cycle.
ine resulting rod worths are lover, and the moderator tem-perature coefficient is more positive than would be ex-pected. . Analysis of various experimental data regarding reactivity levels indicates a possible uncertainty of ap-proximately 15 ak/k. The excess reactivity of approxi-mately 2.h5 ok/k as represented by the 2h0 pps boren c.bove illustrates basic design concervatism.
- b. Centrolled by Inserted Centrol Pod Assemblies The reactivity value specified for transient xenon cen-trol was set by core =aneuvering requirements. The par-O ticular goups of rods selected for this purpese vill be v
m wm s.1_1 0003 097
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. I l
chosen such that the resultant power peaking, for inser-tion or withdrawal, will not exceed design values , nor vill the value of any one of the rods exceed the value used in the ejected rod safety evaluation. Uncertainties in peak renen as associated with the activation of this bank should affect only the relative =aneuverability of the core. Analysis of experimental rod vorths indicates that the uncertainty associated with the calculation of the tran-sient bank vculd be relatively s=all, and under the selec-tien criteria stated above would affect only the relative core =aneuverability similar to the peak.xenen.
- c. Centrolled by Movable Centrol Rod Assemblies (1) The basic uncertainty in the Doppler deficit and the associated Doppler temperature ccefficient is the fuel te=perature. Vari &tions of as =uch as 2300 F have been investigated, although uncertainties of 2200 F are considered reascnable. This maxi =us vari-ation of 2300 F results in a 20.3% ak/k sving in the Doppler deficit, and a variation of less than 21.0 x 10-6 (ak/k)/F in the Doppler coefficient.
(2) Variations in the equilibrium xenon and the equilib-rium xenon centrol bank vorth are ec=pec sated for by the soluble boren. Uncertainties in the equilibrium xenon are considered as part of the initial reactivity ) uncertainty previously discussed. The same criteria set forth for the transient xenen band (item 2-a) vill be applied in the selection of the equilibri'tm bank. (3) The moderator te=perature deficit, which resul;- from the negative =cderator te=perature coefficient during power changes frer tero to 15 per cent rated power near the end of wre life, varies primarily with end-of-life fuel con lition. Calculations for various core cycles indi a maximum variation of 220 per
- cent. As an adi cal conservatism in this centrol
! balance, an uncertu_nty of =0.25 ak/k, i.e. , 233-1/3 l per cent, was considered. (k) The regulating and dilution control rod bank operates between the 75 and 95 per cent withdrawn positions. Any uncertainty in the value of this bank as cperated within these li=1ts vill be taken up in the soluble boren dilution frequency. (5) The shutdevn =argin of 15 Ak/k is a nini=us require-ment which is a= ply covered by the
- 4-"- available verth as illustrated in the ic11cving Paragraph d-(L).
(c) The tal =cvable centrcl verth required is then stated censervatively as L.0 : 0 55 ik/k. The varia-tices results frc: Paragraphs c-(1) (:0.35) plus c-(3) (:0.25).
'. e' J
-a 0003 098
- d. Available Centrol Red Assembly Worths (1) The total C3A vorth has been calculated for various core conditions and reflects an allevance for the fol-loving vorth-reducing effects:
(a) Boron level (b) Spectrum changes (c) Core cross section variation (d) Control poison burnup (e) Hot, rated power, operating uncertainty Although some of these effects are a function of core lifeti=e, all are considered to be in effect from the beginning of Cycle 1. Therefore, the total worth of 10% ak/k represents a conservative =ini=um. (2) The individual rod worths (stuck or ejected) are taken at time zero, and first cycle conditions, and reflect only the first of the five reducing effects listed above. The boron level does not have a strong effect, and the resulting rod worth represents a near =aximum value. (3) The available CRA vorth of 7.0% ak/k is a minimum cb-tained by subtracting the maxi =um stuck rod worth from a minimum totalI .attern verth. Therefore, the movable CRA vorth avai.able of 5.6% ak/k also repre-sents a conservative miai=um. (k) Comparison of the maximum required =ovable rod vorth (h.5% ak/k) to the =inimum =cvable rod worth available (5.65 ak/k) shows an excess of 1.1% ak/k. h.l.2 As stated in Paragraph a of the &nswer to Question h.l.1, a maximum boron level was the basis for the various evaluations of Table 3 h, thereby reflecting a uaximum moderator tempera-ture coefficient effect. h.1.3 The excess control rod verta of 1.65 ak/k was specified as a basis for conservatism in tl:.e reactivity control balance of Table 3-h. This is illustrated in Paragraph d of the answer to Question k.l.1 above, and results from the =inimum movable CRA vorth available (5.6%) less the total nominal movable con-trol vorth required (.k.0%). Pertinent to this Question h.1 ve supply replacement pages (3-9, J-10, 3-11, 3-12, 3-13, and 3-lk) and figures (3-1 and 3-6) which are included as part M Ahendment 2. o V
' -3 0003 099
5 () Decket 50-289 Supple =snt No. 1 October 2,1967 QUESTICC You design criteria for reactivity shutdown capability (PSAR Sections k.2 1.4.8 and 1.k.9) specify a =inimum shutdown =argin at a steady-state reactor condition. h.2.1 What is your criterion for a =ini=us shutdown =argin during operational transients? Refer to the recently published Geno eral Design Criteria 27 to 30 for guidance in for=ulating your criterion. k.2.2 Calculate the =ini=um shutdown =argin for the less-of-coolant flow and loss-of-electric power transients , using assu=ptions which result in the =ost reactive cenditi'ons. h.2.3 Which of the accidents described in the PSAR results in the
=inimum shutdown =argin and what is the range of the =argin for this accident for the =est probable and =ost adverse (con-servative) conditions?
ANSWER h.2.1 Minimum Shutdown Margin - Oterational Transients The reactor is designed to =eet the criterion that it can be shut down to the hot suberitical condition with a =argin of at O' least 1% dk/k with one control rod stuck out of the core. The evaluation of operational transients - such as moderator dilu-tion without red motion, loss of pumping power, and rod with-drawal - has shown that this =argin is not changed by these transients, because the reactor returns to the hot suberitical condition at the end of the transient. This =argin at the hot shutdown condition also provides sufficient shutdown reactivit3 to keep the reactor suberitical in accident-induced transients which cool the reactor coolant to lower te=peratures, such as a steam line failure. 4.2.2 Minimum Shutdown Margin - Loss of Coolant and Loss of Power _ The less-of-coolant flow and the less-of-electric-power tran-sients are essentially identical in terms of core response. The core -ccolant te=perature rises a few degrees initially and then falls off very slowly, asymptotically approaching a limit of Sk2 F. The table shows the negative reactivity =argin avai: able during the loss-of-coolcat-flew transient. O ' tU 0003 100 h.2-1
Reactivity Martin During the Loss-of-Ccolant-Flow Transient O Reactivity Margin, 5 ak/k Time, seconda 3ot got 10 -5.h -5,5 100 -5.6 -5.0 Steady state (Sh2 F and equilibrium Xe) h.1 -3.h 4.2.3 Minimum chutdevn Margin - steam Line Failure A cident The steam line failure accident described in the 1h.1.2.9 of the PSAR results in a mini =um shutdevn margin inase.uch as this accident leads to the greatest cooling of the reactor coolant system below the hot shutdown te=perature. The coolant te=- perature drops to a minimum of 537 F. Using the most negative temperature coefficient which occurs at the end of core life, the minimum reactivity margin at this reduced temperature is
-2.3% ak/k with the most reactive rtuck rod. If all rods were inserted, the reactor vould be 5.3% ak/k suberitical. Even in the event that the reactivity of the core were such that is just met the 1 per cent suberitical margin at 5h2 F vith one stuck rod, the suberitical margin at the minimum temperature following the steam line break accident would be -0.85% ak/k ,
i vith one stuck rod, and -3.85% ak/k if all rods were inserted. O
.o. 4i 0003 101 l
h.2-2 l
i f_/ s Docket 50-2s9 Supple =ent No. 1 October 2, 1967 QUESTION Submit DN3 rst:.os for the unit cell, side and corner channels as cal-k.3 culated by the W-3 correlation for the vorst co=bination of nuclear and engineerirg hot channel factors. ANSWER The DNB ratir.s for the unit cell, side and corner channels are shown in Table h.3-1 and h.3-2. All DNS ratios have been calculated for lik per cent rated pcver. Table k.3-1 presents the nominal case with a mixing coefficient of 0.03, which one would expect in smooth tubes. The W-3(2) correlation for DNB analysis utilizes the most recent modification for unheated vall effects. The results for a postulated vo st case ecndition are shewn in Table h.3-2. Frr this condition, it i: assu=ed that the channel flow area is reduced to the minimum value ever its entire length instead of a statistical deternined a=ount; the highest nuclear peaking, which occurs with twice the nor=al sp.scing between assemblies,is used; and the sixing coefficient is reduced by a factor of three. Mixing acros the perforated vall of the assembly has not been taken for credit in either c;.e. O
'- 5uelear power and engineering hot channel factors used in this analys are reported in 3.2.3.1.1 of the Three Mile Island Nuclear Station PSAR.
Table h.3-1 Nominal Case (a = 0.03) Cell Tyre DNER(3&W) H. Stu/lb DNER(W-3)(1) H, Stu/lb DN3R(W-3)(2) H, Stu/ Corner 2.20 648 1.75 65h.2 1.85 '660.9 Wall 2.11 6h8 2.09 65h.1 1.89 660,9 Unit 2.01 6h8 1.89 660.6 , 1.89 660.6 Table h.3-2 Postulated Worst Case _(a = 0.01) Cell Tyre DNBR(3&W) 3, 3tu/lb DNBR(W-3)(1) H, Stu/lb DNER(W-3)(2) H, Stu/ Corner 1.70 692 1.k9 681.1 1.3h 700.1 Wall 1.65 688 1.73 685.8 1.38 696.5 Gnit 1.73 668 1.h6 692.h 1.k6 692.h ('~% I s '. .' l [ f
- 1
\,,) h.3-1
FERENCES O L) Larsen, P. S. , et al. , DNB Measurements for Utvards Flov of 'Jater in an Un-heated Souare Channel with a Single Uniformly Hented Rod at 1600-2300 PSIA, Third International Heat Transfer Conference, Volu=e II. August 1960.
- 1) '4eisman, J.,'denzel, A. H., and Tong, L. S., Exterimental Deter =inatien of the Derarture frem Nucleate 3 oiling in Larre Rod Bundles at Hi;th Pressures, AIChE Paper Presented at Ninth National Heat Transfer Conference, August 1967 O,
l l 0003 103 g s ;. ' w o., u.3-2
Docket 50-239 Supplement No. 1 October 2, 1967 QUESTICN Discuss in detail the scope of the folleving research, develop =ent, k.h er test progrs=s including projected cc=pletion dates for varicus phases of the progra=s and test equip =ent descriptions. To the ex-tent possible, resv.lcs of the progrs=s to date should be stated. h.h.1 Ther=al design, f neluding DN3 and flev distribution. (Will loss of a core F.arrel check valve be si=ulated in the flov tests?) k.h.2 Centrol rod drives. h.h.3 Steam generator including blevdevn tests. (Diseu.ss the de-sirability of insulating or otherwise maintaining the shell at a high te=peratu e to si=ulate the ther=al transient that might be experienced in the actual generator during secondary system blevdevn.) h.h.h Cora barel check valves. (Discuss the progrs= for testing the valves or a scaled prototype under operational and ac-cident flew and temperature conditions including vibrational effects during operation and mechanical forces during blev-V(N devn.) h.h.5 Material tests at high burnup. (Discuss which =aterial prop-erties are critical, the results expected and the manner in which the results will be used. Could significant data of a confirmatory nature be obtained by re=cving and testing fuel from the reactor environment at intervals in the future. If other test programs, currently in progress, are relied en for fuel rod failure =echanisms, describe the secpe and schedule of these tests and ce= pare your requirements in detail.) h.h.6 List important milestones in the design of the facility, such as when the core design must be frozen. ANSWER h.h.1 Thermal Desi.rn l
- a. Decarture frem Nucleate Boilinz Heat Transfer Investigation i
In the late fall of 1961, The Babcock & Wilcox Cc=pany began the j design and construction of a large heat transfer facility for I the purpose of doing DN3 testing at pcVer reactor operating con-ditions. In this facility, which is located at the B&W Research Center, Alliance, Ohio, testing over a vide range of variables covering practically all of the situations one might expect to encounter during nor=al and expected transient operation of wate: p cooled reacters is possible. The facility is supplied with 1.8 V , =egawatts of d-c pcver and a fully autc=ated data acquisition system. It can be operated within the folleving limits: gi s a h h_1 0003 104
Pressure - 100 to 2,700 psia O Inlet subecoling - 20 to 250 F Mass velocity - 0.2 x 106 to 3.5 x 106 lbs/hr-ft2 in a 9-red as-se=bly. Present specimen size rod asse=bly with a heated length of six feet. DNB detection with thermoccuples (resistance =easure=ent back-up) Flev - 150 gp= at 295 ft head Since the loop has been ec=pleted, a variety of experiments have been perfor=ed to gain better understanding of the DNB phenc=e=a and to develop e=pirical relationships necessary for the design of water reactors. A=cng the experiments cc=pleted to date have been testa en: (1) Single tubular specimens vith coth uniform and nonunifor= power distributions. Nenunifor= axial peak-to-average powers as high as 1 9, si=ulating inlet and outlet peak lo-cations, have been included in the tests. These tests were conducted as a function of shape, length, and syste= para =- eters. On the basis of these tests, power shape factors for , application of the test results to reactor design have been i determined, and it was concluded that simulation of reactor axial power shapes could be achieved with confidence in test bundles of shorter length than actual reactor fuel assemblies. (2) Annular speci= ens with various ec=binatier.a of inner and outer vall heat generation and nenuniform axial power dis-tributions were also tested. It was determined that the results obtained with the annular data correlated very well vith data taken en bundles. Analytical work done on the tubular and annular specimens has formed the basis for the bundle size and power generation shape to b'e used in a fu-ture test bundle described belev. (3) A 9-red test asse=bly with a uniformly heated length of six feet, si=ulating the reactor fuel red diameter, pitch spac-ing, and spacer grid details, has been tested. This was the first test approaching actual gec=etrical conditions as well as operating conditions for the core. Of principal interest were the effects of spacer grids , cold valls , inter-channel =1xing, and instabilities. Data frc= these asse=- blies are still being analyzed, and work is progressing on a new ONB correlation. Results to date indicate that the analytical =ethods used in the design of the reacter core are conservative and that no critical areas exist. l The Sabcock & Wilecx Cc=pany intends to =aintain an active and O l aggressive progrs= in the field of ENB heat transfer. Sc=e of . K' '< . s,, 0003 105
O the principal progrs=s which vill be conducted in the near future are cutlined below along with a tentative schedule: (1) A 9-rod asse=bly with the same rod diameter, pitch spacing and spacer grid used in the fuel assembly vill be tested. A nonuniform axial pover generatica profile vill be e=picyed over six feet of the bundle length. The pcVer profile vill be representative of that partion of the core experiencing the most severe heat transfer conditions and the most prob-able location of a DNB. Testing of this assedbly is sched-uled for the first two quarters of 1968. (2) An annular specimen with nonuniform pcVer distributica on the outside tube vill be tested for additional verification of the effects of length and nonuniform pcver generation. Power may be supplied to various portions of the specimen so that length effects up to the full 12-fcot long test regica of the speci=en may be examined. Testing for the annular specimen is expected to begin in the fourth quarter of 1968. (3) A 9-rod bundle test e= ploying nonuniform radial power dis-tributien vill be tested in 1969 A definitive program and schedule for this series of tests is not for=ulated.
, (h) Depending upon the results obtained frcm the previcus tests, 4
) additional tests vill be devised as part of the continuing basic heat transfer and core optimization program. Tests under consideration are for additional radial and cxial power distributions, larger test assemblies, investigation of different grid designs, and transient simulation.
- b. Mixing Studies Related to the studies for DNB are additional programs conducted to determine the degree of mixing in the fuel rod channels. Flev tests involving a h-rod assembly have been conducted to deternine f mixing effects. Flow tests en a =cekup of the outer two rows of '
fuel rods and the can panels of two adjacent fuel assemblies have been conducted to det--mine the friction effects at the perforated l vall boundary. These tests have confir=ed that the larger flev ' cells at the periphery of the bundle ec=pensate !cr higher equiv-alent fricticn adoquately. These effects are shown nuterically in response to Question h.3. Additional tests to extend the in-vestigation to larger sizes, and more elaborate gecmetry for the purpose of confir=ing the analytical model and value for =ixing coefficients, are described belev. (1) A 9-rod mixing test assembly, of the same bundle geometry as . the DNB bundle described previcusly, has been constructed to I determine the degree of mixing present during the DU3 inves-g-s tigations. Testing with this assedbly is currently in pro-( ,f gress and is expected to be cc=pleted in Dece=ber 1967. s.-mo t i!:ro h.u-3 0003 106
(2) A 16-rod asse=bly with the si=ulated juncture of four per-O forated, fuel asse=bly cans meeting at the corner is under construction. Testing with this asse=bly vill enable one to determine the degree of a1xing which occurs between fuel asse=blies , and vill give more detailed infer =ation on ve-locity distributiens and mixing in the peripheral cells of the fuel asae=bly than did the h-rod tests. The current core analysis considers only mixing within a fuel asse=bly and does not take credit for mixing external to the assem-bly. It is expected that testing with this asse=bly vill begin in January 1968. (3) A facility large enough to accept a 6h-rod assembly is cur-rently under construction. Tests for this facility are not yet fir =, but it is expected that sc=e of the preliminary work for calibration of in-core ther=occuples and pressure differential instru=entation vill be dene in this facility. Initial plans vece to construct a low pressure facility large enough to accept a full si:e, cross section fuel as-re=bly. This has currently been replaced with the 6k-rod assembly, and its need vill be re-evaluated. Testing in this facility is scheduled for the first quarter of 1968.
- c. Vessel Model Flev Tests A 1/6 scale =odel of the reactor vessel, the internals, and the reactor coolant piping frc= the pu=ps to the reactor vessel is currently being tested at the Research Center. Portions of the
$ I reacter vessel and internals are constructed of transparent plas-tic to facilitate visual observation of flev patterns within the vessel. The reactor core is simulated in the model with individ-ual fuel asse=blies censtructed of perforated sheet =aterial and calibrated crifices at the top and bottc= of each asserbly. Pres-sure sensors and thermocouples are provided in all fuel asse=blies to determine the flev distribution at the core inlet. Additional pressure sensors and ther=ccouples are provided in other pertiens of the vessel and core so that overall mixing and pressure drop determinations may be =ade.
Preliminary investigations in the =cdel and the analysis des-cribed in the answer to Question 5 1.5 indicate that it vill not be necessary to si=ulate the interna 2 check valve cunstruction, operation, or any =alfunctions in the vessel =cdel flov tests. Testing should be ec=pleted in =id-1963. L.L.2 Centrol Red Drives
- a. Ce=renen: Tests h e pur;cse of this progra: is to seek cut ;ctentisi =aterial and/or desi.n proble=s prior to production unit testing. Se Oc=;cnent tes pr:grs= :::sists Of (1) Evalua:1:n :f vari:us grades of '.,rs;hitar-bearing =sterials in an aut:: lave a: 1,6:: psi, 60:: ?, a.i va:er :he=istry A '
6 _.___ 0003 107
O V with 13,000 ppm H 3 Boy. - The bearing materials are statically loaded against a 17-4 PH shaft such that the developed stress is greater than vill be present in the actual centrol rod drive. (2) Environmental dynamic gev and bearing tests under loads equivalent to the control rod drive operating conditions.- In this. test, the bevel gears, the pinien, and the bearings supporting these gears are being tested in an autoclave at 2,250 psi, LOO F, and reactor water chemistry. The purpose of this test is to obtain vear characteristics of the gear
=aterial ccmbinations and projected life of the bearings.
(3) Si=ulated drive test.- A cc=plete =echanis=, which simulates the drive with the exception of its overall length, is aeing tested under no-flev reacter operating conditions of : Jpera-ture and pressure in an autoclave. An accelerated wear and life test through a short stroke vill be co=pleted in cen-junction with the life-testing of the prototype mechanism. (h) Autoclave testing at reactor operating te=perature and pres-sure of buffer seal, splines , and bearings.- In this test, the spline joints of the drive rod assembly are being tested under static, no-load conditiona for corrosion. g-') (5) Autoclave testing of shortened drive rod assembly under static as_/ load conditions. - This test is similar to the spline testing with the addition of the bevel gear set at the lever end of the drive red. (6) Autoclave testing at reactor temperature and pressure of the bevel gears, bearings, shortened rack, and pinien gear under vibratory loading of the rack to determine the fretting characteristics of the gear train. - This test is a static load test.
- b. Full Scale Prototyre Testing Under No Flev Ccnditions This test vill be perfor=ed in an autoclave per=itting full strek-ing in rocm temperature water and at reactor operating conditiens of temperature, pressure, and water chemistry. The cold tests vill be utili:ed only as an initial checkcut of the drive prior to temperature and pressure testing. The control rod vill be simulated with a du==y weight. Misalign=ent will be introduced to note its effect on vear and overall perfor=ance of the drive rechanism.
- c. Full Scale Prototyte Testing at Reseter C;eratine Ccnditions of Temperature, ?ressure, and Flev The full life test program as defined in the PSAR vill be con-
, ducted under this test. A prototype centrol rod and fuel asses-
- f ('s_/~') bly vill be used in order to establisa the ec=plete drive train sssembly. 0003 108 ((, ' m u o- u.u-3
Inasmuch as the rack and pinion drive concept described in A=end-O ment 1 is sc=ewhat different frem the first rack and pinion drive tested at the Research Center, the test program which has been cutlined above provides an extensica of previous tests to estab-lish verification of drive perfor=ance and adequacy. The previ-ous test program verified the basic =aterial selection, the snub-ber design, and the buffer seal concept for use with a rack and pinion drive. The components test program (items a and b) is scheduled for ec=- pletion by the end of December 1967 Hot icop testing (item e) at full flev conditions of the prototype drive vill begin in Octo-ber 1967, and continud until the end of December E67 h.h.3 steam cenerater The basic steam generater test pregram is discussed in detail in Appendix kA of the Duke Power Ccepany PSAR (Deckets 50-269, 270, and 287). In addition to the testing described in the above reference, secondary system blevdown tests have been carried out, and a reactor coolant (pri=ary) side blevdown is planned wher. sched-ule ce==it=ents permit. The hot water facility of the Research Center is shared by the Control Rod Drive Tests, the Steam Gen-erator Tests, and other experiments and tests. Three secondary system blevdown tests have been ecmpleted. The results of these tests have de=enstrated the integrity of the steam generator under conditions of rapid depressurization and large (greater than 200 F), tube-to-shell temperature dif-ferentials. In addition, the results of these tests are used in the devel-op=ent and verification of analytical =cdels for ster.m system blevdown analyses. The construc' tion of the test stess generater (including insu-lation) is such that the thermal ti=e constant of the shell is lover than that of a full-scale unit. This lover time con-stant results in more rapid cooling of the shell during steam system blevdown than vculd occur in a full-sice unit. The primary side blevdown test will provide temperature con-ditions which si=ulate a ther=al transient greater than that for the full-scale unit secondary blevdown as well as simula-tien of the ther=al transient for primary blevdown. k.k.4 Core Barrel Check Valves , The core barrO che:k valves vill be designed to relieve the , pressure generaw M by steaming in the core folleving the LCCA i so that the core vill remain sufficiently ccvered. The valves . I vill also be designed to withst'and the forces resulting frcs 80i .' u l 4.h-O 0003 109
O rppture of either a reactor coolant inlet er cutlet pipe. Testing of the valves vill consist of the folleving:
- a. A full-site valve asse=bly (seat, locking mechanism, and socket) vill be tested at steady-state conditions at the
=aximum pressure expected to result during the bicvdown.
- b. Sufficient tests vill be conducted at zero pressure to de-termine the frictional leads and clearances in the hinge assembly, the interia of the valve cover, and the deflic-tions resulting frca impact of the cover so that the valve response to cyclic blevdown forces =ay be deter =ined analytically.
- c. The valve assembly vill be pressurized to determine what pressure differential is required to cause the valve to begin to open. A determinatica of the pressure differen-tial required to open the valve to its maxi =wn open posi-tien vill be simulated by =echanical =eans.
- d. A valve assembly vill be installed and removed remotely in a test stand to judge the adequacy of handling equip-
=ent.
Since the temperature differential existing across the valve
, (~T assembly during normal operation in the reactor is only ap-Os / proximately 55 F, and since the same material is used for the valve seat, socket, and cover, there is no need to conduct tests at elevated temperatures.
The valves are located in a region of relatively lov velocity and turbulence, and preliminary analysis indicates that there is insufficient energy in the coolant to cause vibrational problems. Therefore, no testing to prove the vibrational ade-quacy of the valve is planned. Testing should be completed by January 1969 l ru~ ) l l 0003 110 l l .
.' ; ; i ',
b*b"I 3
O h.k.5 High Burnue Fu,el Tests The design of fuel rods for pressure :ycles and ther=al gradi-ents are a= enable to analysis , based en cut-cf-pile properties. In deter =ining the behavier of =aterials under the influence of accu =ulated irradiation the properties of interest are ura-niuc. dioxide greuth rates under restraint by tubular cladding, and the ability of the cladding to absorb strain without fail-ure at re ctor operating conditions. A detailed report of sources of infor=ation for the irradiation of clad and fuel has been presented in the PSAR, 3.2.h.2.2 plus references. In addition to the PSAR references , irradia-tien of fuel assemblies or partial fuel assemblies with Zire-alcy-clad UO2 is in progress in the Saxton and Big Rock Point reactors. These data vill demonstrate the behavier cf fuel asse=blies under the ec=bined effects of irradiation, pressure cycles , ther=al gradients , reactor coolant environ =ent , and fuel-clad restraints. B&W is conducting a progrs= to cbtain a better understanding of fuel growth rates and irradiation effects on cladding, the influence of hydrogen en cladding, and fission gas release at high burnup for the specific design burnup pro.jected for peak pcVer regicns in the reactor. The fuels irradiation progrs= will test fuel specimens at de* sign te=peratures and at exposures in excess of those obtained in the fuel rod. "'he specimens irradiated to the design burn-up are scheduled to be cc=pleted in =id 1969, well in advance of reactor operatien. The progra= will provide infor=ation en the swelling rate of UO2 as a function of burnup, dencity, heat rate, and cladding restraint. Fuel specimens vill be operated at heat rates up to 21.5 kv/ft, which is in excess of the peak specific pcVer in the c0re. The burnup will range up to 75,000 WD/TU. The fuel rods vill cperate with a cladding surface temperature Of 650 F. A progra= has been carried out to deter =ine the effects c' irradiation On the =echanical properties of lir:alcy k. k Tests were 00nducted to te=peratures as high as 775 F. The l su==ary of results '-- "* s pregrs= is as foll:vs : , l
- a. *he rec = te=perature tensile and yieli strea.gths of lir:-
j alcy-- increased with ::a1 neutron ex;0sure f r irradia-tion te=peratures up t: 650 7. S e rate Of increase was greater a: 1:ver irradiati:n te=peratures. his increase , in strength was a:::=pa=ied t-/ a decrease in the :::al and
"-d'--
el:ngsti:ns.
- b. ~he r::= te=perature y'a*4 a-d tensile strengths Of the spe:i= ens irradiated a: "5 7 vere s:=evha 1:ver than th:se Of the spe:i= ens ber:re irradiati:n. .nese ::anges ,
Id{i!j b . .'
" 0003 111
O in properties, hcvever, were not significantly different from these observed in speci= ens aged cut-of-pile for like periods of time.
- c. The roc = temperature unifor= elongation values for both annealed and cold-verked =aterial vere approximately 2 per cent after neutron irradiation at 130 ? to an exposure of h.5 x 1019 nyt (E > 1 Mev).
- d. A difference in irradiation behavior was noted for the longiudinal and tranr/erse speci:: tens , particularly after irradiatien at 775 F. At this temperature the tensile strength in the transverse direction continued to increase whereas in the longitudinal speci= ens the strength decreased.
A sum =ary of capsule specimens is given in Table h.h-1 following, and a tentative schedule is presented in Figure L.k-1. The folicsing is a description of the research now in progress at B&W that is related to the current reactor design: Material Irradiatien Testing Pregra: The purpose of this prcgra= is to deter =ine the effects of irra-diation en the core cc=penents of a central station pcVer reac-A tor. The progra= is divided into three tasks: Task I Lov Burnup Fuels Irradiation Task II Zirealoy k Irradiation Task III High Burnup Fuels Irradiatica Task I - Lev Burnue Fuels Irradiation The pri=ary objective of this task is to investigate the dimen-sional stability of pellet-type fuel rods when irradiated at current and future PkR cperating conditions. The program consists of capsules, sc=e of which are designed
. to operate at 21-1/2 kv/ft. The cladding for these capsules will operate at a surface temperature of about 6h0 F. All of , the capsules will be irradiated, when possible, for one com-plete cycle of the 3AWTR. Under normal operation, this vill amount to about 25 EFFD and a burnup of about 3,500 to h,000 MWD /TU.
The irradiation of capsules, initially operated at a heat rate of 25-25.7 kv/ft, has been ec=pleted. Some capsules received as =uch as 609 pcver cycles at 22.8 to 2k.6 kv/ft. Hot cell exa=inatien is underway. 73 I ! %d 1,. 0003 112 Ll' [i;l b h.L-3
Task II - Zirealcy h Irradiations The Zircalcy h cle.dding in the core operates with cutside and inside surface temperatures as high as 650 and 300 F, respec-tively. A progras was therefore designed to determine how the sechanical properties of Zircalcy-h are affected by irradittien at these temperatures. Longitudinal specimens cut fres 0.k25-in. diameter Zircalcy k tubing are used to determine the preperties in the icngitudinal direction. Ring specimens and flattened rings conforming to dimensions of the longitudinal specimens are used to determine the properties in the transverse direcu en. Sc=e of the tensile specimens were charged with 250 to h00 ppm hydrogen prior to irradiation. Irradiation of the two 300-day capsules is continuing without any operational difficulties. As of June 30, 1967, these cap-sules had achieved an exposure of 306 EFFD. Task III - High Burnue Fuel Irradiations The pri=ary purpose of the High Burnup Program is to determine the swelling rate of UO2 as a function of burnup using fuel rods of the same design as the core. In addition to deter-
=ining the swelling rate, the effect of several other variables incluuing the density, heat rate, and cladding restraint vill )
be investigated. The data in the literature indicate that UO2 swells during irradiation to the extent of 1.5 to 2.0 volume per cent for each 10,000 WD/TU burnup. These data vere obtained from flat plate elements in which the fuel temperature did not exceed 1600 C. Unfortunately, it is not kncun whether these results are applicable to other operating conditions er fuel shapes. The program consists of capsules sece of which will cperate at a heat rate of 18 kv/ft and others at a heat rate of 21-1/2 kv/ft. The pellets, other than U-235 centent, will conform te the reactor fuel specifications. The burnup will range frcm l 10,000 to 75,000 WD/TU with six capsules exceeding h5,000 ( WD/TU. The capsules vill not operate with an external pres-sure. Ecvever, two different cladding thicknesses, 0.015 and k
- 0.025 in. vill be tled to vary the restraint offered by the cladding. The fuel rods vill operate with a cladding surface temperature of 650 F. The diametral gaps between the pellets and cladding vill vary frca h-5 to 7-8 mils, te give s= eared l densities of about 92.3 and 90.3 per cent, respectively. These gaps and s= eared densities are censistent with the fuel red specifications. The insertion date for the first capsule was Septe=ber 5, 1967 e
c, % 0003 113 h . u-u
O The tests are oriented tcvard the deter =inatien of the behavior of materials in an irradiation environ =ent and to determine the optimum gecmetric and material properties for the specific ap-plication. The information is essential for advancement of the art, but is not considered critical in the sense that all of the programs must be completed to insure safe cperation. Renoval and testing of fuel taken from operating reactors at intervals during operation is not censidered necessary. The data on hand, plus programs which are currently under way, should satisfactorily provide the information necessary for assurance of safe operation within the limits required. 'O n U 0003 114 L.k-11
O O O Table 4.h-1 BW liigh Burnup Program - Capsule Fuel Test Burnup Heat Rate Identification Diametral Clad Irradiation MWD /TU Fissions /cc Irradiation Tim Initial. Final, Cap, Thickness, Capsule Fuel Hod Facility x 10-3 x 10-20(1) calendar months ) kv/ft kv/ft mile mils B-1 B-1 RS-3 10 2.5 4 18 17 5 k-5 25 B-2 2.5 4 17 5 7-8 25 B-3 2.5 4 17 5 7-8 15 B-2 B-4 RS-5 20 h.46 7 18 16.9 Powder 25 B-5 50 7 ' 16.9 k-5 25 B-3 B-T RS-6 30 75 11 18 16.1 4-5 25 B-8 7.5 11 16.1 7-8 25 B-9 7.5 11 16.1 7-8 15 Bk B-10 RS-6 45 10.05 17 18 14.9 Powder 25 B-11 11.25 17 14.9 k-5 25 c-B-5 B-13 RS-5 55 13.75 21 18 14.1 4-5 25 f B-14 13 75 21 14.1 7-8 25 M B-15 13.75 21 14.1 7-8 15 B-6 B-16 RS-3 65 14.50 2k 18 13.3 Powder 25 B-17 16.24 24 13.3 7-8 25 B-T B-19 RS-4 75 16.73 28 18 12.5 Powder 25 B-20 18.75 28 12.5 7-8 25 B-8 B-22 RL-2 25 6.25 10 21-1/2 20 5 4-5 25 B-23 6.25 10 20 5 7-8 25 B-2% 6.25 10 20.5 7-8 15 B-9 B-25 RL-2 50 12.5 18 21-1/2 19.3 4-5 25 B-26 12.5 18 19.3 7-8 25 B-27 12.5 18 19.3 7-8 15 B-10 B-28 RI,-2 75 18.75 28 21-1/2 17.8 T-8 25 O B-29 18.75 28 17.8 T-8 25 O B-30 18.75 28 17.8 7-8 15 .O U flased on 200 New per fission. _' ~ - (2) Based on 80 per cent reactor efficiency. W
-_ m__. _ _m _ _ _ _ _ _ _ _ _ _ _ - ._____m _ _-
h.4.6 Schedule The items in the design R&D schedule which follow are deter-mined principally by commit =ents which must be made to ecmply with =anufacturing and erection schedules. To a large extent. the foregoing R&D programs are confir=atory in nature or have been established to optimize the system rather than to estab-lish feasibility or limits of safe operation. Many of the programs, such as the irradiation program, heat trgnsfer pro-gram, and hydraulics program, are of a continuing nature and the dates given indicate the last time when the results could affect the reactor design. (1) Vessel model flow test June 1968 (2) Operatica of a fuel assembly under re-actor conditions of pressure, tempera-ture (isothermal), flew, and coolant chemistry June 1968 (3) Fuel and clad irradiation program June 1969 (h) Heat transfer and related hydraulics programs June 1969 (5) Centrol rod drive line testing under simulated reactor conditions January 1968
'O (6) Completion of check valve design and tests January 1969 (7) Ccmplete confirmation of steam genera-tor performance to permit fabrication release. (The size and configuration of the steam generator have alr.2dy been determined.) Dece=ber 1967 (8) Blovdown tests of steam generator pri-mary (reactor coolant) side have not yet been scheduled. --
RlTERENCE (1) Mechanical Properties of Zircaloy 4 After Irradiatica at 130, 650, and Q 775 F, TP-299, April 1967 k.k-13 - 3 '1b
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(m- Docket 50-289 Supple =ent No. 1 October 2, 1967 QUESTION What size are the steam lines and the steam generator cutlet no::les? k.5 Discuss the effect of stes= line size and arrangement on the conse-quences of a guillotine break of a steam line. ANSWER The =ain steam leads vill be 2k-in. Schedule 60. The steam genera-ter cutlet no::les will be 2L-in. nc=inal diameter (22.125 in. ID). The arrangement of this piping is shown sche =atically in the attached Figure h.5-1. The rupture of one of the 2h-in. lines at ultimate core pcVer (2,535 MWt) would initiate a turbine trip on icv reactor syste= pressure or high flux within six seconds. The accident characteristics during these first six seconds until turbine trip and isolation occurs are nearly identical to those previously presented in the Duke Pcver Cc=pany PSAR Supplements 2 (questiens 6.3 and 6.h) and h (question 16.1), Dockets 50-269, 270, and 287. After turbine trip (at six seconds after rupture) the steam flevs and ecoling rates are less severe than previously reported because of the smaller line size available for leakage to the atmosphere, i.e. , one 2h-in. line as ec= pared to cne 28-in. line used in the referenced analysis. Therefore, the previously presented analysis
/ is a conservative evaluation for the Meted reactor.
(\_ The conclusions for a steam line rupture analysis in the Three Mile Island Nuclear Station are as follevs:
- a. The accident will produce a maximum cooling rate of 3 F/sec, and and total reactor coolant te=perature decrease of approximately h2 F occurs. The affected steam generator feedvater flew is auto-
=atically reduced to 7.5 per cent upcn turbine trip, and is ter-minated by protective controls or operater action in about ene minute.
- b. Using stuck red conditions and the maximum end-of-life condition for the =cderator coefficient (-3 x 10-k ak/k/F), the mini =um shutdown =argin ir 2.35 ak/k.
- c. The resultant stresses , developed in the steam generator tubes are less than half of the yield strength of the =aterial. Therefore, a stes= line failure vill not i= pair the integrity of the steam generator tubes.
- d. The effect of long-term continuation of feeduater flew at 7 5 per
! cent was evaluated, and the tube temperature remains belev the j steam generater shell te=perature. The resultant tube stresses ( sre less than the yield strength of the material. Therefore, no tube fails even if feedvater flow is net reduced to match the de-cay heat rate. (O) u.5-1 0003 !l8
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l 1 Decket 50-289 Supplement :fo. 1 October 2, 1967 QUESTION Discuss the effect of dropping control rods without snubber action. 4.6 What procedures ensure detection of gas accumulation in the snubber? Could timed control. red drop tests be perfer=ed which would detect I the absence of snubbing action and are these planned (for exa=ple, l following a shutdown)? ANSWER The ecmplete absence of any snubbing action of the control red drive
=echanism en a trip action, i.e. , free fall drop, is censidered to be incredible. For such an action to occur under reactor operating cen-ditions, it would be necessary for the rack housing to be void of any ;
water frca the vent cap to the bottom of the snubber cylinder. This ' void could be established only by a bubble of air and/or gases of 590 - in.3 I Operating procedures require that all drives be vented during fill-up of the reactor vessel with reactor coolant. If a drive vere not vented as a result of an administrative error, the entrapped air vould be ccm-pressed to a volume of 14-in.3 at operating pressure. This volu=e is
=uch less than the 590-in.3 volume required to fill the snubber with gas.
7 Gas generation by radiolytic decomposition during operation is pre-( vented by the hydrogen concentration in the water. However, even if a drive were filled with water centaini:.g no hydrogen, the gas bubble generated for a year of operation vould be smaller than the ik-in.3 bubble which might occur from entrapped air. Thus , gas generation during operation could not lead to a situation in which the water in the snubber was replaced by gas. This analysis demonstrates that neither entrapped air nor gas genera-tien could lead to loss of snubber action. Accordingly, neither pro-cedures to detect gas accumulation in the snubber, nor testing to de-tect gas ac:u=ulation in the snubber, nor testing to detect the ab-sence of snubbing action are considered necessary folleving the reac-tor shutdown, and nene are planned. 1 1 1 1 8 l (~'J s_
' 6-0003 120 1
i i O Docket,50-269 Supplement No. 1 October 2, 1967 QUESTION To what extent vill in-core ther=occuple data be relied on during h.7 operation? Describe the location of the in-core thermocouple junc-tion, the expected accuracy of water temperature measurements, and experimental confirmation of the accuracy and applicability of this
- type of measurement.
ANSWER There vill be no incere thermoccuples in reactor. I l I J l
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1 Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Discuss the possibility of control red ejection due to a rod drive k.8 seal or vent failure. ANSWIR The control rod asse=bly including drive shaft weighs in excess of 250 pounds in water. This assembly has a 2.h-in.2, flat projected area which could be available for a piston effect if the pressure in the control rod housing vere lover than the pzessure in the reactor vessel. A pressure differential in excess of 100 psi vould be re-i l quired to cause outward motion of the control rod. The possibility of a control rod ejection due to a rod drive seal failure has been evaluated by assu=ing that the buffer seal asse=bly j offers no resistance to flow fro = the reactor vessel to the at=o-sphere. Figure 3-51 of the PSAR shows that between the reactor ves-sel and the upper seal asse=bly that there are a nu=ber of cross-sectional area changes which in effect provide a series of pressure breakdown mechanisms. By utilizing these restricted flow arets, sufficient breakdown of the pressure occurs such that less than 20 lb/see vill flow from the reactor vessel into the rod drive housing. This flow will produce a pressure differential of approximately 20 psi which results in an additional (over operating) upward force on the control rod of only h8 pounds. This force is a factor of h below (hs //~') that required to produce control rod motion. As can be seen en Figure 3-51, the vent at the top of the rack hous-ing has a very small cross-sectional area available for leakage flow to the atmosphere in the event of a failure of this co=ponent. This 1/h-in, dia= vent line has an area of 0.05 in.2 as co= pared to 0 975 in.2 available for supplying flow into the rack housing fre= the reactor vessel. With this large area difference, a maxi =cs pres-sure drop of 5.5 psi vould be developed across the control rod. This pressure drop is not sufficient to cause control red notion. It is therefore concluded that neither a failure of the od drive seal nor of the vent piping on the rack housing vill produce a suf-ficient pressure differential across the control rod to cause control rod motion, and therefore a rod ejection accident cannot occur as a result of failure of these components. h.8 1 i
1 Docket 50-269 l Supple =ent No. 1 Cetober 2, 1967 O QUESTION Consider supplying feedvater to the steam generator by operating h.9 the electrical condensate pumps frc= the on-site diesel supply as an alternate to the e=ergency steam driven pumps. ANSWER One condensate and one condensate booster pu=p together develop caly 550 psi differentia', pressure. Since this is not adequate to pump water into the steam generators against full pressure, these motor driven pumps cannot be considered an alternate to the emergency steam driven pu=ps. The main turbine driven pumps, which normally boost the feedvater to full steam generator pressure, are driven by condensing turbines which are inoperable during " blackout" due to less of condenser circulating water. Even if the condensate-condensate bocster pumps were adequate pressure-vise, the diesel generators could not carry both. Only one condensate pump can be carried by the diesel generators and this only if both diesels start and after they have been manually ,5 paralleled electrically. The condensate pump alone vill develop 180 psi differential pressure. Two (2) 5% capacity turbine driven pu=ps are provided, each supplied frem a separate source, which offer the required re-dundancy from full steam generator pressure to the temperature when decay heat pumps may be started. 0003 123 O h.9-1 (Revised 12-??-67)
I l O Docket 50-289 Supplement No. 1 October 2, 1967 l l QUESTION Discuss the time limitation imposed by the station battery in h.10 sustaining heat removal systems during a postulated loss of all on-site and off-site AC power. Include a description of the loads which must be carried and the planned technique for shedding normal battery leads. ANSWER Following a complete loss of all a-c power, auxiliaries and controls i of an emergency feed pump may be i= posed on one battery for a period of two hours, along with other plant emergency loads. Battery loads and the period for each load used in calculating the size of each of the two batteries are listed below: Description Period k 1/2 KVA inverters
- 2 hours Engineered safeguards d-c buses 2 hours All switchgear and miscellaneous control 2 hours Reactor building emergency lighting 2 hours Control room emergency lighting 2 hours Emergency bearing oil pump 1 hour Seal oil backup pump 1 hour O Boiler feed pu:rp turbine emergency oil pumps Reactor coolant pump bearing oil pumps 10 minutes 1 hour
*The inverter loads are as follows:
Control rod drive position indication Primary system instrumentation Auxiliary system instrumentation Steam supply system instrumentation Integrated control system Nuclear instrumentation and reactor protection Radiatica monitoring (52 channels) Ccmputer Control room annunciator Miscellaneous recorders Since each battery is sized to carry all essential emergency loads for 2 hours, even in the unlikely event of a failure of one battery, there is more than adequate time to restore a-c rever. If both batteries are available, each would carry approximately half the load with a resulting lengthening of the discharge time. Each of the turbines for the generator and boiler P' ed pumps is provided with zero speed detection. It is intended that this be used for indleation at the control board that the corresponding oil pump car be manually stopped. The reactor coolant pumps are
operation, there is sufficient heat and water available to =aintain the secondary syste= pressure above the design pressure of the reac-ter building (55 psis) for at least 10 hours folleving the accident. The reactor building pressure, however, has been reduced to approxi-
=ately 2 to 3 psig in about two hours by operatien of the engineered safeguards syste=s. Therefore additional time vill be available be-yend 10 hours when the pressure differential vill prevent leakage.
If it is conservatively assumed that 10 hours after the less-of-cool-ant accident, the pressure differential in the stes= generators no longer prevents leakage to the at=caphere, the leakage rate frc= this path would be equal to 0.025 per cent per day of the reactor building at=csphere. This leakage rate is 12.5 per cent of the total allev-able leakage rate of 0.2 per cent per day. If these leakage rates are assumed to be additive (i.e. , 0.225 per cent per day), deses can be calculated using the sa=e fission product inventories and environ-mental release models for the MHA discussed in ik.2.2.k of the PSAR. The 2k-hour dose to the thyroid at the exclusien distance would be 290 re=. The 30-day dese to the thyrcid at the 2-=ile icv population zone vould be 80 re=. The 2-hour dose to the thyroid at the exclusion distance vould not change. All of the above deses are within the allevable limits of 10 CFR 100. This de=enstrates that special previsiens to acec==cdate si=ultanecus l leaking steam generater tu'ces and leaking safety valves at the time l cf a less-of-ccolant accident are not necessary. , 1 0003 138 ! ~>
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Docket 50-289 p Supplement No. 1 \ October 2,1967 QUESTION Discuss the ability to flood the primarf cavity including (1) the 55 level and size of the overflow drains and (2) the cavity volume. ANSWER The reactor building sprays located in the containment building dome discharge uniformily over the operating floor. The discharge falls into the refueling cavity on the operating floor'and in the steam generator cubicles. The water falling on the operating floor drains to the refueling cavity and down the stair well. All
. water collecting in the refueling cavity drains through 2 ft - 6 in drain lines to the reactor vessel cavity. During normal plant operation these lines are open and protected by screens against debris. During refueling operations these lines vill be blanked close with a blind flange.
Water collected in the reactor vessel cavity will overflow to the steam generator cubicle through the small annular space approximate: 2 in. around the primary coolant pipe penetrating the pri= arf shielt Water collected in the cavity will achieve a level approximately 3 f t above the top of the core before it overflows to the steam generator cubicles. Two 1 in. drain lines in the reactor vessel cavity, one at the reactor vessel support and the other in the reactor in-core instrumentation trench, drain to the reactor building sump. Under no conditions is there sufficient head on these drain lines to impair the ability to flood the reactor vessel cavity. The rate of water supply to flood the reactor cavity far exceeds the capacity of the two 1 in. drain lines.
'Ihe reactor vessel cavity has a volume ;f approxi=ately 10,000 f t3 ,
This volume is approximately 90 percent of the primary coolant inventory and less than 20 per cent of the borated water storage capacity. O 5 5-0003 139-
Docket 50-289 S.upple=ent No. 1 October 2, 1967 Q,UESTION Discuss the route which containment spray and injection water must 5.6 take to reach the su=p. Indicate the size and location of drain lines to the su=p, the criterien for sizing the drains and the
=ethod to be used to prevent plugging of the drains and su=p.
ANSWER Except for thosedescribed in the answer to question 5.5, there are no special drain lines provided to handle the flood of spray and injection water flowing to the reactor building su=p following a loss-of-coolant accident. Extra large openings,provided for more normal ressons,are relied en to carry this water to the sump with a minimum of holdup at any location within the building. The discharge from the reactor building sprays is uniformly dis-tributed over the refueling ficor and fuel transfer pool, and the vented tops of the stea.n generator cubicles. Water drops into the fael transfer pool, frem there via the special drain line to the reactor cubicle, then to the steam generater cubicle and finally, frcm the steam generator cubicle to the base =ent flcer level and the reactor building sump via the labyrinth ac.eess open-ing. The building spray water which falls within the steam generator cubicle vent openings falls to the ficor level of the steam generator cubicle >where it then flows to the reactor build-ing sump via the labyrinth access opening between the steam /[^ w generator cubicle and the basement area in the reactor building. The remainder of the water from the reacter building sprays falls on the opsrating floor and flows to the sump via the stairwell openings on the east and west sides of the reacter building. The water being injected above the core will ficw out of the ruptured pipe to the reactor or steam generator cubicle, depending on where the pipe rupture occurred. If the rupture occurred in the reactor cubicle, the water will flow cut of the annular open-ings to the steam generator cubicle and from there to the su=p via the labyrinth access cpening. If the rupture occurred in the steam generator cubicle, the water will only have to flow through the latter opening to reach the su=p. The reactor building su=p is provided with a rectangular mesh screen. This screen will be so sized to provide several times the required flow area to the suction of the decay heat re= oval and reactor building spray pu=ps. The mesh screen will insure the continued cperability of these pumps by preventing debris frem plugging er entering the pump suction. O 5.c-1 0003 140 [4*; 4 i<,-
The special drain line between the deep end of the Fuel Transfer g Pool and the reactor cubicle will be protected from plugging by a W protruding screened box similar to, but =uch smaller than, that over the reactor building su=p. Normal drain lines, from various locations within the reactor building to the su=p, will have their openings screened to prevent items of debris inadvertently being carried via them into the sump. l k . I l l l l O f'b l ' li!!i 5.o-2 nan 3 141
Docket 50-239 Supplement No. 1 Of October 2, 1967 QUESTICN What volume of water is necessary (in the sump area) to provide 5.7 the required NPSH cf the recirculation pu=ps? We believe tnat the su=p should retain a high enough colu=n of water to provide the required head without the necessity of a large water inventary in the containment. ANSWER Per Fig. 6-3 in the PSAR, 13 25 ft H2 O NPSH is required for the Decay Heat Removal Pu=ps at 3000 gpm each. Per Fig. 6-7, 15 ft H2 O NPSH is required for the Reactor 31dg. Spray Pu=ps at 1500 gym each. Preliminary' calculations indicate that a =aximum friction head loss in the suction piping to these pumps is 33 ft H2 O when only~ ! one of the two 12 in, diameter suction lines is carrying the full 6.000 gym recuired to =aintain one decay heat removal and two R. B. ; spray pumps at full capacity. Detailed calculations of friction hea losses and the NPSH available to each eump vill be performed when manufacturer's drawings of the eauinment. to be provided. are avail-able and the final piping lavout has been ecmpleted. Based on an initial partial pressure of air in the containment building atmosphere of 13.75 usia. 32 ft of H2 O head vill be available to these pu=ps from this source. Based on the layout s drawings.15.5 ft of static suction head vill be available to the
- s. Decay Heat Removal Pumns and 16 ft of static suction head will be available to the Reactor Building Spray Pumps when the vater level j in the reactor building sumn is 2 ft below the basement floor level in the reactor building. With this water level in the reactor building sume. the NPSH available *o the one Decay Heat Removal Pumn on the line is (32.0 + 15.5 - 33) = 14.5 ft H2O and that avail-available to the two Reactor Building Spray Pumps on the line is i
( 32 + 16 - 33) = 15 ft H20. The NPSH renuired for the Reactor ' Building Spray Pumps is eaual to the NPSH available. Because of the preliminary nature of the information on which the above calculations are based. it is estimated that the NPSH available to both the Decay Heat Removal Pumps and the Reactor Building Spray Pumns will be greater than that required when the vater level in the reactor building Sume is full tt :he ficor level. In all of the calculations no credit has been taken for the :20t reactor building eressure, during which time these pumes vould be in oceration. 1 i (^)
'% /
j l l 5.7-1 (Pevised 1-6-68)
I i Dceket 50-289 : Supple =ent No. l October 2, 19t QUESTION Provide an analysis of local pressure forces i= posed by pri=ary 5.8 coolant piping breaks within the pri=ary cavity. 5.8.1 What is the largest break which the pri=ary cavity can withstand (the pressure transient and criterien for failure ) should be included). i 5.8.2 What is the largest break si:e possible within the cavity or shield, what piping restraints vill be provided and what pressure transient and loading does this i= pose. 1 5.8.3 Perfor= a si=ilar analysis of local pressures resulting 1 j free a break outside the pri=ary cavity. 1 ANSWER An analysis of this pre 1g= vas perfor=ed by =eans of the latest version of the CONTIMPT l' Cc=puter Progra=. Data vere generated by the I3M 7Chk Cc=puter for various main ecolant pipe rupture si:es (0.k. 1.k. 3.0, 8.55 and ik ft2) and a range of relittf areas. A con stant back pressure (pressure in the containment building) of ik.7 psia was assu=ed. The preliminary results of this solution are shev in Figures 5.8-1 and 5.8-2. 5.8.1 Figure 5.8-1 presents the preli=inary results of the =ax1=u= pressure in the reactor vessel cavity for varicus pipe ruptur f sizes within the cavity. These results are based en the pre- l li=inary pri=ary shield gec=etry. The transient pressure l buildup in the reactor vessel cavity for a ik ft2 rupture I is shown in Figure 5.8-2. The vent area for this analysis was 225 ft2 This vent area was assumed created by blowing cut the seal ring. Actual vent area available vill be deter-
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=ined at the ti=e final shield gec=etry is established and {
the seal ring design is cc=pleted.
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5.8.2 The pri=ary cavity will be designed to withstand pressures consistent with final shield gec=etry. If necessary, piping j restraints vill be provided to limit rupture areas ce=patible 3 vith design.
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5.8.3 The =ini=u= relief area which =sy be provided in the sten = generator cc=part=ent is 910 ft2, Based upon a IL ft2 rupture blevdown, Figure 5.8-3 indicates that the =aximu= pressure would be 15 psi. he transient pressure buil:iup for the 910 ft2 relief area is shavn in Figure 5.8 k,
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O V Docket 50-289 Supple =ent No. 1 October 2,1967 C.UESTION Discuss the reasons for not providing check valves in the recircula-5.9 tion line between the su=p and pu=ps and in the containment spray line upstream of the thiosulfate tank to prevent backflev. ANSWER Recirculatien Line A check valve in the recirculation line between the su=p and pu=ps would prevent draining the borated vater storage tank to the reactor building su=p in the event that both a borated vater storage tank cutlet valve and a reactor building su=p outlet valve vere opened si=ultaneously. Since this involves the autc=atic cpening of cer=al closed valves , the accident is censidered incredible and check valve are not required. Centainment Sersy Lines A check valve is not considered necessary in the contain=ent spray lines upstres= of the thiosulphate tank to prevent backflov because of the physical relationship that vill exist between the borated water storage tank header and the thiosulphate tank outlet lines g and the spray pu=ps. This vill be provided by connecting the thio-sulphate tank outlet lines t.o the header between the borated water stcrage tank outlet lines and the spray pu=p sucticn. In essence, the flow frc= the berated vater storage tank will enter the pu=p suction header and flow in two directions: In ene direction, it vill flev to the decay heat pu=p suction and in the other direction it will flew to the reactor building spray pu=p sucticn past the connection frc= the thicsulphate tank. Because the building spray pu=p capacity is =uch greater than the flev frc= the thiosulphate tank, no thicsulphate flev vill eccur in a direction opposing the ner=al flew and censequently, cc thicsulphate can flew to the decay heat pu=p suctien. This flev pattern would prevail no =atter hev
=acy spray pu=ps or decay heat pu=ps are cperating.
1 s 0003 148 i 1 s i.' '- l
Dockot 50-289 Supplement No. 1 October 2, 1967 O QUESTION Diseass the physical isolation provisions for the recirculation 5.10 pu=ps for protection against flooding including provisions to isolate tne leak. ANSWER Each of the two Decay Heat Re= oval Pu=ps and the two Reactor Building Spray Pumps will be isolated in indiviaal cubicles within the pit area of the Auxiliary Building. Each of the four cubicles vill be provided with a small sump having an overflow detector and annunciator alarm (located in the control room). The drain line betveen each of these cubicle sumps and the Auxiliary Building Sump vill be of small dia=eter to accommodate mail leaks (of little signifigance to the continued operability of the equipment in the cubicle) without alarming. A larger diameter drain line vill extend above the floor level in each cubicle to some pre-determined level below the bottom of the lowest vulnerable point on the motor ',*iving the pump in the cubicle. This larger
. diameter drain vil? prevent water from a major leak within the cubicle from backing up within the cubicle to a level that vould endanger the pump motor. The Auxiliary Building Sump Pumps vill be of sufficient capacity to handle the largest leak considered credible within the Auxiliary Building Pit Area with a slow draw-down rate on the sump water level.
Since there are no valves in any of the cubicles, the only credible sources of leaks considered therein are pump packing and flange p gaskets of the pump suction and discharge. All pipes entering and ? leaving each cubiele vill do so at a relatively high elevation within the Auxiliary Building Pit area. This provides an added degree of security from flooding in that considerable water storage capacity is available in the bulk of the Auxiliary Building Pit Area before water could overflow into the individual cubicles. The capacity of the Auxiliary Building Pit is approximately 1700 ft3 /ft of water level. With the present arrangement approximately 17,000 ft3 of water could be held up within the Auxiliary Building Pit area before it would overflow into any of the cubicles. A 1000 gpn leak, without any sump pump-out, would require over 2 hours to flood out any cubicle. Waterproof doors provide ready access to each cubiele for inspection of equipment while preserving the leak-tight integrity of the cubicle. The inlet lines to each of the four pumps have remetely operated isolation valves just outside of each cubicle to permit rapid remote isolation of a badly leaking unit. The discharge lines from , each unit are automatically isolated by check valves, also located just outside the cubicle. These are backed up by manually operated outlet isolation valves. O
- 5. w-1 0003 149
() Decket 50-289 Supplement No. 1 October 2, 1967 QUESTICN Provide the preliminary design for the fan coolers. In particular, 5.11 incicate the geccetry and heat transfer coefficients that will be utili:ed to ensure that the units are conservatively designed to remove heat in the accident environment. ANSWER The design details for the fan coolers will be developed by the manufacturers of this equipment.. In ner=al practice, design details become available only after an order for equipment is placed. - The purchase specification prepared for obtaining proposals frcm manufacturers will include as a requirement that the ability of the proposed equipment to perform be shown by theoretical analysis and substantiated by test data. When this information becomes available, we will be able to indicate the geometry and heat transfer coefficients that will be utilized to ensure that the units are conservatively designed to remove heat in the accident environment. When the information beccmes available, it will be submitted to DRL as information. O
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l l 1 ( (12) l 5. u-1 0003 150 { e
l Docket 50-289 Supplement No. 1 October 2, 1967 QUESTICN '4 hat :riterion is proposed with respect u removal of engineered 5.12 sateguards : mponents for =al.ntenance? For example, vould the plant ba shut devn if one high pressure pu=p vere unavailable for use sinet a single failure :riterion ::ald not be met in the event of an acci-dent? ANSWER All engineered safeguard systems contain sufficien. redundancy so thz any active rotating ::mponent =ay be removed for mair.tenance. Hov-ever, it may te necessary to take additional action to assure the availability of suffi:1ent :apacity to handle the desi.1;n basis condi- l tion if a unit is taken out of service for =aintenance. The addi- ; tional action might in:lude an increase in the test frequency of the remaining equipment cr operati:n in a standby mode. The specific ac-tion to be taken in.each case vill be determined during the detailed design and specified in the technical spe:1ft:ations. The Station vill not be shut down in the event of maintenance on one makeup thigh pressure $ pump. If one of the pumps is shut down, the other two full capacity pumps vill he available for emergency oper-ation. In the case of rea: tor building cooling safeguards, two full-capacity independent systems are available fer maintaining the reactor build-ing below design pressure. In ci.is case a combination of components from the two systems :an adequately provide the required cooling ca-O M pacity, and permit removal of a :cmponent from either system for main tenance. Two systems are provided for control of fission products following th accident, i.e. , the reactor building sprays, and the penetration pres surization and fluid ble:k system. Either of these systems is capabl of protecting the public in the event of a hypothetical accident. Th spray system :entains two redundant pumps. If one of these pumps is out of service for =sintenance, either the other pump er the penetra-tion pressurization and fluid block system will assure public protec-tion in the event of an a:cident. O s . m. --1 0003 !SI l
W (~s Decket 50-289 \s,) Supp]e=en: No. 1 October 2, 1967 QUISTION Provide the felleving with respect to the thicsulfate injection sys-5.13 tem. It development progrs=s are required in any of the following areas, specify the secpe and schedule of the programs. If other re-search progrs=s are to be relied on to develop required infor=ation, c:= pare the scope and schedule of these progrs=s with your require-
=ents in detail.
5.13.1 Discuss the interaction of the thicsulfate with the centain-
=ent, reacter, and recirculation system environ =ent under accident conditions.
5.13.2 Discuss the stability of the thicsulfate solutien over 1:ng time periods in the acciden envire=nent, including the effe-of radistica and the for=ation of byproducts in the solutien 5.13.3 What is the ability of the thicsulfate spray to re=cve other than ele = ental iodine frc= the contain=ent at=csphere includ-ing iodine en particulate =atter and = ethyl icdide? 5.13.k What is the effect of condensate en the cutside of the spray drop en the = ass transfer rate of the iodine? rs Q) 5.13.5 Discuss the redundancy provided in the spray syste=. 5.13.6 Provide the basis for the injection rate of the thicsulfate solution for varicus primary syste break si:es. Fcr exa ple if the spray syste: runs at full available capacity but the core injection syste= is required at a 1:v =akeup rate ( for e small break), would the thicsulfate be exhausted before the borated water s:Orage tank vas e=ptied and the re:ir:ulation
=cde begun? Cculd this result in a period in which water without thicsulfate was sprayed in 0 the centa --- **d E S'4IR 5.13.1 In 9raction The thicsulfate solutiens should not react signifi:antly with any of the surfaces prcpesed for the centa*--- - (reacter building), the reacter, er the recirculatien syste=.
The centain=ent surfaces are painted carben steel and painted concrete; both paints will be resista= to -ka -*'d -k -*-al attack of the alkaline thicsulfate s=lution. If the paintei surfaces are scratched er the pain : verage was in:enplete, the thi: sulfate solution still vill n0: result in a : rr:si:n proble: because carben steel is. generally resistant te sodiu= thicsulfa:e and slightly alkaline pH's. (Iarben steel pre-cessing vessels are s:=e:1=es used f:r the =anufa::ure Of 30-ilu= thicsulfa e. ' e- ?.ea:: r syste= sur'a--* - -'--= ':y , st ainles s st eel, and km ,s) ::::nel. The re:ir:=lati:n sys - * ' -"- spray syste= are c ,; - og, . . . . (f1 5 si i!' #* ()003 1 E>2
stainless steel with the exception of the carben steel spray lh nozcles. All of these materials are resistant to attack by the sodiu: thicsulfate-beric acid solution over the applicable pH range of 7 to 10. Research and develop =ent pr gra=s are in pregress to confir= that, under postaccident conditions, there are no har=ful in-teractions between the thiosulfate spray solution and the =a-terials of construction. Elevated temperature corrosien tests en several materials and paints ec==enly used in reactor sys-te=s are planned in the C?3L Spray Technology Progrs=, and 3MI is presently testing the interactions between several paints and likely spray solutions. These programs vill be reviewed to establish that acceptable corresien rates exist under postaccident conditions and that the paints vill not create any hacardcus interactions. 5.13.2 Thicsulfate Solutten Stability The stability require =ents for the thicsulfate solution over long time periods in an accident enviren=ent are
- a. The co=penents of the solutien =ust remain che=ically and physically ec=patible.
- b. The solution =ust retain adequate capacity for iodine re-
= oval and retention.
- c. The decc= position products =ust not result in excessive pH changes, excessive a=ounts of solid precipitates or l i excessive gas formation, or in any way reduce the concen-tration of the soluble poison dissolved in the solution.
Initial test results, which are su==ariced in the folloving paragraphs, indicate that alkaline sodiu= thiosulfate vill satisfy all these require =ents. Sodiu= thicsulfate acd beric acid are chemically quite cc=- patible. The addition af sodiu hydroxide er other suitable base results in the fer=ation of a pH-bcffered sclution which yields greacer chemical and radiation stability. It also significantly increases the iodine and = ethyl iodide abscrp-tica capabilities. Sufficient iodine re=cval capacity is
=aintained throughout the accident period.
Although the buffered solution undergces sc=e radiclytic de-cc=pesitien, the products of this decc= position do not ad-versely affect the icdine re=cval efficiency, the overall perfor:s:ce of the icdine re=cval syste=, or the cooling ca-pacity of ei-"- *ka e-- gency Core Cccling Syste=s er the Reacter 3uiliing Spray System. Results Of preliminary tests at OR5L vere similar te 3 W re-sults, but additional research and develep=ent is necessary to ec= fir: the test results under conditiens which =cre ac-curately st=ulate accident enviren=ent.
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. 3.u-a 0003 153
l fs 3oth B&W and the CRNL Spray Technelegy Progra: are presently (,) engaged in stability experi=ents. Althcugh final acheduling of all experi=ents is not ec=plete, it is anticipated that the scheduling of the ec=plete progrs=, which is secped to i answer all questions pertinent to stability in an accident ; enviren=ent, vill be ec=pleted shcr:1y. The experimental ! phase vill be perfor=ed on a schedule which vill make the final data available in ti=e for inecrperatien into the sys-te= design. 5.13.3 Sersy Re= eval Ability Iodine is expected to exist during !GA primarily in four forms (a) elemental iodine, (b) hydrogen iodide, (c) particulates, and (d) Organic iodides (of which =85 per cent is methyl iodid Hydrogen Iodide The rate at which elemental iodine is removed frc= the centair
=ent at=csphere is discussed in detail in ik.2.2.3.5 of the original PSAR. Hydrogen iodide, which is a very reactive chemical species with a high solubility in vater and aqueous sclutiens , is re=oved by a gas phase-centro 11ed, = ass trans-fer =echanis=, as is the elemental iodine. Therefore, the equatiens developed in lb.2.2.3.5 apply equally as well to the hydrogen iclide. And, since the hydrogen iodide has a higher diffusivit? than ele = ental iodine, it will be removed at a faster rate. The re= oval constant , As, for hydrogen iodide is about 1.5 ti=es the re= oval constant for elemental iodine.
I T
. ,/ Methyl Iodide Testing at CRNL(1),(2),(3),(h),(5) and IDo(6) indicate sub-stantial re=cval of methyl iodide by sodiu= thiosulfate.
Alkaline thicsulfate re= oves = ethyl iodide quickly (h),(5) and efficiently (5),(6). Wind tunnel para =eter studiesik) at ORNL recently deter =ined the = ass transfer constants necessary for the calculatica of = ethyl iodide re=cval by sprays of sodiu= thiosulfate colutions. Based on this data, ve evaluated the
=3thyl iodide re=cval constant, As, for our syste= and fcund it to be 0.67 hrs-1 With this value, the methyl iodide con-centration after two hours vould be reduced to =25 per cent of its original concentration.
The latest experi=ent in the Nuclear Safety Pilot Plant (NSPP; at Oak Ridge National Laboratory confir=s rapid removal of methyl iodide by an alkaline, sodiu= thiosulfate-boric acid solution. A =isting spray test (5) witn this solution in air at a=bient conditions effected a decentamination facter of k after only five =inutes of spraying. This corresponds to a = ethyl iodide re= oval half ti=0 of %2.5 =in. Si=ilar tests, with larger spray dreps and in a condensing stea=-air at=c-sphere simulating the accident enviren=ent, are scheduled to be run in Septe=ber and October 1967 OV
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- a. , .. , . ..sn <
0003 154
The thicaulfate spray syste: vill prevent any significant
= ethyl iodide ft.,rmation within the reacter building because
- a. The reacter building at=csphere vill be rapidly eccled through the te=perature range which is ther=cdyna=ically suitable to = ethyl iodide for=ation.
- b. The elemental iodine available for = ethyl iodide forma-tion vill be essentially eliminated by rapid re= oval through reaction with the che=ical spray.
- c. The for=ation of methyl iodide by reaction on catalytic surfaces vill be prevented due to the spray covering the surfaces with a liquid fil= of thicsulfate solution.
Thus, the = ethyl iodide in reactor building is essentially restricted to that released frc= the core. Recent literature (7),(0) suggests that a conservative estima-tien of = ethyl iodide release during a less-of-ccclant acci-dent is 1 per cent of the iodine inventery. It must also be realized that, although measure =ents have been =ade of the methyl iodide released directly frc= irradiatad fuel, = cst of the reported data indicates higher = ethyl iodide concen-trations (based on =ay-pack samples) then can be justified by chrc=atographic analyses. Also ncne of these experi=ents were performed in the presence of an intense radiation field, which causes decc= position of = ethyl iodide. Ilk) ) Particulates Stinchecebe and Goldsmith have shown that sub=leren particles can be rem stes=.(9),9ved (10) with 95-99 per cent efficiency by condensing They shev that the thermal and vapor pres-sure gradients which exist in a condensing steam environ =ent drive the sub=icron particles tevard the condensing surfaces where they becc=e trapped in the liquid phase. In addition, the particles serve as cendensation nuclei, which grew in si:e until gravitational and in rtial forces result in rapid deposition of the particulates. 9) High concentratiens of s=all particles are very unstable be-cause of the rapid aggic=eration into larger particles ,(ll),(12) which are then effectively re=cved by i=paction with the spray dreps, by vashout frc= the condensing steam, and by-settle =ent. As a result nearly all particulates are expected to be re-
=oved frc= contain=ent at=csphere.
Fur;ther=cre of Rochester'.lb),based en experi=ents at the Universit,r it see=s trat only 1 per cent of the iodine
=ay be associated with the aerosol particles in the reactor building.
Mi>A 5.u-u 0003 155
1 l I i i l l f The Spray Technology Pregra at CRITL includes plans to perfer: aerosol testa, one of which is scheduled to be run before the end of this year. These cests are expected to :entir: that earlier experi= ental work is applicable to the anticipated accident conditiens. The CPliL Spray Technology Progrs= has plannad an extensive program to de=cnstrate the effectiveness of a chemical spray syste= for the re= oval of all iodine species. Since the de-tails of the progra= are available frc= CPllL, the ec. plete pre gra.n vill not be presented here, but a brief su==ary of the three areas of the progrs= vnich pertain to this questien are:
- a. Single drop experiments in a vind tunnel vill determine the pars =eters and constants necessary for calculating the re=cval rates and efficiency for the varicus for=s Of 10-dine (including cethyl iodide). A significant portion of this verk has already been c:=pleted, and has contributed valuable data on the re=cval rate of meth"1 iodide by spray drops.
- b. There are lo experi=ents in the :Tuclear Safety Filet Plant scheduled for cc=pletion by the end of this year. These experi=ents vill provide the necessary data en the re= ova 2 of elemental iodine and methyl iodide by chemical spray syste=s using thiosulfate and other reagents under accider conditions.
,Jh i
- c. Chemical stra; syste= experiments using thicsulfate and other reagents are also scheduled to begin this Fall in the large facility at the Containment Systems Ixperi=ent (CSI). The results of these experiments will provide the final confir=ation of the data generated in the ISPP and in the vind tunnel experi=ents.
These experiments and the entire CFl*L Speay Tuchn01:gy Pre-grs= is directed tcvard the development of experi=entally verified engineering correlations which vill enable the ac-curate evaluati:n :f :he=1:a1 spray syste=s for the re=0vs1 of all species Of radi:10 dine. ~he OR'I*. Pr:grs= is scheduled to be ::=pleted early enough 30 tha: the data can be incorpc-rated into the Tnree Mile Island spray syste= design. U~ -
= .
o . ..._.=. m 'Ono 0003 156 9
5.13.h Effect of Cendensate Condensate on the cutside of the spray dr:p should have no significant effect on the performance of the syste= because:
- a. Under the postulated postaccident eccditions, the spray drops are expected to increase in dia=eter by about 6 per cent (regardless of their initisl size). In small
(%100 micron) spray drops the condensate fil= is 0.003
=m thick. If a thin condensate fi1= did form, an ex-tre=ely large concentraticn gradient would exist between the ions in the sodiu thiosulfate solution within the drop and the pure water ions in the thin film of conden-sate. Since the icns ia the aquecus thicsulfate solution are very =obile - as evidenced by the negligible liquid film = ass transfer resistance - and since the large gra-dient exists across only a 0.003 =m fil=, the ions should rapidly diffuse through the fils and equalize the gradient.
Theref:re, the existence of a condensate fils, as a separate entity, creates a streng driving force tcvard a unifors
=ixture, and the film is self-destructive.
- b. In large (=1,000 micron) spray drops, 6 per cent in-crease in dia=eter would result in a tils of condensate 0.03 =s thick. Although the potential film thickness is larger, the internal circulation within the large drops aids in the dissipation of the film as an entity. Thera-fore, in the large drops the fil= is destroyed by both diffusional forces and convective forces.
gg Calculations are planned to ec= pare the heat transfer rate and condensate film for=ation rate with the ionic diffusion rates and to esti= ate the effects of internal circulatien. However, the final prcof of the effect of condensation en the thiosulfate drops will be derived fres experimental data en tne rate of removal of iodine from a stea=-sir at=osphere using a sodiu thics21 fate spray. The CRNL 3 pray Techn010gy Progras plans to obtain this exper-i= ental data in three different experimental facilities. Single drop expert =ents in a vind tunnel vill deter =ine the effect of hu=idity, condensatien, temperature, drop size, and Other para =eters en the mass transfer 00 efficient of iodine and = ethyl iodide. A :ensiderable a=0unt of experi= ental work has already been perfor=cd i= this facility on the = ass transfer properties of = ethyl iodide, and additional experi-
=ents in coist at=osphere are planned f:r later this year.
""-~~ ax;eri=ents have been perfor:ed in the Nuclear Safe:y Fil:: Plan: :n the re=cval of 1: dine fr:= air a:= sphere us-ing a thi: sulfate s; rap. The experimental plan f:r the NS??
includes nine spray experi=ents en the re= val of 10dino fr:= stes:-air a =: spheres which appr:xi= ate the accident
- nditi:n. These ex;eri=en;s vill star in early Septe=ter 1^67 and sh:uld be ::=;1sted by the end Of the year. ::nfir-
=ati:n :f the results Ot:sined in these t-* * ' *- 'e 'as vill be btainsi by seva-*' ax; '-a- s in the ::::m*- en: Sys-gg j t a:s er;- * - vhi:h =-= '- *"' =' - ' -gin inring Fall 136 .
> uu;3 3/.i .
00u03 157
i 5.13 5 Scra? Syste= Rednndane?
'At all points in the reactor building spray system, the ac-tive cc=ponents are duplicated. This includes the valving at the outlet of the borated water and sediu= thicsulphate storage tanks, it includes the pu=ps, and it includes the main inlet valves L==ediately outside of the reacter build-ing. In each of these cases, the active ec=ponent is pre-
- vided in =ultiples of two so that a single failure in addi-4 tien to the reactor coolant syste= accident does not preclud complete operation of this system. To provide balanced flev between the two individual pu=ps and the two separate banks of spray no
- :les inside of the reactor building, two headers are provided with a normally closed manual valve between the two headers just downstream frc= the pu=p isolation valves.
In this arrangement, each bank of spray no::les is supplied by a separate pu=p. 5.13.6 Injection Rate The design of the thiosulfate injectica syste= precludes the possibility of spraying vater without thicsulfate into the reactor building. The configuration of the borated water an. thiosulfate solution storage tanks calls for these two tanks to be of essentially the same height. The fluid levels in these two storage tanks vill be maintained at essentially the sa=e elevation. Following engineered safeguards pu=p actua-tion, the outlet valves from these two tanks lead to a ec=- O mon header. With this piping arrangement and tank gec=etry, and since both tanks are vented to atmosphere, a proportional a= cunt o: borated water and thiosulfate solution vill be pu= ped into tha reactor building independent of the number of pumps oper-ating or flow paths available to the pumps. In essence, j these two tanks vill be oriented so as to for= a self-regu-lating configuration sbailar to a large =anc=eter wherein gravity will tend to as*ntain the two levels at essentially equal elevations until they are empty. With these design provisions, water sprayed into the reactor building at=csphe: vill kivays contain thicsulfate solutien. O ([ f
.]j)j; 5.13-7
l l rERE' ICES l Parker, et al., Reaction of Molecular Icdine and Methyl Iodide with Sodium l Thiosulfate Sprays, ORNL h071 (ARC-1966), pp 189-191. l Adams, R. E., Soldano, B. A., and Ward, W. T., Eehavier of Fissica Products i in Gas-Liquid Systems, ORNL h071 pp 179-183, Dec.1966. I Parker, G. W., Properties of Fission-Produce Aerosols Produced by Overheated Reactor Fuels.
)Soldano, 3. A. and Ward, W. T., Behavior '.ff Fission Products in Cas-Licuid
(
- Systems. Draft for Nuclear Safety Bimonthly report July - Aug. 1967 Parsley, L. F., Jr., Results frem NSPP Runs 20-22 (unpublished).
Ha==er, R. P., Evaluation of Scrub Solutions for Removal of Methyl Iodide i frem Gas Streams, Chemical and Process Development Branch Annual Report ( Fiscal Year 1966, IDO-lh680 ; 93-98.
)
'Keilholtt, G. W., Filters, Sorbents and Air Cleaning Systems as Engineered l
Safeguards in Nuclear Installations, ORNL-NSIC-13 p 139, Oct. 1966.
) Collins, D. A., et,al., Experience in Trapping Icdine-131 and Other Fission Products Released from Irradiated AGR-Type Fuel Elements, TID-7677, p 113 Stinchcembe, R. A. and Goldsmith, P.. Removal ef Iodine from Atmospheric by Condensing Steam, Journal of Nuclear Enerra Parts A/B, 2O, pp 261 to 275, 1966.
Stinchccmbe, R. A. and Goldsmith, P., Clean-up of Submicron Particles by l Condensing Steam, AERE-M-121k.
)Keilholtz, G. W. and 3arten D. J., Behavior of Iodine in Reactor Centain-l nent Systems, ORNL-NSIC h Feb. 1965 1
l Collins, D. d., ,et,al., t Experience in Trapping Icdine-131 and other Fission Products Released frem Irradiated AGR-Type Fuel Elements, TID-7677 p 113. l Craig, D. K., An Investigatien of the Interactions that Occur Between Radic-l nuclides and Aerosols in the Respirable Si:e Range, UR-636, University of Rochester, 196h. Mishima, J., Review of Methyl Icdide Behavier in Systems Centaining Airborne Radioicdine, 3N'aT-319. l l l l fe i %,.6 O /
,,1,_3 0003 159
Docket 50-289 Supplement No. 1
. October 2, 1967 QUESTION Consider adding flexibility to the emergency service water 5.14 source by providing a cross-tie between the secondary services and nuclear services pumps so that any of the pumps could serve as an emergency water source.
ANSWER A cross-tie with isolation valving vill be provided between the secondary services river water pumps and the nuclear services river water system,but the former vill not be con-sidered part of the engineered safeguards. O O 5.14-1 QQ{]} }<Q
() Docket 50-269 Supplement No. 1 October 2, 1967 CUESTION Discuss the consequences of a small leak and a large line break 5.15 within the containment in the closed loop nuclear service system after an accident. Would the source of a large line break be detectable in time to prevent major loss of inventory frem the system and resulting dilution of the borated recirculation water in the containment if the only indication is surge tank level? ANSWER The nuclear services cooling system is shown in Figure 9-10 of the PSAR. This system has a finite inventory, approximately 20,000 gals. of water. This system provides cooling vater to the emergenej coils of the Reagtor Building Ceoling System. The emergency coil of the reactor building air cooling system is under full system pressure during nor=al plant operation. Normal valve line-up for the emergency cooling coil vill be with the cooling vater supply header valves open and the cooling vater return header valves closed. This valving ar-rangement maintains this system in a standby condition. Flow is established in the coils by =erely opening the return valves of each of the thru coolers. The condition of the emergency coils are always known in the c ;?. above system. Monitoring alarms and level instrumentation s on the surge tank of the nuclear services closed cooling system vill provide for (1) leak detection in the entire system and not just within containment and (2) detection of small leaks. In addition flow orifices are provided in each of the reactor building cooling units. A differential pressure measurement between the orifices vill indicate a leak and initiate an alarm in the control room. This signal vill alert the control room operator and permit a check of the surge tank level and appro-priate correction action. The entire inventory of water in the nuclear services closed loop cooling system would not sufficiently dilute the borated water being circulated following an accident to per=it an approach to criticality. An analysis has shown that the inventory of approx-imately 20,000 gallens vould dilute the circulating borated water to a boron level of 2075 ppm. The preliminary design of the nuclear closed loop cooling system shows that its contents I are not sufficient to dilute the borated vater concentration below the levels discussed in section 3.2.2.1.3 of the PSAR for ! the BOL case with all control rod assemblies stuck-out . l i t .
- I 5.15-1 l
l
With the design provisions incorporated in the nuclear closed loop cooling syste=r, leaks in the system vill be readily detected and the integrity of the system insured. g l h) ic - cc. 0003 162 g
. 5. .5-2 l
Docket 50-289 . Supplement No. 1 October 2, 1967 QUESTION Provide calculations of the environmental effects resulting from an 6.1 accident which released TID-lh8hh fission product fractions to the containment but in which 5 per cent of the core iodine inventory is considered to be methyl iodide and that it is all released to the containment atmosphere (that is, 20 per cent of the iodine in the containment atmosphere is in a nonremovable form). Also, a volumet-ric rather than a virtual sourcs should be used for the release with a shape factor of one-half. Calculate the doses with and without the thiosulfate spray. ANSWER The postulation that 20 per cent of the iodine in the reactor build-ing atmosphere is in a nonremovable form or in methyl iodides is not , ' consistent with the infor=ation available. All data obtained to dat l on CH 3 I release from fuel was obtained in a very low radiation field i With a few exceptions, less than 1 per cent of iodine was found in ' the form of methyl iodide. Work at Brookhaven referenced in SNWL-319(1) cud other reports indicates that CH 3 I is unstable in high ra-diation fields. For example, %90 per cent methyl iodide deccmposi-tion was observed after absorption of %5'.5 x 107 rads.(1) In addi-tion several experimenters have demonstrat d thermally unstable at higher temperatures. 1)that methyl iodide Therefore, a more is rea. O listic, but still conservative, value for the r9 ease dide from the fuel is from 0.2 to 1.0 per cent.t2,3) 1 of methyl io-i i Recent litern ure suggests that <1 per cent of the iodine inventory is expected to exist in the form of CH3I assuming no credit for sys-tems, such as the thiosulfate spray system, which effectively de-stroys all conditions favorable to formation of CH3 I within the reac-ter building. (See answer to question 5.13.2 herein.) i Formation of CH 3 I in the reactor building is based on a catalytic reaction on surfaces. Experiments in steam atmospheres indicate that iodine behaves as if all surfaces were water. No catalytic re- j action is observed.(k) Furthermore, with the thiosulfate spray sys-tem for the Three Mile Island Nuclear Station the valls are expected i to be coated with the alkaline thiosulfate solution which will react l with the deposited iodine, thereby precluding the catalytic forma-tion of CH3 I iodide from this source. In addition, the entire build. ' 1 ing atnosphere vill have a negligible, iodine concentration due to rapid removal by the sprays. ! As sectioned in answer to Question 5.13.3 herein, the alkaline thio- t sulfate spray is effective for removing methyl iodide and aerosols. ; I For these reasons, the analysis presented in 14.2.2.h of the PSAR, ; ' which assumes that 5 per cent of the iodine in the' reactor building ' atmosphere is in the nonremovable form, is considered to be conser-vative. J O "' >u" l 0003 M3 1
\
6.1-1 t {
Dispersien factors for an initial 12-hcur accident period est;aated using a virtual source, local diffusien = ode'l are presented in Table 2-L of the PSAR, and are plotted as Curve 1 of the attached Figure 6.1-1. Cc= parable dispersica f actors esti=ated using a volu=etric source, local diffusion model are shown as Curves 2 and 3 en this figure. Curve 2 is based on the assu=ption that the projected vertical plane cross section of influential buildings is the mini =um of those build-ings indicated on Figure 1-5 of the PSAR 1.e., about h,000 m2 and that the shape factor is 1/2. But =cre than half of the vertical cross section is rectilinear as is indicated by examining Figures 1-5 through 1-10 of the PSAR. Therefore, one might assu=e a shape factor greater than 1/2. If such a factor vere 3/h, esti=ated dispersien veuld be as in Curve 3 of Figure 6.1-1. In either of these cases (Curves 2 and 3) esti=ated dispersion is greater at the site boundary than that esti-mated using the virtual source case (Curve 1). Further, these esti-mates do not take account of additional dispersion which can occur due to aerodynamic turbulence which would be generated by cooling tower structures. The preceding discussion has shewn that the at=cspheric dispersion at the exclusion distance using a virtual source model is more conserva-tive than the dispersion calculated using a volumetric model. Doses eticulated using the volu=e source model vill be less than those cal- ! culated in the PSAR using the virtual point source method. Assu=p-tiens regarding the per cent of nonre=ovable iodine and the operability of the spray system do not 6ffect the dose at the exclusion distance I since the penetration pressuri:ation and fluid block system limits the dose by terminating reactor building leakage. The analysis presented in 1k.2.2.4.2 of the PSAR did not take credit for any iodine reduction during the 1-minute time period required to terminate leakage, and therefore the calculated dose of 0 51 rem at the exclusion distance is independent of the assumed percentage of unremovable iodine or spray system operability-The percentage of unremovable iodine affects the dose at the exclu-sien distance only if failure of the leakage prevention systems is postulated. For this case, an assumed 20 per cent unremovable ic-dine species would produce a dose of 265 rem with both spray pumps operating and 285 rem with one pump operating. The questien further l postulates complete failure of the spray system. For this case, the dose at the exclusion distance vould be 1,230 rem. k*e do not believe that it is credible to assume that the spray system, the penetration pressurization system, and the isolatien fluid block system all fail at the sa=e ti=e that a MHA occurs. 0003 1 4 9 9 6.1-2
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J I
- REFERENCES Mishima, J., Review of Methyl Iodide Behavior in Systems Containing Airborne Radiciodine, BNVL-319, June 15, 1966.
Keilholt:, G. W., Filters, Sorbents and Air Cleaning Systems as Engineered Safeguards in Nuclear Installations, ORNL-NSIC-13, p 139, October 1966. 3)Co111ns, D. A., a g., Experience in Trapping Iodine-131 and Other Fission Products from Irradiated AGR-Type Fuel Elements, TID-7677, p 113. (k) Parsley, L. 3. , Jr. , ORNL, Private Comnunicat'ans. 1 I i 4 (O l . O 6-0003 145 4
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' Docket 50-289 Supple =ent No. 1 October 2, 1967 QUESTION Calculate the effect on the peak loss-of-ccolant accident containme.
6.2 pressure of (1) a steam generator blevdovn due to a massive failure during primary system blevdown, and (2) a steam generator blevdor.1 through a number of tubes ruptured during the primary system blevdo' ANSWER Massive Failure This analysis was performed for the 3.0 ft2 rupture of the reactor outlet piping, assuming core injection of 6,500 gpm, and operation of two core flooding tanks and three reactor building emergency coolers. The total mass and energy centained by one steam genera-ter plus feedvater coastdown is 47,500 lbs and 291 x 106 Btu, re-spectively. For the case of massive failure during reactor coolant (primary) syt tem blevdown it was assumed that the above additional mass and enerE was released to the reactor building during blevdown. This results in a peak pressure of 5h.3 psig (see Figure 6.2-1) which is below tt 55 psig design pressure of the reactor building. Tube Rupture () The second case was analyzed releasing the above additional mass anc energy linearly over a period of 2,000 sec following the rupture. This is equivalent to the failure of two steam generator tubes. The results of this calculation shows virtually no increase (0.1 psig) i the peak building pressure. At the time peak pressure occurs, i.e. , approximately h0 sec, very little additional energy has been release 4 O 6.2-1 (Revised 1-0-68 ) 00b f,})Y
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(~'% \~ / Docket 50-289 Supple =ent No. 1 October 2, 1967 QUESTION Discuss the effect of assuming heat transfer to the stea= generator 6.3 during blevdown on (1) the peak contain=ent pressure, and (2) the core thermal transient. ANSWFR Peak Contain=ent Pressure An analysis to determine the amount c. heat transferred in the stes= generator during the less-of-coolant accident has been made for the 3.0 ft2 hot leg rupture. This rupture produces the peak reactor building pressure. There are three areas of interest:
- a. Heat transfer to steam generator during early blevdern.
- b. Heat transfer to the reactor coolant frc= the steam generator during the latter part of the blevdown.
- c. Heat transferred after depressurization of the reactor coolant syste=.
Heat is assu=ed to be transferred to the steam generator until the I)
reactor coolant inlet temperature to the steam generator reaches 552 F, the saturation te=perature at 1,050 psig which is the stea=
generator safety valve set point. Heat is then transferred fro = the stea= generator through the tubes to the reactor coolant. Heat trans-fer coefficients were esti=ated based on 'the flow in the loop as predicted by FLASH. Heat transfer to the reactor coolant steam at=esphere after blevdown is over is assu=ed to take elace with a heat transfer coefficient of 2 Stu/hr-ft 2-F using the to'tal heat transfer surface of the tubes. For this rupture approxi=ately 9 =illien Stu are transferred to the stea= generator during Stage 1 (15 seck and approxi=ately the sa=e a=ount is transferred back over a period of 30 sec. Thus , an anergy l balance exists at the ti=e the peak reactor building pressure occurs. Centinued heat removal frc= the stea= generator between the end of blevdown and the second pressure peak, which is 2.9 psi less than the~first peak, is on the Order of 0.2 =illien Stu. The net energy addition t; the reactor building during the first 130 see is less , I tha: 1 =illion 3tu and has ne effect on the reacter building 7,43h pressure. C:re -'he.~11 Trans tent The effect of this energy transfer On : Ore 00 cling is regligible as the energy transfer is s-* ::= pared to the ::tal in the syste=.
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3 # e e 0003 1691
1 1 l l This transfer vould be seen as only slight changes in ecolant tem-peraturd and stea qualities. 3 hee these changes are first in ene direction and then reverse durin6 the second portion of the blevdevn, compensating effects vill occur. l l O i l l l l 9 p .. , o.3-2 0003 170 c ,,<,,
s Decket 50-289 Supple =ent No. 1 October 2, 1967 QUESTION Justify the assumption of a heat transfer coefficient of 20 Stu/hr 6.h ft2 0F in the upper half of core when the core is One-half filled with water af,te. a loss-of-coolant accident. ANSWER The water coming in through the emergency injection lines frcm the core flooding tanks very quickly starts to recover the c=re with be-rated water. The ccre at this time still retains a lot Of its sensi ble heat, and is generating heat due to the decay of fission product and other heavy isotopes. When recovery of the core starts, heat is transferred to the water; and after heating to the saturation point, steam is produced. By the time the core is half-covered, the steam production rate is et the order of 200 to 300 lb/sec. The convective heat transfer coefficient for the variable steam gen-eration rates existing during refilling has been evaluated using the following heat transfer correlation: h = 0.23 k/D(R,)0.8(p )1/3 Without considering radiation, the he developed fer a 200-lb/see flow ranges frem 23 to 30 Stu/hr-ft -F2 for steam temperatures rang- ' , ) ing frem 400 to 1,200 F. In the clad temperature range of 1,600 F, radiation from the clad to 800 F steam can contribute approximately another 3 Btu /hr-ft 2-F. However, even though the upper part of the core experiences steam cooling, core covering vill not terminate at the core milplane. There is enough water in the core ficoding tanks to fill the core past the three-quarter point taking no credit for the increase in volume due to steam formation. Thereafter, in less than 30 see the core vill be ecmpletely covered, and will remain so due to the actio: of the high and low pressure injection systems running at tve-thirds of their capacity. Om
"- 0003 171
1 Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Provide an analysis of the loss-ct-coolant accident for a spectrum 6.5 of break sizes and locations, including those breaks in the emergency core cooling systems nich also reduce the injection capability. In-clude curves of the ; ssure, water level, and fuel cladding temper-ature transients for och break size considered. Submit a chart showing the overlap and redundancy in the systems which cover vari-ous break sizes. In the discussien include the following: 6.5.1 A detailed description of how the steam bubble velocity as a function of pressure was deter =ined. Include a description of the physical process occurring in the reactor vessel during blevdown and discuss the limits of applicability of steady-state bubble rise data to blevdown analyses. 6.5.2 A comparison of pressure and water level transients for the variable-bubble-velocity model and the constant-bubble-velocity model for a spectrum of break sizes. In addition, compare these vith Loft semiscale blevdown data with and without inter-nals in place for various break sizes. 6.5.3 How the design of the high pressure injection system is affect-h. v ed by the assumed bubble velocity model. ANSWER An analysis of the loss-of-coolant accident has been =ade for a spec-trum of leak si:es and locations. This information has been analyzed and is reported according to the following grouping: (a) hot leg ruptures, (b) cold leg ruptures (c) injection line failures, and (d) injection system capability,
- s. Hot Leg Ruptures In Section 1h.2.2.3 of the PSAR an analysis of the 36-in. ID, double-ended pipe rupture was presented. This accident produced the fastest blevdown and lowest heat removal frc= the fuel, there-fore producing tPe highest cladding temperatures of any loss-of-coolant accident. This was therefore the basis for design of the core flooding equipment. A decrer.se in the rupture size as-sumed results in decreased =aximum clad temperature during a loss-of-coolant accident.
Evaluations have been performed for a spectrum of four additional rupture sizes using the same basic calculational technique and assumptions as for the large rupture case. These rupture sizes are 8.5 sq ft, 3.0 sq ft, 1.0 sq ft, and 0.h sq ft. The reactor coolant system mass release and pressure-time history for these rupture sizes are shcvn in Figures lk kl and ih h2 respectively of the PSAR. Di > ma - 6.5-1 0003'172
The reactor vessel water volu=e as a function of ti=e after the rupture for the various rupture sizes is shown in Figure 6.5-1. These water volume curves were generated utilizing the flow avail-able frc= core fleeding tanks, one high pressure injection pumps, and cne low pressure injection pu=ps. The pu= ping systems were assumed to have a ec=bined capacity of at least 3,500 g;= with the high pressure pu=ps running on emergency power within 25 see after the rupture, and the low pressure pumps delivering 3,000 gp= vhen the pressure has decayed to 100 psi, or at 25 sec, which-ever occurs later. Figure 6.5-2 shows the hot spot clad temperature as a function of time for the various rupture sizes. As can be seen frc= this figure, the small-sized ruptures yield =axi=u= clad tempera'.2 es which are considerably lover than those resulting frc= the larger sizes. The results of this study are shown in the folleving Table 6.5-1. Table 6.5-1 Tabulation of Loss-of-Coolant Accident Characteristics for Spectru= of Hot Leg Ruoture Sizes _ Rupture Min. Water Level Belev Hot Spot Size, Full-Power Botto= of Core, Max. Temp., ft2 , Seconds ft F ik.1 2.1 -6.8 1,950 8.5 3.h -5.2 1,916 3.0 1.5(,) -2.2 1,235 l.0 1 5l ,) +h.7 1,075 0.4 1.5(*) +12.0 1,015 (*) Blowdown forces on control reds are equal to, or less t Mn, nor=al pressure drop, and control rods will insert vith nor=al velocities. These values are for trip shutdown rather than for a void shutdown.
- b. Cold Leg Ruetures A si=ilar analysis of a spectru= of rupture sizes has been made for the cold leg piping. The rupture sizes tabulated are the double-ended. 28-in. ID, inlet pipe, which yields 8.5 sq ft of rupture area, 3.0 sq ft,1 sq ft, and 0.h sq ft of rupture area.
The reactor coolant system average pressure for this spectr's of rupture sizes as a function of time is shown in Figure 6.5-3 The water level as a functicn of time it shown on Figure 6.5 h. The water level calculation has been based upcn uninhibi:ed flood-ing as the check valves are provided in the core support barrel to equalize pressures and permit the trapped stea= above the core to escape cut the rupture. 6.5-2 (Revised 1-8-68) 0003 173
- r. t.
('a \ l 1) ! V
The hot spot temperature as a function of time for the spectrtv4 OV cf cold leg leak sizes is shown in Figure 6.5-5 The results of this analysis are shown in the folleving Table 6.5-2. Table 6.5-2 Tabulation of Loss-of-Coolant Accident Characteristics for Scectrum of Cold Leg Rupture Si:es Rupture Min. Water Level Belev Hot Spot Size, Full-Power ( ) Botten of Core, Max. Temp., ft2 Seconds ft F 8.5 0.4 *) -6.T 1.T85 3 1.0{*) -4.8 1,575 1.0 1.8(*) +3.6 1,250 0.4 1.3(*) +7.0 1,090 (*) Blowdown forces on control rods are equal to, or less than, normal pressure drop, and control rods will insert with nor- , mal velocity. These values are for trip shutdevn rather than l void shutdown. '
- c. Evaluation of Emergency Coolant Injection Line Failure f(x The evaluetion of a low pressure injection line failure has been made, and the results of the analysis show that the reactor is protected. The rupture of the pipe which connects the core flooding tanks and the low pressure injection flev to the reac-tor vessel was assumed to fail adjacent to reactor vessel and bercre the rirst check valve. (See Figure 6-1 of the PSAR).
This pipe has an internal diameter of 115 in. , and the resultant rupture area is 0.72 sq ft. Interpolation of available blevdevn calculatiens has beca used to evaluate this rupture size, and the data shows that a npture of this si:e vould result in the core being uncovered severs] feet belev the top of the core. Ecvever, the hot spot vill neeer be uncovered, and peak cladding temperatures vill be slightly less than that shown in Figure 6.5-5 for the 1.0-sq ft cold leg rupture. Since this small rupture si:e leaves a censiderable vater inven-tory in the reactor vessel, the remaining core flooding tank in-ventory is more than adeq2 ate to completely reflood the core. (DEITIED) l qL t 6.5-3 (Revised 1-S-68) 0003 174 ($i , # } f, '
The other lov pressure system can supply 3000 gym of water to the 6 reactor vessel and provide coolant to keep the core cooled. The ce=bined capacity of the two high pressure pu=ps is 1000 gpm which is in excess of the boiloff rate (680 gpm) due to decay heat is-mediately after blovdown. With a single 500 gpm high pressure in-jection pu=p the excess water above the core is adequate to pre-vent the core from being uncovered below the three quarter evalua-tion and beyond 300 see, the water level vill begin to increase. The high pressure injection system has two independent chains of flov to supply borated coolant to the system. If a rupture of high pressure injection piping vere to occur in one of the four lines between the attachment to the primary pipe and the check valve, the other chain of this system would have adequate capacity to protect the core against this s=all leak. In the event of a component failure in the second high pressure injection loop, the ruptured flow path can be monitored by the operator and spillage flow can be stopped by isolation of the affected piping. The entire capacity of one pump can then be utilized to handle the small rupture and protect the core.
- d. Evaluation of Emergency Core Injection System Perfor=ance ~
for Various Ruuture Sizes The loss-of-coolant analysis is based upon the operation of one 6 high pressure injection pump (500 gpm), one low pressure injec-tion pump (3.000 gym), and the operation of the core flooding tanks. The capability of other combinations of engineered safe-guards to provide core protection has been evaluated in a prelim-inary analysis. This capability is shown on Figure 6.5-6. In this evaluation the core is considered protected if the com-bination of emergency cooling systems considered vill prevent core damage which would interfere with further core cooling. The high pressure injection equipment with one pu=p operating can acec=modate leaks up to approximately 3 in. in diameter. The l6 preliminary analysis upon which this conclusion is based indicates that one pump will probably have the capability to protect j6 the core for leaks somewhat larger. A combination of one high pressure pump and one lov pressure in- l6 jection pump vill protect the core up to a 0.h-sq ft leak. This is eauivalent to the rupture of a pressurizer surge line. One 6 high pressure injection pump plus two lov pressure injection pumps can protect the core up to leak sizes of 3 0 sq ft. This i is considerably in excess of any of the piping connecting to the reactor coolant system. High pressure injection, plus the core flooding tanks and one low pressure injection pump, can protect the core up to Ih.1 sq ft, which is a double-ended rupture of the 36-in. ID, het leg piping. The core flooding tanks and one low pressure injection pump can protect the core frem about a 3-in. leak up to the lh.1 sq ft leak. Figure 6.5-6 de=enstrates that high pressure injection system provides core protection for nor=al operating leakage and for small leaks in which pressure decay of the system may be slov. For intemediate leak sizes, either the core flooding pgi i thk(s or lev pressure injection protects the core following the 6.5 h (P.evised 1-8-6a> 0 00 3 17 5
O("'s loss-of-coolant accident. For very large leaks in the category of a double-ended rupture of the reactor coolant piping, the core flooding tanks and low pressure injection together protect the core. For these leaks the core flooding tanks previde i==e-diate protection and can protect the core for several =inutes following the rupture. Due to their limited volu=e, they must be supplemented by the high flow frem the low pressure injecticn pumps within several minutes following the leak in order to pre-vent the core from again becoming uncovered as a result of boil-ing off the core flooding tank coolant. This evaluation of emergency core cooling capability demonstrates that the core is protected for the entire spectrum of leak sizes in both hot ind cold leg piping. 6.5.1 Steam Bubble velocity as a Function of Pressure In the process of trying to correlate the LOFT semiscale blev-down tests using the FLASH code, it became apparent that the method employed in FLASH to determine the amount of steam-vater separation did not adequately describe the phenomena. The variable that determines the amount of steam-vster sepa-ration in the FLASH program is the stea= bubble rise velocity. The original value of the bubble velocity incorporated into FLASH vas based on a report by '411 son et al.(1) This report j) shows experimentally deter =ined ter=inal velocities of stess bubbles rising through saturated water. Data vere obtained for the velocity as a function of void fracticn and pressure in two different diameter test vessels. Over the range of pressures considered, the bubble velocity varied between 1 and 7 ft/sec. The authors of FLASH chose a constant value of 2 ft/sec. For a given void fraction, a plot of the sa=e steam bubble velocity vercus pressure data as referenced above shows an increase in velocity with a decrease in pressure. A curve was fitted to the data to obtain the form of the equation. This expression was then inserted into the FLASH program in pir.e of the constant 2-ft/see value. Using the same form of the equation, the coefficients were changed until a bubble velocity was obtained that gave the best fit to the experi-mental data obtained from the LOFT semiscale blevdown tests without internal; in the vessel. During the depressurization of a high temperature water sys-te=, steam bubbles will be for:ed due to the flashing of the water. The rate of formation of-the stes= bubbles depends i on the heat addition and the rate of depressurization. If the steam bubbles are generated rapidly enough, they will rise, six and coalesce, and entrain water droplets. If the rate of for=ation is slow enough, the steam bubbles vill be unable to entrain water droplets, and ec=plete separation of f-~g the steam and water vill occur. U Until recently, caly steady-state data on flashing steam-
- r. " vater systems were available. The use of steady-state data
.d ' ".
0003 176 6.5-5
O is very useful in that certain phencmena can be examined with-out having perturbatic,ns frcm some other critical parameter. For exa=ple, in the case of the bubble velocity experi=ent, bubble velocities were obtained at different void fractions but at a constant pressure. By =aking series of tests at dif-ferent pressures, the pressure effect on the bubble velocity can be more clearly defined. By making use of this steady-state data together with the transient data currently being obtained by the Phillips Petro-leum Ccmpany in connection with the LOFT project, better en-pirical relationships for describing the physical processes can be obtained. As can be seen frem the answer to Question 6.5.2, the relationships now being employed are conservative. 6.5.2 Comparison - Bubble Velocity Models The variable bubble velocity model has been used to correlate the LOFT semiscale blevdown tests. The only data that have been released that show time-dependent pressure traces are that corresponding to a 100 per cent break area and a 6.1 per cent break area. These data are for the blevdown tests with-out the simulated internals installed. Information showing the per cent = ass remaining 3 min after depressurization has been reported for a number of differen; break sizes. Figures 6.5-7 and 6.5-8 chov the FLASH predictions of pressure versus time using the variable bubble velocity as compared / vith the measured data for the 100 per cent break area and the 6.1 per cent break area. Figure 6.5-7 shows that, for the 100 per cent break area, the FLASH code predicts depres-surization to be complete in less than 2 see as compared with 5 see frem the test data. This implies that the two-phase critical flow rates employed in FLASH may be on t3e high side. Figure 6.5-8 shows a good correlation of the test data for the 6.1 per cent break area over the entire blevdown period. Figure 6.5-9 shows the ecmparison of the measure d amount of residual water with the FLASH predictions using the variable bubble velocity me,lel. FLASH runs for break sites smaller than 6.1 per cent areas have not been made. The curve shows that the predicted residual mass is less than the measured residual mass for leaks larger than 7 per cent full opening. Also, the FLASH runs did not take into account the heat added to the water from the hot metal. This becomes significant especially for very small ruptures since the vessel has a longer time to du=p its heat into the water. If this heat in-put were considered in FLASH, it would result in a smaller percentage of water remaining than those shown on Figure 6.5-9 If a constant bubble velocity of 2 ft/see were used instead of the variable, the FLASH program would predict that 2 per cent of the mass re=ains for the 6.1 per cen ; break area. lh As seen frem Figure 6.5-9, 22 per cent of the initial mass
,v. ; c, . ., o 0003 177 6.5-6 1
( re=ains 3 =in after depressuri:ation. Therefore, even though the variable bubble velocity underpredicts the a=ount sf water re=aining, it is better than the constant value of 2 ft/see used in the criginal version of FLASH. Regardless of which bubble velocity =cdel is used, the amount of water re=aining for large ruptures with no injection is very small. For smaller ruptures some differences do exist depending on the leak location. A typical exa=ple is shown in Figures 6.5-10 and 6.5-11. These figuras show the Three Mile Island reactor vessel water volume and pressure for the assu=ed rupture of the pressurizer surge line. Only two of the three pumps in the high and lov' pressure injection syste=s were used. No credit was taken for the core flooding tanks. Figure 6.5-10 shows that the mini-
=um water level in the core is 2 ft higher for thi case where the variable bubble valecity =odel was used. This is due pri-
=arily t1 a slightly higher enthalpy and a corresponding lover density at the leak for the case of the varying bubble velocity.
It is also due to the slightly lover pressure that exists at a given time for the variable velocity case towards the end of the blevdown. This causes the lov pressure injection syste= to start injecting water at an earlier ti=e and provides for a faster recovering of the core. In the reactor the core flooding tanks vould start injecting water at 135 to lho see in both cases. The =ini=um vater level that is reached is less than a foot below the top of the core using the constant VBUB :odel and is never below the top of the core in the variable VBUB :odel. In either case, adequate core cooling is always provided. 6.5.3 Design - High Pressure Indeetten Eauie=ent The design of the high pressure injection equip =ent is not at all affected by the bubble velocity =cdel that is used. Tnis syste= is designed to acec==odate breaks in the system up to 3 in. in diameter with two of the three high pressure injec-tion pu=ps operating without uncovering the core. This is true regardless of the locatien of the leak, i.e. , in the hot or cold leg. The bubble velocity =odel that is used in this analysis is relatively uni =portant since the rate of depres-suri:ation is such that separation vill occur with a bubble velocity of 2 ft/sec. Also, in the pressure range of interest for the design of the high pressure injection equip =ent, the variable bubble velocity only varies frc= 1.5 ft/see at 2,200 osia to 3.3 ft/see at 600 psia. ~e.e variable velocity = ode.
's i therefore co= parable with the constant velocity =odel for these leak si:es and over this pressure range.
O U - 0 t, i >
' . [ofn. .-
6.,-7 0003 178
O
?EFERENCE (1)'411 son, J. F. , n 31, "The Velocity of Rising Stes= in a Bubblin6 Two-Phase Mixture," Transactions of the ANS, Vol. 5, No. 1, p 151, June 1962.
O} 1 O , 1 w, - 0003 179 ! s .tht s l 6.5-6 l
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'Elevat6on of Nozzles d 2.400 , - -- - - - - - - - - - -. - - - 1- - - - - - - - - - - -
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O l9 40 60 30 100 120 140 i60 Time, sec l M0T LEG RUPTURES - AUCTOR VESSEL WATER VOLUME VE TIME INCLUDING EFFECTS OF SOILOFF 20 INJECTION. O FIGURE 6.5-1
, AMEND. 6 (13 63) 0003 180
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l 0 - T, ,,* ; s M0T SPOT CLA00inG TDMRATURE VU TIME FOR SPECTRtm 0F COLS LEG RUI FIGutt 6.5-5 I
O L EGEND: HPl - hlGH PRES 3URE INJECTION LPI - Low PRES $URE INJECTION 2 MPl PUMP 2 HPt PUMP + 1 LPI PUNP 2 MPI PUMP 2 LPI PUNPS 2 MPl PUMP , 2 CORE FLOODING TAMES 2 MPI PUMP . 2 CORE FLOODING TANK 3 + 1 LPl PUMP i 2 CORE FLOOOING TamKS . I LPI PUMP Pressurizer Surge 3 in. 6 in. Line 36 in I ! l I I I I l 0.01 .02 .05 0,1 0.2 0.5 1.0 2. 0 5.0 10.0 Break Size. f t l A <G* i .,.s.l 0003 !85 O '
- HE.tGENCY CORE COOLING $YSTEN3 CAP A8ILITY FIGURE 6. 5-4
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(IO-2-47) l l I 0003 !86 1 l vE33EL PRES 3URE vet 3U3 TM - E-lu. ID TOP AuPTURL . LOFT
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301s-3EL PRE 33URE VER!U$ TlWE . g,h PER CDT BREAK AREA O T SBel3CALE E0M4 TEST 4. 544. 0003 on, Ui FIGURE 4.5 4 l
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2 l i i ! i t t t i \f f It 1 2 5 10 20 50 100 Pipe Break Size, f. Opening l l 1 l l 0 <; ; RESIDUAL WATER IN AN UNOBSTRUCTED REACTOR VESSEL AF O toe s'o oo a (52o e iao 2 aao rsia) FIGURE 6.5-9 0003 188
O l I I i l 2,400 Varia.ble VBUS /
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- g Bottom of Core 800 --- - -__- 1 ' ' _o __ _ 0 400 0
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-4 100 150 200 250 300 350 400 Time, sec CTOR VESSEL {)ATER VOLUME VERSUS TIME AFTER TURE ( 0. 4 FT WITH OPERATION OF ONE HIGH PRESSURE Ann 0'003A SURGE 189 LINE I Low PRESSURE INJECTION PUMP 3.
FIGURE 6.5-10
. , , i , AMEND. 6 (1-6 68) l
O . 24 20 LEGEN0: V8U6 2 f t/sec 16 V8U8 . Variable 7-o D M h 12 I a d o 8 \ ' 2 4
- N IO s
0 , ~ ~ . l 0 50 100 150 200 250 300 350 40 Time. sec l l l i l REACTOR 2 COOLANT AVERAGE PRES 3URE VER3U3 TIME AFTER A SURGE LINE RU (0.4 FT ) WlDi OPERAT!0N OF ONE HIGN PRES $uRE AND rW LOW PRES 3URE INJECTION PUNP3. FIGURE 6.5-11 AMEND. 6 (l*$*68) l 0003 190
1 I I I l l () l Docket 50-289 l Supplement No. 1 October 2, 1967 QUESTION Discuss the influence of the void geccetry assu=ed on the calculatien 6.6 of the reactivity inserted due to a positive moderator tempcrature coefficient during a loss-of-coolant cecident. Would a reiaced den-sity in the center of the core have = ore effect on peak temperatures than a uniform decrease in density? ANSWER The analysis of the core kinetics during the LOCA has been based on a detailed breakdown of the average channel using coefficients gen-ersted by assuming uniform void distribution. The calculation of re-activity effects in a ther=al reactor using the assumption of uni-formity is a good one if the reactivity has scme reasonable distribu-tion through the core. The =axi=us density variation at core exit at ultimate power is only 25 per cent; therefore, the distribution is relativ31y even. The distribution at mid-core is even closee. There is no analytical model available for calculating the void frac-tion as a function of radius and azimuth during a LOCA, nor is there any way of accounting for such a distribution in a kinetics analysis. Since there is such an even distribution of the voids, however, the reactivity feedback is also expected to be uniform. Therefore, the reactivity held in any one fuel assembly (or group) is very lov and jp) \\_- nearly proportional to its volume fraction in the core. It is con-cluded, then, that the average channel si=ulation used by the CHIC-KIN ccmputer code is an accurate one. The program considers six axial se6nents in the fuel, clad, and water; and six segnents radi-ally in the fuel. The code also weights the reactivity feedback ac-cording to the importance of each region. Since the principal void-ing effects are occurring in the axial dimension, the code does a proper calculation of the feedback. In the unrealistic case where it fs assumed that voiding of many central fuel asse=blies occurs, the local flux in this region vill tend to increase and decrease relative to the rest of the core. This would be stnilar in result to a rod ejection accident of very small rod vorth. Since the density variation in the radial direction is les. *han 5 f-* cent, the effect on total reactivity addition and, therefore, on peak amperatures, would be small. O ooa3 !91 6.6-1
3 Docket 50-269 Supplement No. 1 October 2, 1967 QUESTION Discuss the damage and calculate the deses which would result from 6.7 dropping a fuel bundle onto the core during refueling. ANS'4ER ' The concentration of boric acid in the water during refueling opera-tiens is sufficient to insure suberiticality regardless of the final resting position of the elecent. Calculattens of the stresses pro-duced in an element in the core, if it is struck end-on by the drop-ping ele =ent, indicate that possibly all of the fuel rods in the l' struck element vould lose their integrity tad release their gap ac-
*ivity. +
4 The calculated environ = ental release frc= the fuel failure is the same regardless of whether the accident occurs in the reactor build-ing or the fuel handling building since both buildings exhaust to the at=csphere through filters and since no credit is 'taken for de-cay during holdup. A total of 208 fuel rods are involved. There-fore, the d0se rates given in ik.2.2.1 of the PSAR vould increase by a factor of 208 + 58 or 3 59 The resultant doses at the exclu-sien distance vould be 2.62 rem to the thyroid and 2.65 rem to the whole body. O )
,-~ "
(ms/ 0003 19 c.- I e
-, e
O
\s / Docket 50-289 Supple =ent No. 1 October 2, 1967 QUESTION Justify the assumption that the a=0unt of diversion of injection wat 6.8 to the ruptured line during blevdown accident is not significant in the si:ing of the accu =ulator volume. Include a description of the physical phenc=ena in the annulus including the mass and velocity of water and stea= for various break sizes during the period which the accumulators inject water.
ANSWER Three possible mechanisms have been postulated by which the diversic: of emergency injection water could bypass the core and affect the size of,the core ficoding tanks. These are: (a) a stea= bubble in the core could prevent the coolant frc= entering the bottom of the core, (b) blevdcun flev could have sufficient velocity to carry the injected water along with the blevdevn steam, and (c) the high te=- perature reactor ecolant could flash the inec=ing injection water. Hot Lee Ruotures The hot leg ruptures provide a path for the coolant escaping frc= the reactor vessel up through the core and out the rupture. The injec-tien flow is devn the thermal shield annulus to the botto= of the re-actor vessel and then up through the core. Since the flev direction for all het leg ruptures is up through the core and out the rupture,
/(\ /~'s
. none of the above =echanis=s can cause a bypass or diversien of the emergency injection water.
Cold Lee Ruetures However, the large, cold leg ruptures have a leakage path which is down through the core and up the. thermal shield to the rupture. The injection point is between the botte= of the core and the rupture; therefore, the above postulated mechanisms have been evaluated for these rypture locations.
- a. Stea= Bubble in the Core The ste un bubble could divert injection flew if the stea= pressu:
in the core vere adequate to depress the water level in the core and prevent injection frc= filling the core. The check valves ir the upper portion of the core support barrel provide for pressure equali:ation, which eliminates depressicn of the cere water leve' and vents steam directly to the rupture. (These are discussed in detail in the response to Question 5.1 of this Supplement 1.)
- b. Carrvever The second pcstulated bypass mechanism is that of the escaping coolant having sufficient velocity to carry over the injection
'T vater to the rupture. The point of contact of these tvc fluids
~-) is in the thermal shield annulus which has a flew area of appror
=ately.33 f:2 For a pisten effect to eccur, a ficating slug of b i' '
- 4 5i.
0n03 193
cold, dense (56 lb/ft3), injection water en a flowing strea= of O stes= (6.7 lb/ft3, average mixed density of escaping ecolant at time injection starts ) =ust exist. For the slug to sustain its position or =ove outvard, the velocity must remain high enough to support the fluid piston, then pass around the cold water, and escape through the rupture. At the time injection . starts for the double-ended inlet pipe rup-ture, the ficv rate in the core and ther=al shield annulus is less than h,000 lb/sec. Assu=ing this is all saturated stea= at 600 psi, the velocity of the escaping fluid is approximately 8h ft/sec which could produce a stagnation pressure of 1.1 psi to hold up the fluid piston. This pressure is adequate to support a colu=nated slug of water approximately 2.8-ft high. Two seconds later, at 10 seconds after the rupture, the flow is less than 1,000 lbs/seg, and the average density of escaping coolant is 0.65 lbs/ft . The resultant velocity in the annulus is now reduced to 56 ft/see which can produce a stagnation pres-sure of only about 6 in. of water. With this reduction in stag-nation pressure a suspended fluid slug in the annulus must be collapsing and moving downward under the influence of gravity. At 13 sec, er 5 see after injection has started, the annulus flew is less than 500 lb/sec, and the density of the escaping coolant has decreased to 0.35 lb/ft3. The annulus velocity is caly 37.h ft/see and can produce a stagnation pressure of only 1.5 in. of water. The water is again unrestrained against free fall. / Thus, it can be seen that the stagnation pressures developed by the stea= flowing in the injection annulus cannot support a sig-nificant slug of injection water. For a short time the stagna-tien pressures could offer a small resistance to free fall, but this pressure rapidly decreases to the point that the injected fluid can fall freely under gravity.
- c. Flashing s
The third postulated =echanis= is that of the eschping fluid flashing the inec=ing injection water. This has been evaluated by assuming a flashing model in which the injection flow rate and the escaping fluid becc=e perfectly =ixed. Using a 280 F as energy datu= (point at which the condensing fluid would not flash l at the end of blevdown), the escaping coolant could flash the in-l cc=ing liquid for approximately the first 1.5 see of injection. Eeyend this time the injection flow can ec=pletely condense the escaping steam flow in the annulus. The injection up to this l time is approximately 150 ft3 and represents only 8 per cent of the core ficoding tank inventory. This loss of inventory in the , reactor vessel vould be more than ec=pensated by the gain frc= l cc=plete condensation during the re=sining 5.5 see of the blev-down.
*he reactor vessel annulus represents s large flew area available 6 S-2 ,1 <, . , s o 0003 194
z O) (,, for the injection water, and only 10 to 25 per cent of this flov area is required. Thus , no =echanism for intimate nixing is jus-tified. Since the flowing reactor coolant will only offe a s=a. resie:acce to injection coolant flow, and since the reactc. cool. ant cannot flash a significant amount of the injection coolant, there is no way for a significant quantity of the injection cool. ant to be diverted to the leak. A similar analysis has not been nade for the complete spectrum of leak si:es as the double-ended pipe rupture requires the grea-est amount of injection in the shortest period of time. In addi. tion, the smaller break si:es vill leave behind a greater per-centage of the original reactor coolant, and therefore less injec tion coolant is required. k O 6 8-3 0003 !95
Docket 50-289 x Supple =ent No. 1 , October 2, 1967
~
QUESTION Design Loadings and Factors 7.1 7.1.1 Provide scaled lead plots as a function of containment height for =oment, shear, deflection, longitudinal force, and hoop tension resulting individually frc= prestress, dead, pressure, design basis earthquake, wind, liner thermal (normal and accident), and concrete thermal (nor=al and accident) loadings. 7.1.2 Provide stress levels in the dome, shell, cylinder-base junction and in the ring girder region in the containment structure which result frc= the chosen loading co=oina-tions. 7.1 3 Provide a load cc=bination for ternado leading. 7.1.4 Provide the te=perature gradients calculated to exist anross the concrete containment shell under operational and design basis accident conditions. 7.1.5 Describe in more detail the wind load parameters including drag coefficient, gust factor, velocity to pressure con-version, and pressure distribution assumed. Cp ANSWER 7.1.1 Scaled plots of individual loads are shown in Figures 7.1-1 through 7.1-16. The load plots for operating te=perature and prestress are based upon preli=inary te=perature gradients and are being further developed. The accident te=perature load. plot will be ec=pleted upon finalization of temperature gradients. 7.1.2 The stress levels as specified above are tabulated in Tables 7.1-1 through 7.1-4 The stress levels at the base of wall are deter =ined based upon a straight transition of the liner. A detailed analysis of the lines at the base to cylinder transitica . is more fully described in the answer to the question 1 7.8.7 l l The stresses associated with accident and operating te=pera- l tures are based on preliminary te=perature gradients and l will be redesigned based upon final gradients. l l The stresses at initial operation (assumed at =axi=um winter l temperature gradient) are the maximum ce=pressive stresses I induced in the concrete. These stresses are te=porary in i that creep due to Operating te=perature has not occurrad. l (12) l e = '+ 0
.r 7.1 1 0003 196
The asterisk indicates stresses that occur at discontinuities and exceed 0.45f'c but less than the 0.60f'e specified by ACI 318, chapter 26,for te=porary leads. All concrete section: that have temporary stresses that exceed 0.45f *c in ec= pres-sion will be designed witn reinforcement to limit ecmpressive strains. 7.1 3 The following capacity formula (refer to Section 13 of Appendix 5B, " Design Stress Criteria") will be used to determine a load ccabination for tornado loading: C = 0 95 D + 1. cwt + 1.0 Pt Where
*?t = Wind loads based on a 300 =ph tornado.
Pt=oressure lead based on an internal pressure of 3 psi difference between inside and cutside of thi Reactor Buildin6 7.1.4 Fi dure 7.1-17 shows the te=perature profiles through the containment shell from preliminary analysis. The profile shown for normal operation was obtained with boundary conditions of 110 F inside the containment and a -5'F ambient condition outside containment. Sir'? srly the accident profile also had -5 F as a boundary :'ndition. The accident profile represented is the most ai verse temperature transient obtained with the liner te=perature assumed to reach the building maximum internal accident temperature of 281 F. - 7.1 5 The wind distribution for the Reactor Building will be based on infor=ation obtained from Wind Forces on Structures" (ASCE Paper - No. 3269) as described in PSAR Section 5.1.2 3.2. The basic wind velocity (wind 30' above ground) will be based on the fastest mile of wind with a 100 year period of recurrence. The Three Mile Island site will be classified as a coastal area and the wind velocity as a function of height will be expressed as: x V2 =V30 : 30 _ when Vz = velocity at height z grade V30 " f****** "il' # "i d ** 30' 'D V' 8 ""d
- = height above grade x = exponent for Ekmon spiral ranges frem 0 3 to 0.143 7.1-2 "
'9[
For the Three Mile It'*nd site the above will have the O value of: , V30 = 80 m.p.h. z = 144 ft. (the pressure at this elevation will be considered for the entire height of the cylinder.) x = o.263 Vz = 118 =ph The wind force will be applied in the vertical plane as shown in Figure 7.1-17. The wind pressure will be defined by the followin6 for=ula: g=iPVz 2 for standard air at o.07651 lb. per cu. ft. and V2in miles per hr: g - 0.002558 V22 Psr = 0.0000177 V:2 p,i i or 1 g = 0.2k6 psi The wind pressure distribution in the horizontal plane will be as shown in Figure 7.1-18 and will be defined by the Fourier series. f(x) = .1682 + o.0738 cox x + o.228 ces 2x + 0.lc ces 3x -0.0123 ces 4x + 0.0033 cos 5x + 0.016h cos 6x - o.002 cos 7x - 0.0124 cos 6
-o.003 ces 9x A dust factor of 1 3 will be used for those structures designed for tornado loadings.
I i O OP ~.1-3 0003 198
w
,-e P
O
- , TABLE 7,1-1
~ .
1011rTIFICATION: Conditions at Time of Isttial Os.eration _locattom and Meridional Stress - pet Boop Stress - pst Imading Ima t.ta outside Isalde outside laside outside Inside outside ConJittons St eel A- 4 t e sets e mcrete Steel Steel Ce crete concrete h o.95 Dead - 835.11 - 838.64 - 107.83 - 174.02 - 113.18 - 114.08 - 3.25 - 16.49 3 oper. Temp. b6T6.40 L624.30 - 326.44 379.29 - 6892.50 - 6827.80 - 6%4.71 382.w t Prestress -10757.00 -10708.0u - 832.31 - 149.15 - 4T86.70 - 4771.90 - 296.43 - 159.79 Total -16286.51 -16170.38 - 1266.58 56.12 -11792.38 -11713.18 - 9Ah.39 206.71 h 3 0.)5 Dead - 10Zl.20 - 102T.10 - 132.68 - 131.87 - 102.35 - 102.33 t '. 91 1.07 oper. T p. - 8011.60 - T828.90 - 879.31 1020.80 - %.80 - 7794.60 - 874.M 108T.80 g Prestress -11354.00 -11353.00 - 7T0.52 - 760.68 -17955.00 -17955.00 - 1365.5u - 1364.60 htal -20396.80 -20209.00 - 1782.51 128.25 -260ko.15 -25851.93 - 2238.94* - 345.75
-4 h h 0.95 Dead - 107.56 - Tua.18 -
93.8% - 17.73 153.98 155.60 30.07 65.29 L 3 oper. Temp. -10586.00 -10349.00 - !!99.30 3386.20 - 8650.60 - 84%T.So - 924.64 1104.50 t Prestress -20k88.00 -20292.00 - 1552.20 182.39 -11988.00 -11929.00 - 19T.23 - 450.31 h total -31781.56 -31343.18 - 2845.34' 1550.86 -19484.62 -20221.20 - 1691.80 699,k8 0.95 h ad - 627 00 - 62J.83 - 82.18 - 5.54 80.90 81.39 19.25 17.12
% oper. Temp. - 9397.00 - 9135.60 - 1083.00 1348.90 - 7456.20 - 7228.T0 - 80T.58 1853.40 6
Prestress -22887.00 -22825.00 - 1719.80 - 12bo.ko -16k39.co -16436.00 - 1143.00 - 1168.10
.$ Total -32911.00 -32581.43 - 2884.988 104.96 -23614.30 -23583.31 - 1931.33 2.42 Q
p f 0.95 Dead o.er. s Temp.
- 243.86
- 7001.70
- 248.k2
- 6798.30 29.66
- 781.18 82.74 948.96
- 196.40
- 6T61 70
- 199.38
- 6558.00 22.58
- 752.48
- 5 3. St.
986.96
- }
Prea6tess -19288.00 -19340.00 - 1h04.20 - 1786.40 -20005 00 -20035.00 - 146T.00 - 1658.10
'A [
Total -26533.56 -26386.72 - 2221.04 - 920.10 -26963.10 -26T92.38 - 2242.06 - 725.00
.O *14ss tiene 0.6 ('c O
O O O
O v Tatt2 7.1-2 7 TotafflFICAT!op: Test condition - C = 1.oD + 1. TSP
. Incation and leerldtonal Stress - pet hop Stress - pet fondlag Inalde outside Isalde outside Inside outalde Inside outside Conditions Steel Steel Concrete Concreta Steel Steel Concrete Concrete 3 Prestress -1o757.00 -10708.00 - 832.31 - 149.15 - 4786.70 4771.90 - 296.b3 - 159.79
- P = 63.3 pel 10546.00 10456.00 1118.t.o - 692.94 3389.70 3362. % 292.78 - 106.48
+g Dead toad - 854.31 - 857.45 - 110.32 - 180.31 - 115.90 - 116.85 -
3.35 - 17.34 Total - 1065.31 - 1109.45 375.77 - 1022.40 - 1512.90 - 1526.25 - 6.00 - 285.6 Prestress -11354.00 -11353.00 - 770.52 - 760.68 -17955.00 -17955.00 - 1365.50 - 1364.60 P = 63.3 psi St ok." 5104.50 542.56 536.62 8918.30 6918.00 1093.40 1092.20
'g Dead 84 4 - 104 . - 1062.80 - 137.29 - 136.46 - 105.90 - 105.88 0.94 1.10 f Total - 1312.00 - 7311.30 - 365.19 - 360.52 - 9th2.80 - 9142.88 - 271.16 - 271.30
.d 7 Prestress -2M88.co -20292.00 - 1552.20 182.39 -11988.00 -11929.00 - 797.23 - 450.31 vi h" P = 63.3 pet 7943.1o 7876.10 988.86 40.23 2857.40 2837.30 261.04 71.31 o
Dead toad - 144.30 - 7 38.6% - 98.71 18.57 161.82 163.52 31.61 47.63 g Total -13289.20 -13154.54 - 662.o5 2 M.05 - 8968.78 - 8928.18 - 504.58 - 314.37 Prestrese -22887.00 -22825.00 - 1719.80 - 1240.40 -16h39.00 -16436.00 - 11h3.00 - 1168.10 P = 63.3 pet 14001.00 13835.00 1791.80 - 2h2.h9 1070.90 104a.20 - 56.22 - 207.52
- Dead I- d - 659.97 - 653.b8 -
86.50 - 5.80 85.1% 85.65 20.26 18.04 Total - 9545.97 - 9643.48 - 14.50 - 1488.69 -15282.96 -15309.15 - 1178.96 - 1357.58 Prestress -19288.00 -19340.00 - 1 04.20 - 1786.Lo -20005.00 -20015.00 - 1467.00 - 1658.10 P = 63.3 pat 5089.80 5177.40 602.36 1629.70 5244.30 5297.20 619.6T 1165.80 bead load - 256.58 - 261.37 - 31.21 - 87.07 - 206.61 - 209.7% - 23,75 - 56.68 O Total -14454.78 -IteJ3.97 - 833 05 - 2h3.TT -14967.1% -14947.54 - 871.08 - 548.98 O C3 u h3 O O
a -. 4 TABL.E 7.1-3
$ 10ENTIFICAT105: AcetJent Conditions - C = 0.9SD e 1.5p + 1.o?
Lucation sad Merldtonal Stress - rel Maop Stress - s>et ImaJing Inal e outside Inside outside Inside outstJe Inside outstee Cunditions Steel Stee' Ccocrete Concrete Steel Steel Concrete Concrete o 0.95 Dead - 835.11 - 838.08 - 107.83 - 174.02 - 113 18 - 116.06 - 3.25 - 16.69
'J opr. Temp. 4676.40 - 4624.30 - 326.bh 379.29 - 6892.50 - 6827.8o - eA4.Ta 382.99
- Prestreme -1o757.00 -107 2.00 - 832.31 - 169.15 - b186.To - 47T1.90 - 296.43 - 159.79 t P = 82.5 pal 13745.00 13627.00 1T18.20 - 903.12 %417.80 4382.50 382.89 - 141.39
- g Acci. Temp. 0.0 0.0 0.0 0.0 0.o 0.0 0.0 0.0 5 Tot 1 - 2523.51 - 2443.38 451.62 - 847.00 - 7374.58 - 7331.28 - 563.50 65.32 0.95 bend - 1027.20 - 2027.10 - 132.68 - 131.87 - 102.35 - 102.33 c.91 1.07
'J oper. Temp. - 8017.60 - 1828.90 - 679.31 1020.80 - 7982.80 - 1794.60 - 874.35 1017.80
- Prestrema
-11354.00 -1 353.00 - 170.52 - 160.68 -17955.00 -1T955.00 - 1365.50 - 1364.60 0 P = 82.5 pst 6653.40 6652.80 107.13 699.39 11623.00 11623 00 1h25 00 1423.50
'.' Act1. Temp. -22254.60 -22443.80 - 675.41 187.86 -21582.85 -21771.07 - 6T4.45 2u8.26 Total -36000.00 -36000.00 - 1750.79 1015.50 -36000.00 -36000.00 - Ik88.39 4286.01
--J b o 0.95 Dead - 707.56 - 702.18 - 93.84 - 17.73 153 98 155.6o 30.07 45.29 f' 'J oper. Temp. -10586.00 -10349.00 - 1199.30 13f6.20 - 8650.64 - 8k%1.80 - 92k.64
- Prestress 1104.50
-20k88.00 - 2029.00 - 1552.20 182.39 -11988.00 -11929.00 - 79T.23 - 450.31 D P = 82.5 p t 10352.00 10265.00 1288.80 52.bb 3724.10 3697.90 3ko.21 92.94 g Acci. Temp. -1457t.44 -14921.62 - 453.12 139.75 -19239.48 -19476.To - 631,25 150.82 e.
Total -36000.00 -36000.00 - 2009.66 17h 3.05 -36000.00 -36000.00 - 1982.84 943.24
. 0.95 Dead - 627.00 - 620.83 - 82.18 -
5.54 80.90 81.39 19.25 17.12 9 oper. Temp. - 9397.00 - 9135.6o - 1083.00 1348.90 - 1456.20 - 7228.70 - 807.58 1153.60 0 Frestress -22887.00 -22825.00 - 1719.80 - 1240.k0 -16439.00 -16k36.00 - 1143.00 - 1168.10 t P = 82.5 pst 18247.00 18032.00 2335.30 - 316.ob 1395.70 1357 00 - 73.27 - 2To.hi
. Aces. Temp. -21336.00 -21k50.51 - 648.56 245.91 13581.4o C
Q Total -36000.00 -36000.00 - 1198.24 32.83 -36000.00 -36000.00 - 2h39.28 - 13t.15 C LN
. 0.95 Dead - 243.86 - 248.42 -
29.66 - 82.Th - 196.ko - 199.38 - 22.58 - 53.86 j op r. Temp. - 7001.70 - 6T98.30 - 787.18 948.96 - 6761.70 - 6558.00 - 752.48 986.96 Prestress -19288.00 -19340.00 - 1404.20 - 1786.4o -20005.00 -20035 00 - 1467.00 - 1658.10 Q % P = 82.5 pat 6633.60 6747 70 785.07 2124.00 6835.00 6903.90 807.63 1549.40 Acci. Temp. -16100.04 -16360.98 - 479.90 181.10 -158T1.90 -16:11.52 - 482.46 180.30 F* -36u00.00 Total -36000.00 - 1915.87 1385.52 -36000.00 -36000.00 - 1916.89 974.70 e G G
w f s' J G TAh!E 7.1 b ILINTIFICATION: Accident Conditions - C = 0.9"D + 1.OP + 1.OT + 1 OE Imestion and Mertalonel S_trea s - s,el Noop Stress - i.et Imdir.g laside outside Inside Outside != side Outside Inside outside Cu.d tt e us Steel Steel Concrete Concrete Steel Steel Concrete Concrete
,. 0.95 Dead - 835.11 - 838.08 - 107.8) - 174.02 - 113.18 - 11b.08 - 3.25 - 16.49 Q
oper. Temp. - 4616.bo 4624.30 - 326.bb 379.29 - 6892.50 - 6827.80 - 644.11 382.99 Prestress -10757.00 -10708.00 - 832.31 - 1b9.15 - 4786.70 4T71.90 - 296.43 - 159.79 t P = % pst 9163.40 9084.90 11b5.50 - 602.08 29b5.20 2921.60 2 % .26 - 94.26
. Acci. Teenp. o o o o o o o o j karthquate - 1242.15 - 1212.22 - 164.68 - 551.49 - 482.91 - b91 79 -
TT.b2 224.10 Total - 8347.26 - 8297.70 - 2b5.76 - 3097.45 - 9330.09 - 9283.97 - 76er.% 336. M
, 0.95 Dead - 1027.20 - 1027.10 - 132.68 - 131.87 - 162.35 - 102.33 0.91 1.07 d
Oper. Temp. - 8017.60 - 1828.90 - 879.31 1020.80 - 7982.80 - 7794.60 - 87b.35 101T.8o Prestress -11354.00 -11353.00 - 710.52 - 760.68 -179 % .00 -179 % .00 - 1365.50 - 1364.00 Q P = n p.1 bb35.60 bb35.20 471.b2 4M.26 7748.80 7748.70 950.03 9b9.00 3 Acci. Temp. -18514.91 -18l03.65 - 561.91 156.29 -17394.52 -17582.21 - 581.41 179.53 Earthquate - 1521.89 - 1522.M - 197.04 206.27 - 31b.13 - 314.56 - hv.4T 54.86 T t=1 -36000.00 -36000.00 - 2070.05 957.07 -36000.00 -36u00.00 - 1919 T9 o si.M M 0.95 Dead - 707.56 - 702.18 - 93.84 - 1T.73 153.98 1 % .60 30.07 45.29 7 -1
;j
- I oper. Temp.
Prestrees
-10586.00
-20488.00
-10349.00
-20292.00
- 1199.30
- 1552.20 1386.20 182.39
- 8650.00
-11988.00
- 8b4T.80
-11929.00
- 924.66
- 797.23 -
1104.50 450.31
.. P = n poi 6901.60 M43.bo 859.20 3k.96 2k82.80 2b65.30 226.81 61.96 Acci. Temp. -10238.20 -10627.87 - 318.39 98.19 -17532.43 -17776.25 - 575.25 13l.bb t
6 Earthquet. - 881.84 - 872.35 - 117.47 16.40 - b65.75 - 467.83 - 68.61 89.67 Total -36000.00 -36000.00 - 2422.00 1700.41 -360uo.co -36000.00 - 2108.85 988. % 0.9) Dead - 627.00 - 620.8) - 82.18 - 5.54 80.90' 81.39 19.2% 17.32 a oper. Temp. - 9397.00 - 9135.60 - 1083.00 1348.90 - Tb56.20 - 7228.70 - 807.58 !!53.bo j Prestress -22887.00 -22825.00 - 1719.80 - 1240.40 -16439.00 -16b36.00 - 1143.00 - 1168.10 P = % pal 12165.00 12C21.00 1%6.90 - 210.69 930.47 904.69 - 48.85 - 180.31 I Acci. Temp. -Ib410.51 -14609.8b - 438.05 166.09 -12488.b2 -12690.26 - 399.70 125.88 g g Earthquake - 843.49 - 829.7) - 117.08 51.14 - 627 75 - 631.12 - 93.87 - 117.71 C Total -36000.00 -36000.00 - 1883.21 109.50 -36000.00 -360u0.00 - 2bT3.75 - 169.T2 C fN 0.95 Dead - 2b3.86 - 248.b2 - 29.66 - 82.74 - 196.40 - 199.38 - 22.58 - 5 3.St. N l 4 Oper. Temp. Prestresa
- 7001.70
-19288.00
- 6798.30
-193bo.00
- 787.18
- Ab04.20 9k8.96
- 1786.40
- 6764.70
-20005.00
- 6558.00
-20035.00
- 752.b8
- Ab67.co 986.96
- 1658.30 g P a % ps t 4422.40 %498.50 523.38 1436.00 k%6.To b602.60 538.b2 1012.90 N u Accl. Temp. -1368T.69 -13911.11 - 407.99 354.48 -13278.05 -13497.b8 - bo).62 150.84
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$1DE SURF ACE OF REACTOR SUILDING 000-3 N0 FIGURES 7.118&7.119
. Docket 50-289 Supplement No. 1 October 2, 1967 QUEST!ON Seismic Analysis 72 7.2.1 Provide the analytical model (including consideration of mass distribution, stiffness coefficients, and vibrational modes) and the analytical procedures used in arriving at a leading distribution en the structure. 7.2.2 clarify the " algebraic" addition of vertical and horizontal seismic components. ANS'4ER 7.2.1 Refer to the answer to question 7.4 for the analytical model and the analytical procedures used in arriviO6 at a loading
~
distribution en the structure. 7.2.2 The vertical and hori: ental seismic components at any point on the shell will be added by taking the sum of the absolute value of the contributing vertical frequencies and adding it to the sum of the absolute value of the contributin6 horizontal frequencies. O l () 7.2-1 0003 221
f') N/ Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Large Openings 7,3 731 Provide criteria with regard to what si:e opening constitutes a large opening; hence, meriting special design consideratier 732 List the number and indicate the sizes of :he large openings for the containment. 733 Indicate the primary, secondary, and thermal loads that will be considered for design of these openings, including design load combinations. 7 3.4 Provide more detail on the analytical methods that are being used in the design of large openings. ANSWER 731 The Equipment Hatch and Personnel Lock will be analysed by using the finite element solution as described in the PSAR, Appendix 5C, and the answer to question 7 3.4 , The next largest opening will be the purge line sleeve which will have a diameter of approximately k8 in. The diameter to wall thickness ratio will be about 1.14 This opening
,("% and all other smaller openings will be designed as described 3s) in the answer to question 7.8.5 732 Large openings in the reactor building:
- 1 - Equipment Hatch, 22'-4" inside diameter 1 - Personnel Lock, 9'-6" inside diameter 733 The equipment access and personnel access openings will be designed for the loads and load ccmbination as specified in Sections 1.2 and 13 of Appendix 53," Design Program for the Reactor Building," of the PSAR.
7 3.h As indicated in Section 2 of Appendix 50 of the PSAR,the equipment access hatch and the personnel access lock will be analyzed by using the finite element technique leveloped by the Franklin Institute Research Laboratories. A dis-cussion of the abcVe =*thed of analysis is presented in the Fourth Supplement to the Preliminary Facility Description and Safety Analysis Report for the Robinson Station of the Carolina Power and light ecmpany, Docket No. 50-261. l'h G 7.a-1 0003 222
Oceket 50-289 Os Supplement No. 1 October 2, 1967
@ESTION Shell Analysis 7.h 7.k.1 The general shell analysis precedures are not provided in sufficient detail that a judgment =ay be =ade as to their adequacy. Provide a description of the analysis procedures for the structure to include a detailed descriptica of the analytical technique used, infor=ation verifying the accep-tability of the technique, sample calculations, the general geometry utilized to determine structural stresses and the consideration given to structural stiffness and discontinuity effects.
7.k.2 The analysis procedures for treating nonaxisym=etric loadings such cs vind lateral loading are not clear. Provide a detail explanation. 7.k.3 Describe the analysis procedures for the containment base slab design, particularly with respect to accaxisymmetric loadings. ANSWER neneral O 7.4.1 The shell of the Reactor Building vill be analyzed to dele mi: 7.k.2 all stresses, mcments, shears,and deflections due to the static and dynamic loads listed in Section 1.2, Appendix 53, of the original PSAR. Static Solution The static load stresses and deflections that are in a thin, elastic shell of revolution are calculated by an exact numerical solution of the general bending theory of shells. This analysis e= ploys the differential equaticas derived by E. Reissner and published in the "A=erican Journal of Mathematics," Vol 63, 19kl, pp. 177-18k. These equations are generally accepted as the standard ones for the analysis of thin shells of revolution. The equations given by E. Reissner are based on the linear theory of elasticity, and they take into account the bending as well as the =embrane action of the shell. The =ethod of solution is the multiseg=ent method of direct integratien, which is capable of calculating the exact solution of an arbitrary thin, elastic shell of revolution when subjected to any given edge, surface, and temperature leads. This =ethod of analysis was published in the n " Journal of Applied Mechanics," Vol 31, 196h, pp h67 h76 Q and has found vide applicatien by =any engineers concerned with the analysis of thin shells of revolution.
- h. ' ( i[',.
7 '-t 0003 223
The actual calculatica of the stresses produced in the shell and foundatica was carried cut by = cans of a ce=puter program written by Professer A. Kalnins of Lehigh University, g Bethlehe=, Pennsylvania. This ce=puter progra= =akes use of the exact equations given by E. Reissner, and solves the= by means of the multisegnent =ethod mentioned above. The program esa solve up to four layers in a shell and these layers can have different elastic and ther=al properties and can vary in thickness in the =eridional direction. Applied loads can vary in seridional and cirewnferential directicas. Dyna =ic Solution The stresses and displacements of the response of a shell of revolution to the excitation of an earthquake can be calculated by superi= posing the non=al =cdes of free-vibration of the shell. The =cdes of vibration are calculated by
=eans of the general tending theory of shells derived by E.
Reissner. The translatory inertia ter=s in the nor=al,
=eridional, and circu=terential direction of the shell are taken into account. The = ass distribution is the actual
= ass distribution of the shell and no approxi=ations are
=ade. E. Reissner's shell theory is such that it predicts exactly the ec=plete spectrum of natural frequencies of the shell without any approx 1=ations.
The differential equations given by E. Reissner are solved by means of the multiseg=ent direct integration method of ll solving eigenvalue proble=s, which was published by A. Kalnins in the " Journal of the Acoustical Society of A= erica," Vol 36, 1964, pp. 1355-1365 According to this method, the eigenvalue proble= of a shell of revolution is reduced to the
., solution of a frequency equation which vanishes at a natural frequency. The frequency equation consists of exact solutiens of E. Reissner's equations, and no approxi=ations are made.
The calculation of the natural frequencies and the correspond-ing = ode shapes of each = ode of free-vibration is perfor=ed by =eans of a cc=puter program vritten by A. Kalnius. The cc=puter progra: has been used for the calculation of the dynamic characteristics of =any types of shells of revolution and its results have been verified with experiments on many occasions (a listing of previous applications is attached). The progra: calculates the natural frequencies of any rotationally sy==etric thin shell within a given frequency interval and gives all the stresses, stress-corresponding to a natural frequency, consultants and displace =ents at any prescribed point on the =eridian of the shell. 1 i The nor=al = odes of free-vibration need only be added in order j to construct the response of the shell to an earthquake. The relaticuship between free-vibration and a given excitation is given by the following equation: l l 3 .i - 7.b-2 0003 224
/ ,', , . til ,
N /~' Ylx,t) = Yi(x) Ci Svi Wi Ni i=1 where Y (x,;) = fundamental variables of the respense Yi (x) = funda= ental variables of the 1"h = ode Ci = constant for the ith mode Wi = natural frequency of the i th mode Ni = h constant for the i mode Svi = maximum velocity from the response spectru for a single degree of freedom system for a given value of Wi for the ith mode For analysis purposes the Reactor Building shell is divided into structural parts, and each part is divided into a specified number of segments as shown in Fig. T.h-l. The Static Analysis and Dynamic Analysis have been used by i the following companies for the analysis of thin shells:
- 1. Martin Company - Orlando, Florida
- 2. Pratt and Whitney - Aircraft, East Hartford, Conn.
() 3. Central Electricity Generating Board - London, England The Static Analysis has been evaluated by H. Kraus, in Welding Research Council Bulletin, No. 108, September 1965 The Dynamic Analysis has described and its results compared to experiment by: J. J. Williams, " Natural Drought Cooling Towers - Ferry bridge and after," in the Institution of Civil Engineers publication, 12 June 1967. 7.h.3 The nonaxisy= metric loads i= posed upon the Reactor 3uilding base slab will not have a contributing influence upon the design of the shell; therefore, the foundation slab will be designed for nonaxisymmetric loads by considering a circular j slab on an elastic foundation. j i 1 1 o G 0003 225 7.k-3 i i
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Docket 50-289 Supplement No. 1 October 2, 1967 Q,UESTION Missile Protection 7.5 7.5.1 Describe the analytical procedures to be used in design of missile shields. ANSWER 7.5.1 The missiles within the Reactor Building are relatively small as compared with the mass of the surrounding rein-forced concrete structures. The high velocity and small impact area will cause the missiles to penetrate into the protective structures until its energies are dissipated. The analytical procedure in design of missile protection shielding will be divided into three groups:
- 1. The Impact Area . !.
- 2. The Local Effect 3 The Overall Structural Effect
- 1. The Impact Area The depth of penetration will be computed based on an empirical expression contained in the " Design Criteria
' Structural Theory for Protective Construction," Depart-ment of the Navy Bureau of Yards & Docks, April 1951, j as mentioned in Section 5.1.2.7 of the PSAR.
The protective reinforced concrete shield : mist be of a thickness not less than three times the depth of penetra-tien.
- 2. The Local Effect Ref. a: " Structural Design for Dynamic Loads," C.H. l Horris et al., McGraw Hill l Ref. b: " Plastic and Elastic Design of Slabs and Plates ," R. H. Wood, Ronald. l The collapse mechanism that will be investigated has the form of a circular fan, Ref. b, p. 29 Energy absorbed by the protective structure, according to the Yield Line Theory, will be based on a n ideal condition of plastic impact, Ref, a, p. 128. Mass ratio between the missile and part of the structure that will be considered effec-tive in absorbing the energy during the impact is assumed equal to 0.1. Based on the above condition, the kinetic energy to be absorbed by response of the collapse mechanism can be computed as W, = W2/2g(V12 )2, O
7 5-0003 227
Ref. a, p. 128. The work dissipated in the fan is Wi 4M, Ref. b, Eq. 4, p. 28. S.en,by equating W, = W4 , the required cir .=u= ulticate =ccent capacity of tee structure, acce .' ding to ACI 318-63, chapter 16, can be computed as M : W,/4.
- 3. The overall structaral Iffect The protective structure will be investigated for a con-centrated load applied at the point of missile impact.
In determining the collapse =echanism according to the Yield Line Theory, the kinetic energy to be absorbed by the structure, vill be as computed above. The internal strain energy dissipated by the collapse =echanism will again be equated such: We : W i, from which the ultimate moment capacity M can be solved. 0003 228 9, t i O 7 5-2 s
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j () Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Penetrations 7.6 7.6.1 Describe the analysis that shows that the piping penetration details as shown assure that pipe ruptures at the contain-ment shell will not result in loss of containment leakage integrity. What loading combinations are considered in the penetration design? 7.6.2 Provide the general criteria for and methods of reinforcing the concrete structure at and around penetrations. 7.6.3 The meaning of your "5.1.2.6.1 d criterion" is not under-stood. Provide further definition of this. criterion. ANSWER 7.6.1 All piping penetrations will be designed to ensure that the liner is not breeched due to the rupture of any process pipe. The load imposed on the contain=ent shell will be based upon the full plastic moment capability of the pipe with the moment calculated on the basis of the ultimate strength of the pipe. This load will be considered, provided there is a sufficient moment arm and thrust to potentially develop such a moment. In addition to the foregoing,the penetrations will be designed for those loadings detailed in Items a. through c. of Section 5.1.2.6.2 of the PSAR. Therefore the piping penetration and its anchorage into the' liner will be designed as a stronger element than the piping system. The method of analysis for thermal loading of the shell at the penetrations (openings) and all pipe reactions and moments is performed by methods suggested by A. K. Maghdi and A. C. Eringen in an article entitled " Stress Distribu-tion in a Cylindrical Shell with a Circular Cut Out," presented for publication in Ingenieus-Archiv during August 196h and in Stress Concentration Around Holes by G. N. Savin. Stress concentration factors used to analyze membrane stresses around the penetration are based upon these references. 7.6.2 Reinforcement will be provided as required for all radial, hoops and meridional =cment' .nd shears. Minimum rein-forcement will comply with 'c requirements of Chapter 26 of ACI 316-63 for shear re reement of webs. 7.6.3 sased upon the foregoing th9 .cading condition described in item d. of Section 5.l'.2.6. of the PSAR will be revised to read as follows:
" Pipe thrust leads to ensu 4 the vapor barrier is not breache n\~' due to the rupture of any piping system."
7.6-1 0003 229
i Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Shear 7.7 7.7.1 Reliance on ultimate values of shear (used as a measure of beam strength in diagonal tension) does not seem particularly applicable to shell structures. Provide amplification and justification. 7.7.2 In view of the indicated nonconservatism of the ultimate strength provisions of ACI 318-63 for combined loadings, your design criteria in this area must be more explicitly indicated. ANSWER 7.7.1 The formation of diagonal cracks in one-way slabs takes place in approximately the same manner as in beams. However, in two-way slabs, the stresses in the third dimension influence the ability of the material to resist the stresses in the other two dimensions. Thus, the behavior of a slab cannot be directly compared to the behavior of a beam. It has been verified through experimenta3 investigations of reinfor:ed concrete slabs that under certain conditions the slab will behave similarly to a beam in shear, Ref. /
" Shear and Diagonal Tension," Report of ACI-ASCE Committee 326, Fig. 8-1, Equation 8-14 As the ratio of column size to slab thickness, r/d, from the above report, approaches infinity the slao will have ecmparatively little slab action and will tend to behave like a wide, shallow beam.
Therefore, as "r/d" approaches infinity, the value of 4h would approach the corresponding shear strength of a beam. The circumferential stresses in the shell are normally uniform; that is, meridial shear is zero. Consequently, the radial displacement of the shell is uniform, with little slab action. Therefore, the ultimate shear strength for a beam will be used as a =easure of diagonal tension for the shell structure.
,. The, ultimate shear values used in the design will be in I ( ', accordance with Chapter 26, " Prestressed Concrete" of ACI
' 318-63',"except as noted belou in par: graph 7.7.2.
7.7.2 The load factors utili:ed in the criteria are based upon the lead factor concept employed in Part IV-3, " Structural Analysis and Proportioning of Membo s - Ultimate Strength Design" of ACI 318-63 The lead raetor of 0.95 applied to Dead Load represents the accuracy of deal load calcula-ticas (i.e. ,!5%) considering the greater severity of O' 0003 230 7 7-1
l l l 1 reduced dead loads for tension members. The load factor I applied to the pressure loads due to the Maximum Hypothetical Accident of 1 5 is consistent with that suggested by Waters h and Barrett (1,2) as the limit of low strain behavior on i prestressed concrete pressure vessels for nuclear reactors. l This factor is also consistent with the proposed set of
" French Regulations Concernin6 Concrete Reactor Pressure Vessels" wherein it is stated that "The design pressure shall not exceed 2/5 of the pressure calculated to bring about destruction of the structure by rupture of the cables." The load factor considering a stress of 0.60 fu at factored load, would thereby equal 0.6 + 2/5 or 1.5.
The load factor applied to the design earthquake load is consistent with that utilized in ACI 318. The reduction is the load factor applied to the pressure load when the design earthquake is experienced and is also consistent with the reduction in ACI 318. The shear stress limits and shear reinforcing for radial shear used in the design will be in accordance with Chapter 26, " Prestressed Concrete" of ACI 318-63, except ss follows: (1) In Equation (26-12) the shear increnent between flexural and diagonal tension cracking (0.6b'd fc') will be modified based upon the results of testing under the direction of Professor Alan Muttock of the University of Washington. The resulting equation will be Vei = K bd %+ m. Mer d
+ Vd v :
where K y = 1.75 - 0.036 + k.0 np np In accordance with ACI 318 the factor K 3y will not be considered to be greater than 0.6. (2) Requirements for minimum shear reinforcement as called for in Equation (26-11) of ACI 318 will be provided only at discontinuites. 0003 231 1 1
- w. :
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Docket 50-289 Supplement No. 1 Octcher 2, 1967 QUESTION Liner 7.8
- 7.8.1 Discuss the typss and ccabinations of leading censidered with regard to liner buckling. What are the stress levels in the liner and how do they relate to the buckling stress.
7.8.2 Describe the geometrical pattern, type and spacing of liner attachments and the analysis precedures, boundary conditions and results with respect to buckling under the leads cited in answer to "1" above. Describe the analytical procedures and technique to be used in liner anchorage design, including example calculations. 7.8.3 Describe the pressure / thermal load variatiods considered. T.B.h Describe the failure mode and failure propagation character-istics of anchorages. Discuss the extent to which these characteristics influence leak tightness integrity. What design provisions vill be incorporated to prevent anchorage failures frem jeopardizing leak-tight integrity. Describe the anchorage design considerations given to and tolerances [ on liner plate out-of-rcundness, liner plate ritup, liner plate thickness and liner yield strength variation and these bases. 7.8.5 Describe the procedures for analysis of liner stresses around openings. Also provide the =ethod of liner design to accom-mocate these stresses and the related stress limits. 7.8.6 Describe the design approach that will be used where loadings must be transferred thrcugh the liner such as at crane brackets or machinery equipment = cunts. Also, provide typical design details. 7.8.7 Provide the liner detail to be used at the base-cylinder liner juncture, the strain conditions imposed at the juncture, and an analysis of the capability of the chosen liner detail to absorb these strains under design basis accident conditicas. 7.8.8 Discuss the extent to which containment vacuum can influence liner buckling and the capability of the chosen liner / attachment arrangement to resist possible vacuus loading, f ANSWER 7.8.1 The Reactor Building liner vill be designed for the leads that are specified in Section 1.2 of Appendix 53, " Design Losds" and vill be ocmbined as specified in Section 1.3 of O . Appendix 53, " Design Stress criteria." 0003 232 7.8-1
The stress levels in the liner are as tabulated in the answer to Question 7.1.2. The liner vill be designed so that the critical buckling stress will be greater than the proportional li=it of the steel. Present analysis indicates . that the basic accident conditions produce a strain of approxinately 0.0022 in./in. in the liner. 7.8.2 The Reactor Building liner anchors will be vertical angles as shown in Figure 5-1 of the PSAR and will be spaced horizontally at 18 inches center to center. The liner vill be analyzed as a flat plate. This assu=ption is conservative in that the liner vill have to buckle against its own curvature. For analysis it is assumed that the liner is fixed at the angles and that there vill not be any differential radial =ovement of the boundaries. The folleving analysis is based on interaction curves given by A. Pflu*ger: "Stabilit'atspre,bleme der Elastostatik" Springer-Verlag, Berlin, 196k. The critical stress resultants N i g and N20 are the stresses
. induced in the plate (Figure 7.8-1) and are defined as:
N1
= K$Ne where K5
= 6.97 N2 K3Ne where K3 = k.00 and Ne = 772ggd 1 x
12(1- Y 2) b4 It is seen frc= the interaction curve (Pfluger) that for a = oo the influence frc= N1 can be neglected.
.'. N(critical) = kT.0 kai Ihe liner anchors are designed and spaced so that the critical buckling stress will be greater than the proportional li=it of the liner.
The liner anchors vill be designed to resist the leads induced when a section of the liner between anchors may exhibit greater stresses than the adjacent panel. These leads vill be as r.hown in Figure 7.3-2. 7.8.3 Figure 7.8-3 is a preli=inary analysis of the reacter building pressure and reacter building liner te=perature as a function of time after the MHA. This analysis is per-for=ed using the digital ec=puter ccde " CONTE 4PT" develeped by Phillips Petroleu= Cc=pany. Results of this cede are used in deter =ining what =axi=u= pressures and te=peratures result frc= the MHA and serve to descritte what variations of pressure and te=perature leads the liner is subjected to with ti=e. 7 S.k The liner anchors vill be designed such that the velds connecting the anchors to the liner vill fail before the l c ., . . 0003 233 a.. 7.3 2 t I 1 i
liner is breeched. Where the anchor angles do net confor=
~\ to the curvature of the plate, such that the specified veld cannot be =ade, the angle shall be reshaped to ecnfor= to the configuration of the plate. The design of the velds between the liner and anchers will be based c= the =ini=u=
acceptable thickness, and the -"-" guaranteed ulti= ate strength of the liner. 7.8.5 The procedure for the analysis of the liner and the concrete shell is provided in Paragraph 2 of Appendix 5C of the PSAR and the answer to Question 7.3. The stress concentration in the liner around the re=aining opening vill be deter =ined by the =ethods suggested in the references noted in answer to Question 7.6 and the paper entitled " Reinforce =ent of a S=all Circular Hole in a Plane Sheet Un:ier Tensien" by Levy, McPherson and Smith appearing in the Journal of Applied Mechanics, June 19h8, page 160. Reinforcing of the liner at openings will be in accordance with the ASME Soiler and Pressure Vessel Ccde, Secti:n VIII unless analysis indicates greater reinforcement is required. Stress li=its are in accordance with the ASME Nuclear Vessels Code. A = ore detailed description of the analytical =ethod for penetrations is contained in the Fourth Supple =ent to the Preli=inary
,' Facility; Description and Safety Analysis Report for 3rockwood nuclear Station, Unit No. 1.
7.8.6 For transfer of lead through the liner, see crane bracket I I~' details as described in the answer to Questien 3.2.h. i Equip =ent anchored in the base slabs vill on occasions be required to be bolted down through the base liner as shown in Figure 7.8 k. The controlling factors in doing so are based on the fact that the total net uplift and overturning forces are too large to utilize the slab above the liner to transfer the forces to the nearby valls. 7.8.7 The Reacter 3uilding cylinder to base junction .ill be as shown in Figuce 7.8-5 The ec=pressible =aterial vill be such that the ituckle plate can defer = and absorb the strains produ:et by Operating and design basic a::ident conditiens. The analysis of ihe knuckle plate vill te carried cut by the
=ethods des:ritet- in the answer to Questien 7.L. 3ased upcn the present anal / sis,the felleving leads, =cve=ents and strains vill be applici :: the knuckle:
- 1. Onternal pressures ::rresponding :: the design accident
- nditi:n.
- 2. A vertical strain Of -0.0036 in. and a lateral ==ve=ent of -0.00603 in appliei a: the : p Of the knu:kle due :
dead 1:ad and prestress 1:ai. ( 3. A verti:al strain :f -0.;;30 in and a lateral =:ve=ent of ,0.000~- ':,n. applici a: the ::p of the knu:k'.e ius :: the design a::iien ::nitti:ns.
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Based upon a si=ilar solution described in the fourth supple =ent to the Prel1=inary Facility Descriptica and Safety Analysis Report for 3rcokvced Unit No.1 for =cre h severe motion and pressure loading, it is concluded that
=axi=u= tensile stresses vill be less than yield stress.
7.8.8 Contain=ent vacuu= can only occur during operating condition and therefore could only influence liner buckling duri 6 that ti=e. The deflection associated with a vacuu= load of 2.5 psi and an axial stress of about 24 ksi is apprcx1=atel: 0.010 in. The height of the middle ordinate due to the - curvature of the plate is 0.05L in. Considering this deflection and tolerances in construction the vacuus vill not influence the buckling of the liner. The liner anchors and welds vill be designed to withstand a vacuu= icad of 2 5 psi and vill also be designed such that the veld er anchor vill fail before the liner is breeched. l 0003 235 l l [ O , 1 l ( I l l l l 9 7.S k
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O B ASE CYLlHDER JUN LINER DET All FIGUR E 7.3 5 i (
Supplement No. 1 Docket No. 50-289 October 2,1968 O D QUESTICN Anchorage Zone Design of Prestressing Tendons 7.9 7.9.1 Describe the analytical procedures used for anchorage :ene analysis and typical analytical results. 7.9 2 Provide typical details of anchorage zone reinforcing. 7.9.3 Provide test data supporting the acceptability for your reinforcing =ethod to resist the' i.:: paired (1:.. posed) anchorage loading (particularly under extende? icading). ANS'n'ER 7.9.1 The calysis of the anchorage ::enes for the prestressed tendens is based upon chapter 9 "Trans=1ssion of the Prestressing Forces to the Concrete" in Prestressed Cenerete Design and Censtruction by F. leonhardt (second edition), and upon ACI-318-63, chapter 26. Two factors have been censidered: (1) bearing stress , and (2) transverse tensile forces (" splitting forces") . A typical analysis for the hori: ental tendens is as follevs: (For tenden and buttress layout see Figure 7.9-1) D _esign Criteria 7pd f'c = 5000 psi fs = 240,000 psi A3 = 169-1/k" dres = 8.31 in.2 Max. prestress Force = 70% of ULT. strength of tendon
= .70 x 2ko x 8.31 = ik00 KIPS Analysis
- 1. Bearing stress en Buttresses !
Ab = 20. 52-172 = 381.5 in.2 h Bearing stress: ikOO = 3670 psi 381 5 Allowable stress frc= ACI-318 vertical spacing of tendens == 17" distance frc= t, bearing lt. to outside of buttress T 16" Ab ' = 32 -17 = 986.5 in.2 by ACI-318 f(allevable) = 0.6 fe 3} Ab ',/A , l q = 0.6 x 5000 x o.5/331,5 I V
= 3000 x 1.39 = kl60 > 3670 l
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However, it should be noted that f in the above for=ula is li=itid to the design strength of the concrete. If the actua. value df f were e used (approximately 6000 psi) l the allovable stress vould be: f(allevab10) = 0.6 x 6000 x 9865/331,5
= 3600 x 1.39 = 5000 " 3670 The bearing stress is considered to act uniformally over the area of the bearing plate. The use of a square shi=
with a square bearing plate and with the controlling di-l =ensions as shcvn in Figure 7.9 h provides sufficient dis-tribution of the tendon load to justiff the assu=ption of unifor= bearing stress. Nevertheless , a check was =ade on the extre=e condition if a circular effective bearing area for the bearing plate is assumed. The following i stress vould be produced. Ab = 1 ( 20 5 ) 2 _1 ( 7 )2 = 291. 5 in . k k l Bearing stress = lh00/291.5 = h.60 ksi Fro: ACI-318 Ab ' " 1 (32)2 _ 1 (7)2 = 763.5 4 h 1 JAf/Ab = g 763.5/2915 = 1.37 O / i The allowable bearing stress using the actual concrete strength is f(allovable) = .6 x 6000 x 1.37 = h930 psi. l Fro = the above it can be concluded that the bearing ( stresses produced by the square bearing plate does not
! exceed code allowables even when very conservative l assumptions are =ade.
l 2. Transverse Tensile Forces The transverse tensile forces produced in dis-tributing the tendon force fro = the bearing plate to the cross section of the =e=ber is a function of the tendon force, bearing plate si:e and the eccentricity of the force.1,2,3 The " equivalent pris="1,2 =ethod produces transverse splitting stresses as shown in Figure 7.9-5 for the buttresses . The total splitting force is 112 KIPS. An example of t i ,,i;., a spiral to resist this force vould be a i t; . - 3/h" d bar with a radius of approximately 10 in, ana a pitch of 5 in. Additional re-inforce=ent vill be provided in areas in which tensile stresses could produce spalling. 0003 242 7.9-2 (F.evised 6-21-68)
O 7.9.2 A typical anchorage feinfore;.:g detail is shown en Figures 7.9-2 and 7.9-3 7.9.2 The problem of analyzing the end block in a prestressed structure is three dimensional in nature. However, little work has been done to investigate the problem as such. Most tests and theoretical solutions are based upon simplifi 2-di=ensional systems. The theoretical soluticas and tests that have been =ade seem to verify the conclusions reached b F. Leonhardt regarding the nagnitude and location of the transverse tensile force. The =agnitude of the transverse tensile ferce varies between 0 and 30% of the prestressing force, depending upon the size of the bearing plate relative to the pretreseed cross section, and the spacing of the tendens. The transverse tensile force exists from a point close to the bearing plate to a point at a distance equal to the depth of the structure. The tensile force reaches a maximum value at a distance of about 3d from the bearing plate. Ref. 1) Leonhardt, Fritz - Prestressed Concrete Design and construction. Wilhelm Ernst & Schn 196h
- 2) Guyon Y - Prestressed Cccerete - John Wiley & Sons Inc.1960 0 3) Lin , T. Y. - Design of Prestressed Cencrete Structures - John Wiley and Sen Inc.1963 h) K. T. Sundara Raja Iyengar - Two-Dicensional Theories of Anchorage Zone S Stresses in Post-Tensioned Prestressed Seems. Journal of the Amer-ican Concrete Institute, October 1962
- 5) T1 Euang - Stresses in End Blocks of A Post-Tensioned Prestressed Beam. Journal of the American Concrete Institute, May 196h 0003 2 O
7 9-3 (Revised 6-28-63)
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i O 170 wlRE ANCHOR HEAD l 170 wlRE SHIM y--REARING PLATE 204" = 20M' m 3%'
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Docket 50-289 Supplement No.1 'O October 2, 1967 QUESTION Reinforcine Steel 7.10 7.10.1 Considering the critical nature of the structure, a materia specification on splicing in conformance with ACI 318-63 does not provide adequate assurance of structural ductility Revise your material performance criteria in this regard and provide more explicit information with regard to the type of cadweld splicing intended. 7.10.2 Indicate the extent to which splice stagger vill be achieve 7.10.3 Indicate the location of and extent to which splicing or tacking of reinforcing steel vill be made by velding. 7.10.k Discuss in detail the extent to which NDT requirements vill be imposed on the reinforcing steel. Also, indicate how quality control vill be exercised to insure that these re-quirements are achieved. (If no requirements are imposed justify the omission. ) Discuss similar requirements for the prestressing vire and anchorage hardware. ANSWER 7 10.1 Section 2.3 of Appendix SD, " Quality Control," provides a /('T specification for the random sampling and destructive () testing of the tension splices for reinforcing bars of size larger than #11 using the CADWELD system. This speci-fication is developed on the basis of a testing program which provides a 99 percent confidence level, that 95 perce of the splices develop 125 percent of the minimum guarantee yield strength of the bar. The actual specification for the CADWELD splices will be developed on the basis of using the splice designed to develop the ultimate strength of the bar, or with the use of deformed bars conforming to ASTM A h08-6h, Intermediate Grade, a minimum tensile stress of 70,000 psi. The specification for random sampling vill be amplified to require that the average (X) equals or exceeds the minimum ultimate strength. If the average of the process is above this value, the process is considered to be in control. If the average falls below this value, an engineering investigation will be required to determine the cause of the low breaks and to re-establish control. 7.10.2 Splices at points of maximum tensile stress vill be avoidec insofar as possible. Alternate splices for concrete rein-forcement vill be staggered a minimum of 6 feet 0 inches, when the center to center spacing of, bars is less than 12 inches. O 0003 249 I. ' . ' 7.10-1
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7.10.3 Provision has been made to build a retaining structure l around all below-grade portiens of the Reactor Building to provide access for field insts11ation and tensioning
. of tendons. This retaining structure vill also keep ground water away frem the contain=ent shell, thereby I
eliminating the necessity of grounding the rebars in the containment shell. Therefore, no tack velding of ! reinforcement vill be performed. Arc welding vill not , be used to splice concrete reinforcement as stated in l Section 2.3 of Appendix SD. Tension splices for bar sizes larger than #11 vill be made with a CADWELD splice. 7.10.4 All principal load car:fing components of ferritic materials for the centainment vessel exposed to the external en-vironment will be selected and tested to confirm that their nil ductility transition temperature is at least 300 F below the minimum service metal temperature. The ferritic mater-l ials exposed to the external environment consist of the penetrations and large openings (equipment access hatch and personnel locks) for which materials vill be selected to conform with the ASME Boiler and Pressure Veesel Code, ( Section III, for Class "B" Vessels. Material specifications for the penetrations are more completely described in Appendix 5F. The containment vessel vill be designed so that it is not susceptible to a lov temperature brittle fracture. This does not mean, however, that each element in the structural system which is a ferritic material vill comply with a NDT + 300F criterion. The transition temperature, defined g by an impact test, is not considered relevent for the design of the concrete containment shell for the following reasons: (1) History (a) To the best of our knowledge, no prestressed or l , reinforced concrete members have failed due to a lov temperature brittle fracture. (b) No suspension spans, all of which use high strength vire, have failed due to a lov temperature brittle i fracture. (c) To the best of our knowledge, no prestressed I concrete primary vessels, during proof test or l vhen tested to destruction, have failed due to a lov temperature brittle fracture. (d) Modern design concepts, such as ultimate strength design and energy absorption methods for aseismic design, acknowledge that the criterion of NDT, as defined by Charpy impact tests, is not applicable to uniarially stressed rebars and tendons. O w ..m T.10-2 003 250
(2) Uniaxial Stress F.ssentially only uniaxial stresses are applied to the mild steel reinforcing and tendon material. The . triaxial stresses which may induce brittle behavior at higher temperatures by restricting plastic flow are thus avoided. Field inspection vill ensure the absence of all mechanical and metallurgical notches from the material. (3) Strain Rate The tendons are stressed more highly during the jackin operation than they are during any design condition. Prestress losses exceed the minimal increase of tendon stress during the load application. Because the strai during pressure leading of the prestressed elements is primarily a function of the concrete strains and because the application of the accident pressure load is considerably slover than the application of an 1 impact load, the steel elements would not experience ' the high rate of strain associated with impact loading l (k) Residual Stresses i No splicing of mild steel reinforcement by are velding ! vill be permitted,thereby avoiding residual stresses produced by velding. The tendon material is stress
/ relieved, thereby beneficially tempering heat affected zones and favorable altering the material's micro-structure.
(5) Fatigue Tests Lehigh University has advised that fatigue tests being performed on 270 K vire at room temperature and at 0 F show no change in properties at the lover temperature. l l These facts coupled with experience to date in concret construction, indicate that transition temperatures, as defined by an impact test, is not indicative of the behavior of concrete structures at lov temperatures whether they be mild steel reinforced or prestressed. 0003 251 O i 7.10-3 l
Supplc=ent No. 1 Docket No. 50-289 October 2,1968 QUESTION Prestressing Materials 7.11 7.11.1 Sub=it a detailed descriptica of the prestressing =aterials and hardware selected. 7.11.2 Justify the prestressing syste= selection. Include data with regard to ulti= ate tendon strength, elengatica, anchorage strength , hardware dyna =ic perfor=ance, etc. ANSWER 7.11.1 The prestressing syste= to be used for the reactor building is the BBRV syste: utilizing a =axi=u= cf 169-1/h inch diameter wires. The wires vill consist of high tensile steel *, bright, cold drawn and stress relieved confer =ing to ASTM Ah21-59T Type BA, " Specifications for 'Jncoated Stress Relieved Wire for Prestressed Concrete." The BERV system uses parallel wires with cold for=ed buttenheads at the ends which bear upon a perf rated steel anchor head, thus providing a positive mechanical means for transferring the prestress force. The anchorage hardware is designed and fabricated for the use of 170 vires. Hevever, one hole vill be used to acco==odate an unstressed surveillance wire as described in the answer to Questien 8.7 included in Supplement No. 3 to the PSAR.
/ ' The materials to be used for anchorage conponents are as follows:
Ite= Material Washer h1h0 Heat Treated Washer Nun h1h0 Heat Treated Cc=posite Washer h1h0 Heat Treated Split Shi=s ASTM A36 Bearing Plate ASTM A36 The k1ho =aterial used for anchorage ec=penents has a mini =u: yield stress of 163 ksi, ulti= ate tensile stress of 180 ksi and a Rockwell "C" Hardness of h0 to kh. Details including di=ensions of anchorage co=ponents are as shown en Figures 7.11-1 through 7.11-6. 7 11.2 The wires for the prestressing tenden vill develop a mini == guaranteed ultimate strength of 2h0,000 psi and a =ini=u= elongation of h.0 percent when measured in a 10 in gage length. Quality control measures for the wire are detailed in Appendix 5D, Section 2.6, "Prestressing Tendons" of the PSAR. [*, *dd f"> b 0003 252 711-1 (Revised 6-26-66) L
The specifications for the prestressing tendons require that the anchorage system be a positive =echanical type with cc. penents capable of developing the =in1=u= guar-anteed ulti= ate strength of the tenden. The specifications further require that each type of anchorage harteare be tested using one or more asse=blies censisting of a length of tendon with an anchor at each end. The test asse=bly/ asse=blies are required to be statically leaded a=d measure-ments =ade to ensure that the require =ents for ultimate strength are met as set forth in tne tenden material speci-fication. The total elongatien of the asse=bly is to be not less than 2.5 percent =easured over not less than a 10 ft gage length. The anchorage harteare are to shev no significant per=ar.ent physical distertion when tested to 100 percent of the =ini=us guaranteed ulti= ate strength of the tenden. The anchorage cc=p:nents are designed for a factor of safety of 1.5 based upon a guaranteed ulti= ate tenden strength of 2002.3 kips for a 170 wire unit. A detailed report of the design and testing of the anchorage cc=ponents which verifies their adequacy is included as Attach =ent No. 1 to this question. Supple =enta2/ data on static tests perfor=ed on the anchorage ec=penents are reported in A=end=ent No. k to the "Preli=inary Safety Ar.slysis Report for Fort St. Vrains Nuclear Power Station. All test data indicate that the total elengation of the asse=bly is not less than 3 5 percent =easured ever a 30 g ft gage length. W j The BBRV systes provides a pcsitive anchcrage with excellent properties when subjected to cyclic loadings. The specifi-cations for the prest- ssing tendons require that dynamic tests be perfor=ed wherein the test asse=bly/asse=blies are to consist of a single ele =ent prototype tendon having no less than 10 percent of the full scale tenden capacity that vill duplicate insofar as possible the behavior of a full scale tenden including anchcrage. The test as se=bly/asse=blies are to withstand without failure 500,000 cycles of lead fres 60 to 66 percent of the =ini=us guaranteed ulti= ate strength of the tenden. This test is intended to represent an extre=e indication of cyclic loads anticipated during the service life of the structure. A test is scheduled to be performed using a prototype tenden censisting of 70-1/k inch diameter wires loaded for 500,000 cycles at the specified stress range at a frequency of approx 1=ately 360 cycles per =inute. The previous fatigue tests reported in the "Forth Supplement to the Preliminary Fac:.lity Description and Safety Analysis Report for the 3reckvcod Nuclear Station Unit No.1" provide assurance that these specified perfor:sace require =ents can readily be =et. A second test which is also specified in-0003 253
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7.11-2 (Revised 6-23-6c)
O volves asse=bly/asse=blies which are to withstand without failure a =ini=:= cf 50 cycles of Icading correspending to the folleving percentages of the =ini-" guaranteed ultimate tenden strength of: 60 ! 2000 L + 100 where "L" is the length in feet of the tendon in the structure. For the reacter building the shortest tenden which occurs in the dome is approximately 100 ft which results in a stress range as a percentage of 4-"- guaranteed ultimate tenden strength of 60 10 percent for a stress variation of ikk ksi ! 2k ksi. This second test is intended insofar as practical to de=enstrate the capability of the tenden to respond satisfactorily to dynamic loads such as those resulting from an earthquake. The number of cycles selected for this test is a rational estimate of the number of cycles of earthquake motion corresponding to the frequency of the structure. Be most severe tenden stress variation vill occur in the vertical tendons of the cylinder. The tendon stress variation for the test when applied to the vertical ten-dens results in a concrete stress variation of I 154 psi. f The theoretical concrete me=brane stresses resulting from
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the =aximum hyecthetical earthquake average for all nat-ural frequencies over the cylinder height are approxi-
=ately 120 psi. It can thus be seen that the test stress variations are therefore a conservative measure of the expected stress variations due to the seis=ic loadings.
A test is scheduled to be perfor=ed using a prototype tenden of the same size previously described which vill be loaded at approxi=ately the sa=e frequency. The type of =echanical anchorage provided by the 33RV systen as well as the previously perfor=ed fatigue tests provide assurance that these perfor=ance require =ents can readily be met. It is well kncun that all prestress tendons are subject to the = cat critical stress during initial tensioning. Because of the expected prestress 1csses, the stress in the tendens even when subjected to the hypothesized loads vill not be as high as during the initial Jacking cperatien. This means that a failure of a tendon under either design accident or test loading is quite remote. Although it is difficult to predict or even centemplate the possibility of hypothetical tenden failures , a study was perfor=ed to deter =ine the effect of the total icss of three adjacent tendons either vertically or circumferentially in the cylinder or in the < de=e. This study indicates that the icss of three adjacent tendons vill not jeopardize the capability of the structure to withstand the design accident loading condition. 0003 254 u b' .n 7.11-3 (Revised 6-20-od)
O An evaluation has been =ade of the potential for brittle fracture of those ferritic =aterials which are expeced to the external enviren=ent ec=prising the anchorage ec=penents. For all practical purposes the washer, vasher nut, and split shi=s are subject to only shear and cc=pressive stresses and consequently should not be subject to brittle fracture. The bearing plate as shown on Figure 7.9-h included with the answer to Question 7.9 very nearly approximates a pedestal or shear block. Nevertheless bending =ay occur. For this reason the =aterial used confor=s to AS24 A-36 "Specificatica for Structural Steel" including the optional requirement of this specification of silicon killed fine grain practice for steel used at te=peratures where i= proved notch toughness is i=portant. Tests are seneduled to be perfor=ed to detertine the nil ductility transition te=perature of the bearing plate
=aterial using the Charpy 7-notch i= pact test ( ASO4 A-370 Type A)
A study was made regarding the use of large curved tendons. The most severe condition due to the radial force i= posed by curved tendens occurs in the vicinity of the large opening where the radius , spacing center to center of tendens , and spacing between the edge of the hole and the first tendon all an the least di=ensions . The analysis of the large opening is described in Appendix SC and the answer to Question 7 3 , included in S pple=ent No. I to the PSAR. In this analysis the tendon radial forces are i. posed as concentrated leads at node points and concrete stresses are deter =ined in cen-junction with other leads for the pertinent leading ec= bin-ations. The results of this analysis insofar as it is presently ec. pleted do not indicate that excessive concrete tensile stresses vill occur. As a check on this analysis a s1=plified model was assumed as described in Figure 7.11-7 This anal-ysis indicates that critical concrete tensile stresses vill not exist due to the radial force. Another =cde of possible failure that was analyzed was that resulting frc= pure shear between tendons . 3e least di=ension center to center of tendens is 10 in. "'he =axi=u= shear stress l based upon a parabolic shear distributica for the least spacing l is 325 ;si. Although ACI-313 does not provide a definite shear stress li=it, it is sc=ewhat si=ilar to the shear per-
=itted between stirrups which for verking strength design is li=ited to 35h psi. Censequently this = ode of failure is also not critical.
In regard to the curved tendens in the ic=e the least dimensica censidered '-- "-a a= - -= " "-* d---- - st ba.1 ef tendens " = --- ds 12 inches. 37 inspection it :an be s een that this ---dd -' -- 's less severe ths= that described for the large opening. i - , ,, I , g i8 s J . 0003 255
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7 ro = ll' - 2 ' Y = VARIABLE LEAST RADIUS OF DRAPED t = 84~ P = RAOIAL TEWDON F'ORCE l l 0003 262 l O EFFECT OF TENDON CUR FIGURE 7.117 AMEND.11 (6 28 48) l
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" Attachment :To.1" to Questien T.11
" Report of Design and Testing of End Anchcra6es for the 170 Wire 3ERV System' 0003 263 ,
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!O TECHNICAL REPORT NUMBER 8 THE WCS 2.0 Mep/170 W POST TENSIONING SYSTEM INTERIM REPORT:
CHAPTER 3 END ANCHORAGE. JANUARY, 0003 264
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@ 1968 Western Concrete Structures Co.,Inc..
O TABLE OF CONTENTS for 0.MAPTER 3.0 END ANCHORAGE 3.1 GENERAL 1 3.1.1 COMPONENTS 1 3.1.2 PERFORMANCE CRITERI A 1 3.2 PROTOTYPE DESIGN 2 3.2.1 GENERAL 2 3.2.2 SPLIT SHIM . BEARING PLATE INTERFACE 9
'3.2.3 COMPOSITE WASHER SPLIT SHIM INTERFACE 10 3.2.4 9-3/8 INCH DI AMETER THREAD 11 3.2.5 6 INCH OI AMETER THREAD - WITHOUT SHIMS 12 3.2.6 6 INCH DIAMETER THREAD .WITH SHIMS 13 3.2.7 WIRE HOLE WEB SHEAR 13 3.2.8 FAILURE MODE ANALYSIS 15 3.3 PROTOTYPE TESTS 16 3.
3.1 DESCRIPTION
OF TEST PROGRAM 16 3.3.2 PROTOTYPE ANCHOR AGE HARDWARE 16 3.3.3 TEST SERIES PL PRELIMINARY.4' x 170 W TENDONS 16 3.3.4 TEST SERIES A 30' x 170 W TENDONS 19 3.3.5 TEST SERIES B AND C. GENERAL 23 3.3.6 TEST SERIES 81 WEB SHEAR WITH SHIMS 26 3.3.7 TEST SERIES B2. WEB SHEAR WITHOUT SHIMS 29 3.3.8 TEST SERIES C2 6 INCH THREAD WITHOUT SHIMS 30 3.3.9 TEST SERIES C1 6 INCH THREAD WITH SHIMS 32 3.3.10 ANALYSIS OF FAILURE MODE FROM TESTS 34 3.3.11
SUMMARY
CONCLUSIONS 34 0003 265 O : l
. 1
(_ 'l
CHAPTER 3.0 END ANCHORAGE 3.1 GENERAL l 1
.1 COMPONENTS i end anchorage ha dware of the WCS 2.0 Mep/170 W Since neither the mean ultimate load capacity ({} nc
- t. Tensioning System is made up of the comoonents listed standard deviation (a) are iraown until after prototype Table 3.1 1. The terminology "A end" is used to designate are completed, it is necessa. , to establish a preliminary cr end of the tendon wnicn has the Washer installed and for design purposes. Previous experience indicates that d es shop headed during tendon fabrication. The tendon tube ing for a Safety Factor of 1.5 will oroduce test results me he A end has an enlarged section of sufficient diameter and the basic criteria. This gives: Component Design Ult Jth to allow the Washer to be recessed appro'.imate!y 6 Strength (P') = 1.5 x 2002.8 = 300 t.2 kips.
. mside the face of the Bearing Plate, so that tne unheaded es can project approximately 6 feet beyond the Bearing Another independent consideration influences the desige e at the opposite end (B end) of the tendon. The "8 end" mate capacity of the end anchorage caraconents. Due t se end of the tendon (opposite the A end) which allows a critical structural application, a proof load Mst of all cr noosite Washer (or optionally a Wasner and Washer Nut) to components to minimum tendon ultimate (2902 8 kip-nstalled on the projecting wires, which are then field headed. been established as an essential part of cuality assuranct g o.;,g o, . cedures, it follows that all components must be belcw i l*f& "cx l mcx yield point at the proof test load in order that the proc l
[c ve o. 4 .o ,, r, be a non-destructive procedure. For 4140 steel heat treat
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(==.+ 77.r . ~~ %, #., *. c therefore follows that the design ultimate strengtn (P')- 3LE 3.1 1; Tendon end anchorage components of the WCS 2.0 Meo/ weakest failure mode for each component should be: Post-Tensioning Svstem. P170 2002.8
.2 PERFORMANCE CRITERIA P* > 0.90 x 0.90 0.81
( basic criteria for performance of the end anchorage of an sonded tendon system for a prestressed concrete reactor Taking all of the above into account, the preliminary cr
;el (PCRV) or other nuclear containment is that it must for prototype end anchorage component design ult-ably: 1) sustain the permanentlong term load on the tendon strength is 3004.2 kips for the most critical failure mode the life of the structure,2) sustain any variations in tendon the final criteria for performance being:
1 for the life of the structure, and 3) have sufficient over. J capacity to allow the full actual ultimate strength and ((- 3 ol x (F v+ F )u > 2002.8 kips mate r Angation of the tendon wires to be developed.
)ressed more simply, the end anchonge must be stronger n the tendon which it anchors, for all types of loading dition.
I actual pnysical and mechanical properties of the tendon e can be determined by statistical analysis of test data. As
- ussed in Section 3.3.4, a long (> 30 feet) tendon composed 170 individual ASTM A421 wires of 0.250 inch diameter be expected to produce an ultimate load > 2002.8 kips, I an ultimate elongation > 3.5%. Due to the mode of failure i multiple wire tendon, resulting from variation of the indi-Jai wires, the average or the maximum values of either ulti-te load or ultimate elongation will not greatly exceed the o<<
iimums. The ultimate load capacity of the end anchorage 1ponents can be determined by ultimate load tests con-0003 c00
- ted on prototype components so as to test all critical fail-modes. It can be assumed that the ultimate load capacity otoJuction anchorage components will fall within a range the r:1ean ultimate load capacity of the mcst critical failure
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de ((} plus o minus three standard deviations (a) of test ' g alts. Many specifications require that end anchorage compo-ts may not yield at the minimum guaranteed tendon ngth. Therefore thy basic performance criteria for the end norage of tne,2.0, Meril70 W System can be establisned i P = ([ 3 ol x (F, e Ff) > 2002.8 kibs. 1
3.2 PROTOTYPE DESIGN O
.1 GENERAt.
)rication drawings for prototype end anchorage components net effect of all these factors can be reduced to a simple shown in Fig's. 3.21 thru 3.2 5 as follows: concept called tne rupture factor (kr) which is the ratio failure load as determined by calculation (F u x A')
, t COMPONENT NAME FIGURE actual failure load as determined by ultimate load test - !
component (P"). Therefore kr = (Fu x A;) + P", where Washer 3.2 1 nominal arer. of steel. kr is normally greater than 1.C Washer Nut 3.2 2 value of kr is determined from previcus testing of similar Composite Washer 3.2-3 anism designed in accordancewith the same type of calcul Split Shims 3.2 4 Each series of tests allows determination of revised rt Bearing P! ate 3.2 5 factors, so that calculated failure loads become more ac-as more testing experience is gained. The rupture factors i load on the tendon at all maior loading conditions is pre- ly used herein are taken from WCS Technical Report Ni ted m Table 3.21 as a function of the guaranteed minimum 7, Behavior of the WCS 520 k/44 Post-Tensioning S l don strength (P'170) whien is 2002.8 kips. Under Static Loads" ! co~oirior. l aun.cnry .- l \. , Mechanical properties for the various steels used in the
~ aa.a o .,. ui w us is soo. :
type end anchorage components are shown in Table 3.2
-w- : se ,* s.c 3 2. i t.o scos.s various strength levels. Strength levels are listed by equi r., mu s ,,. o, ..~ 2 .-, w. o e a .,,,, ,, wi.,
act sie o.e ison.s hardness on the Rockwell 8 or C scales (Rg or RC) , iso : l . = a .a s w -> aci sie o.i s.or.o quality assurance is based upon determination and cont l . . N=a s . We act sis se i soi .i hardness. Values for mechanical properties shown in 1 3.2 2 are derived from curves contained in Fig. 3.2 6 in v BLE 3.2-1: Tendon load at various conditions presented as a function 1) the curve for ultimate tensile strength (Ftu) vs ha juaranteed mensmum tendon strength (P'170L (Rc or Re) is constructed from information contained 1965 SAE Hancbook,and 2) curves for other mechanical
<eral factors may cause the calculated ultimate load based on erties are plotted as they retate to Ftu based on inforn >
l alytical calculations to differ from the actual ultimate load. contained in MIL HOBK 5 " Metallic Materials and Ele iong these are; a) stress concentrations due to notches for Flight Vehicle Structures", and from appropriate . l flor geometry, b) variation in material strength, and c) specifications, iation in the .srea of material resisting applied Idads. The I i SYM80L ASTM AISI AISI or SAE 4140 et R. MfCHANICAL PROPERTIES (ksi) A7 A36 1025 40 41 42 a3 u l Ultimate Tensile Strength F ., i 60 - 75 58 - 80 55 180 187 193 200 207 Tensile Yield Strength F,, 33 36 36 163 168 173 176 183 Compressive Yield Strength F., 33 7 36 2 36 179 186 192 198 205 Ultimate Shear Strength F., 38 7 37 2' 35 109 113 115 119 1 21 l Shear Yield Strength F., l Ultimate Bearing Strength 7 . F ,, 98 2 95 2 90 326 335 Ju 355 364 l Bearing Yield StrengN ' i 'FS,', 25 6 265 272 280 289 Notes: /[ For e/ D = 2. 0 i Derived using ratio (F,,, ' F, ) os indicated for AISI 1025 times Fn For A7 or A36 TABLE 3.2-2: Mectionical properties of varicus steels used in end anchorego components. Refer to Fig. 3 2-6 for derivattve curves. 0003 267 2
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170 wire Washer; R & D Part No. 730-02; R & D Drawing No. 2 Morerial: 4140 commercial grade, hot finished, leaded, annealed, 6-1/2" diameter bar. Heat Treat after all machining to Re 40-44 0 'e. = > .,0 c , -. . . o" . 0003 268 3
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l I l vl k sioe view "*/4,"n'dj 3QI1' TMt lumraCES OF Cow *e:Suse a st? tmaLL m.fut statts T wa4Y most TMa4 . 0 4 t *, , 170 wire Split Shiry.s; R & D Port No. 730-06; R & D Drcuing No. 3 i Material: A7, hot finished, 2 inch plate. Flame cut 5 inches x 10 inches with 5-5/8 ind diameter hole. 2 pieces per set. Heat Treat: None
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ao a=. c P L,A N to I#1.* 3eUAAE r one n k 170 wire Bearing Plate; R & D Part No. 730-09; R & D Drawing No. 3: Material: A7, hot finished, 4-inch p!:te, flame cut 20-1/2 inches x 20-1/2 inches with 7-1/16 dia:neter center hole. Drill and top four holes for 1 inch diameter x 8 t.p.i. bolt on a 20 inch bolt cir. Heat Trect: None O 'e u =.' e-- 0003 272 7
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0003 273 9 FIG. 3.2 6: Mechanical prooerties vs. hardness. The curve for tonsale ultimate (Feul, snowing UTS clotted against Aockwell Hardness (R RC) is derived ff0m information contained in the 1965 SEA Handbook, pages 107 and 109. curves designeced Fbru.Pbry, Fgy, P ty and Show other mechenecal Dr00erties relative to Fgy and are derived from T40eet 2.2.1.1 and 2.3.1.1 (al of MIL HCBK 5. 8
m I possable failure modes are listed by components in Table and criticality of failure modes must be established as a s '.3, which also shows for each failure mode the t)De of stress. of prototype tests reported in Section 3.3. ulated ultimate load, maximum applied temporary load. J maximum applied long term loads. Safety factors are shown Stresses, strains, and ultimate load capacity of como each applied load and calculated as the ratio of the calcu- influenced by the supporting concrete are dependent up rd ultimate load to the applied load. For each component, strength, elastic modulus, creep characteristics and reir critical failure mode is that having the lowest safety factor. ment in the anchorage zone concrete and are not witr tidity of calculated ultimateloads and resulting safety factors scope of this section. Predicted Mos. Lead Mo . Permanent Fail. Component Foilme Mode Type of Stress UT5 (Ten o. Ovedooo) Lood Mod 4 oo 5F i n. :od 5s C'i'i' Supporting Concrere Anchoreve Zone Principal feasion Bearing f. Interface Compression Tendon Tubing biol Compress;en Failure ;s dependent on mechanico and Anchorage Zone Anchoror;e Zone Itodial Con.p,,ss;on > pDsICol properties of the swpporting concrete and is not cons;dered in this section. Beoring Plate Concrete Interface Compression Internal Flenwrol j Shim Interface Bearing 3527.9 2002.8 1.76 1201.7 2 94 Split Shims Bearing f. Interface
- Beoring 3527,9 2002.8 1.76 1201.7 2.94 No Washer Interface Bearing 3357.7 2002.8 1.68 1201.7 2.79 Yes Composite Waiker Shim laterface
- Bearing 7908.2 2002.8 3.95 1201.7 6.58 No Web
- Sheer and Flexure 2864.4 2002.8 1.43 1201.7 2.38 Yes 9-3/8" Thread Sheer 4342.7 1602.2 2.71 None = No Washer Nut Shim Interface
- Bearing 7908.2 2002.8 3.95 1201.7 6.58 No 9-3/8" Threads Shear 4342.7 1602.2 2.71 None = No 6" Threads with Shims
- Shear 3276.5 2002.8 1.64 1201.7 2.73 Yes f
Q Washer Web Shear and Flexure 2864.4 2002.8 1.43 1201.7 2.38 Yes 6" Threads with Shims
- Sheer 3276.5 2002.8 f.64 1201.7 2.73 No I
TA8LE 3.2 3; Possible Failure Mooes of 2.0 Moo /170 W System End Anchorses Cornponents. Safety Factor (S.F.) is the predicted ulter ioad dev oed by tne sooned loed. + indicates forure rnodes to be tested. 2.2 SPLIT SHIM BEARING PLATE INTERFACE (Ref. Fig. 3.2 7) 0625 8 Nominal Area: A* = 10.0 *
- 4
, = 60.83 so. in.
Rupture Factor: k,= 1.0 Fe, . 36 ksi for A36 per Table 3.2 2. . 9 Fey = 32.4 ksi LOADING FORMULATE FOR LOAD STR ESS CONDITION P or f (kipsi (ksi) REMARKS Calculated UTS P=FxA; 3527.9 58 > 3004.2 Predicted UTS P = F x A', x 1/k, 3527.9 58 > 3004.2 Proo? Test Load f = P + A; 2002.8 32.93 < 33 Jacking & G Unloaded till trans. Anenoring 1402.0 23.05 < 32.4 = .9 F,y D h e, Max. Final 1201.7 19.76 t n i
. , .i,t.., I 0003 274 9
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' ..WASweQ MUT" d y s ecTioM A-A'-
- /l-f PIG.12 7: Arrangement of Anchorage Components
.3 COMPOSITE WASHER . SPLIT SHIM INTERFACE (REF. FIG. 3.2 7)
( Nominal Area: A' = # '
~ $~0 I = 44.18 sq. in.
Rupture Factor: k, = 1.0 F ev = 36 ksi for split shim LOADING FORMULATE FOR LOAO STRESS CONDITION P or f (kips) (ksi) REMARKS Calcule'ed UTS P=F=A', 3357.7 76 @ > 3004.2 Predicted UTS P = F = A', x 1/k, 3357.7 > 3004.2 76 @ Proof Test Load f = P + A', 2002.8 <F = 179 for R C 45.33 @ ey l Jacking - - No load till trans. l Anchoring 1402.0 31.73 < 32.4 = .9 F,, Max Final 1201.7 ~ 27.20 1I U l Nots: t 0 equivalent D = 9'3 6 = 1.875
@ 2 , , f equivalent e = t + 8
= 2 + 'j = 2.94 for t = 2.0 min. , e
- z. e/o = ,2. .a7s
= 1.57 > i.5 Fe,u tfor e/D = 1.5) = .8 Foru (for e/O = 2.0) 0003 275
,' ,Foru,(for e/O = 1.5) = .8 x 95 = 76 ksi Ig , i> i.n;;;
10
i f s
@ The bearing stress of 45.33 at Proof Test Load exceeds Fey = 36 ksi for the A7 or A36 split snims wnich would tt i
fore be expected to show permanent deformations. The split snims do not require a proof load test and the be stress is well below ey F of 4140 steel at Re 40.
@ For the composite wamer side of tne interface F,,, = 326 ksi and F ey = 179 ksi. !.4 9 3/8 0.D. THREAC A'5 = I
- UI * * * ; where:
2 A*, = nominal shear area L, = length of thread engagement
- 3.250 inches n = number of threads per inch =4 p = pitch = 1 + n = 0.250 inches O = nominal diameter = 9-3/8 = 9.375 inches E = nominal pitch diameter
= 0 0.3 p = 9.375. (0.3 x 0.250) = 9.300 inches A', = (3 25 0 25) x 3.1416 x 9.30 = 43 83 m. in.
2 p*=F g g, ,Pxk,
. "r As P' = Predicted ultimate load F,, = Ultimate shear strength (See Table 3.2 2) k, = Rupture factor = 1.1 (Ref.: WCS Technical Report No. 7) f, = Calculated shear stress 0003 ,76 2 LOADING HARD. LOAD STRESS CONDITION Re (kips) (ksi) R EMAR KS Calculated UTS@ 40 4777.0 109 > 3002.4 Predicted UTS 40 4342.7 109 41 4502.1 113 42 4581.7 115 43 4741.1 119 44 4820.8 121 Proof Test Load 40 2002.8 50.27 < (.9 F,y = .9 x ,9 x F,, = 88.3)
Jacking @ 40 1602.2 40.21 < 0.4 F,, Notes: h Calculated UTS does not make use of k,, *. P, = F,, x A,'
@ The 9 3/8 thread is unloaded after transfer 11 L-
2.5 6 INCH 0.D. THREAD . WITHOUT SHIMS (REF. FIG. 3.2 7) A,' = (L, . p x x x E ; where: A', = nominal shear area L, = length of thread engagement = 3.250 incnes n = number of threads per inch =4 p = pitch = 1 + n = 0.250 incnes D = nominal diameter = 6.000 inches E = nominal pitch diameter
= 0 0.3 p = 6.0 - (0.3 x 0.250) = 5.925 inches A' = ( .25 0.25) x 3.1416 x 5.925 = 27.92 so. in.
Calculated Ultimate Load: P, = P, = F,, x A',
- Predicted Ultimate Loads and Stresses:
P' = P'
, = F, x A*' ; and f, = ' ; where f
F,, = Ultimate shear strength (See Table 3.2 2) k, = Rupture factor = 1.1 (Ref. WCS Technical Reoprt No. 7) f, = Calculated shear stress LOADING HARD. LOAO STR ESS CON 0lTION Re (kios) (ksi) REMARKS Calculated UTS 40 3043.4 109 > 3004.2 Predicted UTS 40 2766.7 109 41 2868.2 113 I l 42 2919.0 115 43 3020.5 119 l 44 3071.3 121 Proof Test Load - 2002.8 78.9 < .9 F,y (.9 x .9 x 109 = 88.3) Jacking - 1602.2 63.1 = 58 x F,, (min.) Anchoring - 1402.0 55.2 =51x ; Max. Final - 1201.7 47.3 = .43 x 1I 12
.6 6 INCH O.D. THREADS . WITH SHIMS (Ref. Fig. 3.2 7)
The 6 washer bears on the solit shims over an area An, and thus results in a ferce, Po, = fer X Abr. This bearing force (f clus the shear force (P ) as derived in Section 3.2.5 reacts against any acDfied load (P), so that: P = P, + oP ,. At ultimate I levels, all component mater *als are stressed within the plastic range, and we can expect fbr to be equal to Feru which is a
'agn in terms of Feu, but is ratner indeterminate. MIL HDBK 5 gives data for For, of AISI 1025 steel for e/D = 2.0, but n the average stress at ultimate rather than the peak stress since it is based on a round pin of diameter D in a slightly oversi , ole. If we assume a sinusoidal stress distribution Fbru (averaget = 0.636 Furu (pesk), or Fbru (peak) = 1.57 Fbru (aver:
Data for AISI 1025 steel indicates that Fbru = (90 + 55) F,, = 1.64 Feu. The wMer.solit shim interface is a plane sur-unere it can be assumed that average and peak bearing stresses are tne same.e F , for either A7 or A36 steel can be determi soproximately from the Re hardness. Therefore, we can derive an approximate expression for Por as follows: Po , = fn,, x A ,b= 8.79 Fru; where: Abr = (6.00s 5.6258) = 3.42 sq. in. Fbru
- 1.64 F ru u 1.57 = 2.57 F ru F,, is determined by hardness test from Fig. 3.2-6.
Therefore P' = P', + Por = P', + 8.79 F ru ihear stresses in the threads resulting from an acclied load can be determined in much tne same manner except that for Ic ubstaintially below ultimate, we must assume a relatively uniform stress over the entire bearing surface as follows: P = fa , x A br (total) = 44.18 fer; or fbr = . p ere: 2 8 An, (total) =f(9.375 - 5.625 ) = 44.18 so. in. IO N Therefore: Po , = fe , x Abr " 44.18
- Since: P = P, + Por= P,+.077r' P, = 0.923 P = f, x A, = f, x 27.92 (Section 3.2.5) f, = = 0.033 P 2 92
.7 WIRE HOLE WEB SHEAR 0003 278 As snown in Fig. 3.2 8, shear failure of the web between the were holes can occur along either of two critical paths. The 1 3pplied by the wire heJds to the portion of the wasner inside the shear plane (P ) is less than the total acclied load (P) si 3 art of the load applied by the wires on the shear path is applied to tne portion of the wssher outside the shear plane.1 ratio of load distribution and the number of webs on the shear plane can be determined by inspection of Fig. 3.2 8. wt pves the following results:
. WIRES RELATIVE SHEAR "\
- TO SHEAR PLANE TOTAL LOAD RATIO NUMBER OF STRESS RA*
PATH INSIDE OUTSIDE WIRES Ro = P,i P WEBS (N) R, = Ro /N Shear Path 1 141 29 170 0.829 44 0.0189 Shear Path 2 133 37 170 0.782 40 0.0196 or any given condition, the web widtn (w), the washer thickness (t) and the applied load (P) are constants, so that the c O :uted snear stress (f,) varies
) t directly wi h the o stress ratio (R, = R ,N) as follows:
f,= Ab = R,xP R ., P s Nuwst , Y** , w a t
=Rs*C 13 I
l
=,
l It can thus be seen that the shear stress along shear path I will be slightly lower than that along shear path 2, which wil used in following calculations, however, the difference is small and, dde to manufacturing yariables, web shear failure car
- expected to occur along either shear path.
PATHI l e r-O ._ e '.' '90 Gi!!e!:!* O oe
- 0 L - Oo PATP 2 I
k FIO. 3.2-8; Alternate shear paths for were hole web shear f aslure with Path 1 shown above horizontal ( and Path 2 below. Path 2 is slightly more critical than Path 1 While the center to center spacing between adjacent wire holes can vary : 0.010, or 7.5% of the nominal web width (w = 0.I' the total spacing along any line of holes has the same tolerance of : 0.010, which is only 0.2%. Therefore the average we the nominal center to center hole scacing (0.397) minus the hole diameter (0.260 nominal or 0.264 maximum). The are / steel resisting web shear (A,) along the critical Path 2 is therefore: A; i N x w' x = 40 x 0.137 x 3.750 = 20.55 so. in. As. min. = N x wmin. x t = 40 x 0.133 x 3.750 = 19.95 so. in. N = number of webs along Path 2 = 40 w = nominal web width = 0.39' 0.260 = 0.137 in. wm;n, = minimum web width = 0.397 0.264 = 0.133 in. t = washe thickness = 3 3/4 = 3.750 in. Calculated ultimate load (P')e and predicted ultimate load (P') are thy same since the rupture factor (kr) is taken as 1.0. Lt and stresses are given by: p*, , F iu x A'_ , ,109 x 20.55 = 2240 kios 0003 279 P' = 2864.4 kios 0.782 f, = R, x P , 0 782 x P = 0.0381 P ksi where: P = any apolied tendon load P' = tendon ultimate tensile strength P', = predicted ultimate shear (test) load R, = load ratio = 0.782 from enart above k, = rupture factor = 1.0 (Ref. WCS Technical Report No. 7) l A, = nominal snear area for Patn 2 = 20.55 so. in. I 14
LOADING HARD. LOAD STRESS CON 0lTION RC (kips) REMARKS O (k si) Predicted UTS 40 2864.4 109 41 2969.5 113 42 3022.1 115 > 3004.2 43 3127.2 119 II 44 3179.7 121 0
' roof Test Load 2002.8 76.2 < .9 F,y (.9 x .9 x 109 = 88.3)
Jacking 1602.2 61.0 = .56 F,, Anchoring 1402.0 53.4 = .49 F., Max. Final ,t 1201.7 45.7 = .42 F su For shear failure along Path 1: Ai = N x w' x t = 4' : 0.137 x 3.750 = 22.61 so. in. A, min. = N x w,,,in, x t = 44 x 0.133 x 3.7% = 21.94 so. in. [ P', = Fsua A,' 109 x 22.61 = 2464.5 kips P' = equivalent tendon ultimate s.rength
=fa..P,,g, , 2464.5
= 29 '2.8 k Ro 0.829 0.829 f, = R o x P , 0.829 x P = .0366 P ksi A s 22.61 2.8 FAILURE MODE ANALYSIS
.I 1 !,
- I 318 63 limits the concrete compressive stress on the bearing Table 3.2 3 lists all failure modes for each component. I es supporting a tendon bearing plate to:
component, the critical failure mode is that having th feo = 0.6 th JAh/A b; but < f't e Safety Factor (S.F.). The purpose of the prototype te ported in Section 3.3 is to determine actual ultimate i customary practice. f eg is considered to be a uniform stress all critical f ailure modes, id the bearing plate thickness is then set to limit flexural ress in the bearing plate to Fry at minimum paranteed tendon timate. Although this procedure results in satisfactory per.
,rmance. it does not represent the actual conditions which <ist and has no significance in analizing the mode of failure hile f e, = P/Ab grves the average stress on the bearing area, e distribution is not uniform in any case. and is dependent Mn the modulus of elasticity, poissons ratio, and creep prop.
0003 280 ties of both the plate and the supporting concrete. The max. O ' sum concrete stress is a function of bearing plate deflection. nce the bearing plate flexural stress could not possibly exceed tv without causing concrete failure. the bearing plate can not it. The failure mode is thus determined by the supporting merete, not the plate. and will not be considered in this N: tion. This subject is covered in WCS Technical Report No. 2. 15
3.3 PROTOTYPE TEST
1.1 DESCRIPTION
OF TEST PROGRAM
, e test program was designed to test the ultimate load, ulti- It should be noted that prototype anchorage hardware e its elongation, and failure mode of 170 wire tendons of sions and material do not agree exactly with drawin gths up to 30'; and to determine the ultimate capacity of manufacturing standards snown in Section 3.4. Even t e individual anchorage hardivare components when loaded tests on the prototype hardware were entirely satisfacto such a menner as to test the critical failure mode of each and conform to design criteria,some dimensions were er moonent, in the interest of standardization. In particular, the th of the washer and composite washers were increasec sting was d!<ided into four categories, Series PL . ultimate 3-3/4 inches to 4 inches in order to conform with the +
id tests of 170 wire tendons and anchoragas 4' long; Series nut; and alloy tubing for the washer nut and alloy bar 4 ultimate load and elongation tests of 170 wire tendons and round) materials are shown as preferred over flame cu
- horages 30' ic , Series B . ultimate shear load of web .ised for prototype hardware. Since these changes are 2neycomb); and Series C . ultimau shear load of 6 inch dia. the conservative side and increase the strength of the rest eter thread. part, the prototype tests are applicable. Predicted failun for the production parts have been estadished by lin-crease of prototype test results to account for the inc 1.2 PROTOTYPE ANCHORAGE HARDWARE strengin due to these changes. All such changes are .
noted in the analysis of each series of tests.
)totype hardware was fabricated in accordance with draw.
is per Figs. 3.21 'ru 3.2 5. Pnor to testing, components re designated by R and D drawing and part numbers. After 3.3.3 TEST SERIES PL tshadvalidated design, final drawing and part numbers were igned. These are listed in Table 3.31 for reference. The objective of Series PL tests was to provide prefir ___ results by loading prototype anchorage hardware to the t e a o i. .a... + o ultimate in such a way as to limit the release of energy
**** 7,*,'"*,' ,,t
[,* l '** l ' C .*** l event of failure of an anchorage hardware compone, j ~
., no.0 24 i o-ai imies . m ,
anchorage hardware components used for: 1) Series 4 , no o. us io.v inic. . a ' (170 wire x 30 foot long tendon ultimate tests),2) 17 I Y' .y a [o. no-o* ssi io.is I mier . = structural tendons in the 4.0 Men test bed, and 3) G. Atomic ultimate tests of both straight and curved tendom j n8LE 3.31: Prototype Anchorage Hardware Desegnation 100' long were tested in this series. A secondary objecti to provide additiona! statistical data in support of the ich prototype washer, washer nut and composite washer was criteria that the anchorage hardware must be stronger tt signed a serial number for identification and record. Tables minimum guaranteed tendon strength (2002.8 kips). 4
- 3 2 thru 3.3 4 show the rnaterial material supplier, enemical objective was to repeatedly test the stressing equipme-alysis, machining practice, heat treatment, and hole diameter load equal to tendon design ultimate, d spacing tolerances for the washer, washer nut and com-isite washers respectively. Also listed in the same tables are The test set up is shown in Fig. 3.31. End anchorage c measured' dimensions and loading history for each serial ware consisting of a washer and washer nut (typical A-e l
mbered part. the right and either a composite washer (typical B ent l 0003 281 l l WAs.st i.uT - l _ c.. sns eis,,0. os.s. _ g , $ 0 M /' / V U
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i a. aveaes i_ eee see i PL 8 See.es, l Senee a l f oer l Ser.es l Ceae'ee I see I t & C I ason.e f si J0t $ tog J73 si g 020 al 3 390 - PL a 5e . ee 102 S. M2 .074 31 7 .023 as 3 390 - PL 3 A-6 a-2 Se . Se 003 3. M6 #5 31 3 320 41.5 390 - PL-7 to, ob 004 5 M4 076 . 31 4 a2.0 01 5 405 - PL-3 ee. % 005 S. Mi .30 028 , 01 0 48 0 30s - eL t toe Ya $ M1 072 .014 01 41,3 390 P5 es . % GGP 5. 9'S . Ora 01 5 . 38 2 al 0 Jae - st-e 8I0 ora 3. M4 1s0 CIS Of f 42.3 401 - PL 6 5b 109 5. M2 . 'J3 .01 3 Jia as 3 390 - 8L 2 Ise 31 0 3.90 01 5 Jia 42.0 401 373 C2-7 Oil 5 M3 01 4 . 07 5 a1. 0 304 343 C3.r0 3t 2 1 M5 . 220 01 4 40 $ 385 375 C2-4 31 3 5 M1 of 5 01 2 39 0 270 kJ C2 8 01 4 3. e9e 01 4 220 41 0 3ee 33 2 C1 -4 01 3 3. 38 7 .Q23 Ots at . 3 390 a 85 8 IS IS If IS 5 I 3.M2 74 0147 3th se 23 390 3aJ 6 8 0.Joass 00242 3039a 30344 0 127 $3 6 4 ee
. .77% 34% 23 e 22 2 I to f37 2a TABLE 3.3 2: WASHE R - dirnenssons and load history.
O (U Paef Naa( *a1Mt e NUf Paff asumatte 80 710-Ca: Pe.a* #0130-348 7;aeo Peteone 0 e=.ag No. : 10010s ana rt etal , , as53 41 a2. saaee.ee, a .aea aee ee41 e once, se==ce.e4 yees. Pfene eve e-3/a ;aeae, e.es.ee e , =.me 3 1 r2 .ach esa+c he4e eae saaesses matteint $UPetite i . .. L.e me 5eees Ce=e e Cmeaa<4L aNatv$ss ( . 6, C Ma r.>P' 5 53 Ce Me 8b
.A ?6 Jos 023 .240 90 15 -
aaaC =5NsNC a*wt* 'aeth ae saa 'si.oes on e eas=,eaneae6 eg no lose.e. Mt af fetaf Pee MIL se-68730. pre.ece..e saaesence fueasse fe hervea.ag one *e=ese.ag; sietwiee ag e.3 8ee e.sackag, me neaese a ewencaos ee Sa/15
'>sc a M $$ a .asae.cnel . 3-l/2' 'eg* aeeeae+
- eese , 3-l/a' tearse esmeraed ***esen.
I we.,,eee D.= . a. I Lees w. se, e (.. aes r e i i .e,ae. N.se t
'ee eea Mo use l Nees l C'ee eace 4 44.aea l C oe.eee I l l " men se j 5ee.es Tese Geaceet
%=ec 0..=e.e l Ceem i * ,ca s e l A. .or, 0. =e i Nea a.. I a s .. I I e BMN l Lee s l leeIl &&L5er.es
- t. PL l a , asea.e f e 301 9 364 33 023 02' 5 eed Its 021 01 0 40 3 3e3 8t-4 C2 9 J02 e 377 Jes J22 020 $ 855 Ol ' 01 4 al 0 330 363 PL-7 C2-to J03 9 365 Jet 33t :20 5 052 020 01 3 as 9 390 3a3 PL-3 A-I a-2 Joe
- 3ea ima 326 3 340 S 40 0 01 4 00? 373 - pt.$ Se. Se, et, ee. Ib, %
Jos 8 Joe Je* 029 02t 3 870 OJJ 41 7 e00 PL-4 O' 8 332 C2-7 JQ. e la! Ja' 029 023 5 aa3 C23 31 2 40 3 345 . PL 5 !b. ee. es . Io. Se Jo?
- Joe Je* 023 cte 5 4't 432 01 0 40 0 3?$ . PL t A 2 les All see Ye.e JOB # 3*3 Jt7 031 . J19 3 8ea 023 . 00s at 0 sci 343 PL-+ Cf-4 a 9 8 0 4 8 8 8 8 e $
e e31 ;g g 3274 020s $ soad 023I Ot t e 40 40 387 4 343 8 i OC323 JClas J0282 J038' 7t 's 00532 J034' 721 8 15 11 9 e 03 i7 10 3 le 3 20 *30 2* a t 76 2 34 3 25 TA8LE 3.3 3: WASHER NUT dirnensions and load history. i
'3 e
N. 0003 282 17
O Paar Naad CCurC!ift wasmes Paar Nhme(t 80 73MS. Paas, ao 710 3491 Fw pe , e 0, , No. 100105 MAftAiAL i Ails asat Cmamene.es C ee 8.1/2 :asa e.e-ener pre e ac enaended mAfte:AL suretite someo ne= se ee Corsemaea CMeme:AL ANALY11$ C W P $ Si Ce Me Pb 43 93 .00s . 324 20 . FF .19 - mACMcNeteG Rewoh meen.ae end
- ees en o - -ened e.g.ae ieme. Diese er.il en en sweemee e er.it seems fleer.ag .ae. nemse. ag sea.eae.eae 4.e esie we e. we e -8.egames.ee' a
Mt Af f at AI Per m64.a-ed730 peemse..e e w sseen.ag one eemosaag, s.eewieaa, e.1 ser e,easa ag. Ch.eacn stee .a e e.agee s eveash. ,eesensee se gwenenes a SW33. . . e a easse der a everse feel.aees =. HOLE DIAMdite ah.4e sheetes =.sh heie m.o e-eeers eras *C N. Ce' gm.ges asa se men ==== =easi s.se of =$e ease. MCLI CINfta SPACING i mm.= . 3 0 en e,eer wais .at seee wins e .330 en saaee wall e.el see. NaC ant 15 : 3 3/4.aene esis 3 8/4 h sheese ion,sh
*aseau.of Ol=ea..eas tese es..eerv $*"*O *'" N888 CI**eae* **e8easie O'*"*'9' M See.es see.es feie Wes Co el A's> 0.e Am.es A, . areN m. p( a ges ggg a, . f.,
D 9 Jet .30 .030 .J21 es . 3 390 341 88 2 3D2 9.372 .29 . 230 OIS 40.3 383 341 31.s 003 9 24 .075 .230 . 0t 7 a2. 0 ace . PL.i see m 9 h3 .74 .328 . 01 7 *I . 2 24 341 81 3 005 9 34 .22 . C29 . 01 4 a2.0 40t 332 42-e 006 9 362 .72 .03 . 01 7 40.0 373 . ft.2 Bee 00F 9.h9 .W1 . 323 . 01 4 40.0 375 . PL-4 ft.7 Se. Sb. 5s. es. ee. Se. fu 300 9 h3 .71 . 330 . 0t 7 40.0 373 . Pt.3 A. A2 '10. et, ed. 96. % 009 9. h 9 25 232 . 01 8 42.6 aos 375 42-3 01 0 9 364 .22 .029 . 05 9 40.0 P3 . Oil 9.24 .73 Q30 .223 a2.0 40s . 01 2 9 20 .unt .332 . 01 8 se . 0 Joe . 01 3 9 342 070 223 .010 a0.3 383 . 01 4 9 22 .C73 .c27 .orp 41.0 JB0 . 01 5 9.334 .370 -029
. 020 40.0 373 .
I a is is is is is is s I 9.24 . 2 13 .0293 .Gire .0 95 3si.3 33 0 e .32364 .30334 .00173 .30234 .369 t 0. 83 13.21 / e . Ql9 4 S3 4. 0a 14 2 2. t 2 2.30 3.77 TA8LE 3.14: COMPOSITE WASHE R . dimensions and load history. 0003 283 O
~; .. \ $ 9 3.s.
18
1 i ssher an1 washer nut on the left were connected by 170-0.250 anchored by rreans of button-heads to prototype anc ch diameter wires 4 feet long with buttoutads. The washer hardware, i the right hand side of the 4 foot tendon was inserted thru
> lock consisting of eight bearing plates (R&D Part No. 730-09 The obiectms of Senes A tests were to determine: 1) th
, r Fig. 3.2 5). After installation of the washer nut, shims elongation characteristics uo to tendon ultimate, and ! t&D ikart No. 730 06 per Fig. 3.2 4) were installed on the mode of failure. The maximum force which a multi wire - i ft hand side as spacers and the 1000 ton stressing jack was can resist. tencon ultimate, is (nat force at whicn 2 3* l .nnected to the washer nut on the right. In the initial tests of wires fait (3 to 6 wire failures for a 170 wire tendon l is series. the jacking load was increased until failure of two though the remaining wires will continue to elongate t l three wires occurred. in later tests the load was stopped at duced force. The numoer of wires whicn fail at each inc 00 k. All anchorage hardware tested in Series PL was sub. elongation increment can be expected to follow a norr' auently reused in other tests. Elongations at failure were not tnbution curve imposing increasing shock loads and
- orded. Series PL test results are summarized in Table 3.3-5. energy reiease. Since this would nsk injury to test persont visitori. and damage the testing equipment, all withot
. . v n . r ' 'w a '
tnbuting any additional significant information, Senes f o. . h s 't.o l Y .,. l were terminated at the total elongation whicn produce { t... n- . % . i c ~ ~ , . . , '- M.* wire failures. i *.J9 41) ime . . Jc3 Ste oor l M.I 1 3o-47! 20.o . . 306 Jo# Gor l s.2 i.2o il riso . . oas x: oo3 Tests were conducted using the 4.0 million pound capaci s.* r.+r l_ tim , l. zr aoi j ooi bed and the 1000 ton capacity stressing ecuspment. T-45 4.r.or 2:10 .m w ,. cas l ar* Set up is shown schematically in Fig. 3.3-2 and by photo r i 7 $7 3c2 which are typical of both A 1 and A.2 tests iFig. 3.3-sBLE 3.}5. Summary of Series PL Test Results. 3.3 7) An assembled and banded 170 wire tendon h. prototype composite washer on the east end and a pro 3.4 TEST SERIES A washer on the west end was installed in the test oed fro to west. A prototype washer nut was installed on the we ries A tests were conducted on 30 foot nominal lengtn and the tendon centered ready for test. Two 4 enick t. aight tendons made up of 170 wires of 0.250 inch diameter plates are used under the anchorage hardware at both ( nd-o' out -s e e. gstsgo WCST F_ND AsT Mo
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'S 0003 284
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gy, s.- g , . _ - r l FIG. 3.3 3: Series A.170 wire banded tendon installed through FIG. 3.3-4 Senes A. Stressmg lack attached to tendon a center hole of 4.0 Meo test bed. West end for Phase i elongation.
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! FIG. 3.3-5: Senes A. West end after Phase i elongation. show- FIG. 3.3-6: Series A. Prototype comoosite washer at ea I ing 14" to 16" shams is place retaining etongation of prototype end at the same test stage as that shown in Fig. 3.3-5. T; washer - washer nut anchorage hardware, were deflector plate, shown in place, is removed onor to i stallation of 1,000 ton Stressing Ram for Phase il elongatio
#;;7 0003 285 w cr e .- t ,
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i FIG. 3.3 7: Senes A. West end dunng test A-2. Two wires
~
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have failed at a force of 2.054k at an elongation of 17.0 incnes
.+ -
~'
h .' ., w2,'
' and can be seen against the dettector ptates. At this stage of 7.:#d M Y' ~ Mb ._
** ~
the test. Phase !! elongation is being aootied at the opoosite
;;,.d.3h7 - >
g toast) end of the tendon. 20 l I
l er to transfer sheer forces around the oversize (10) holJ in capacity load cell, accurate to 1% full scal 3 traceable tc test bed. These bearing plates are not considered to be a National Bureau of Standards, was added between the w t of the assembly being tested. and the split shims at the left end. Although capacity c g load cell was or. y urw half that of the applied loads, loa 000 ton Stressing Ram having an 8 inch stroke was attached vs either transducut or hydraulic gauge readings showed a he west end for Phase i elongation. (Fig. 3.3 4) At approx. relatioilship to 1000 kips and can be extrapolated linearty tely full 8 inch ram stroke, shims were stacked under the sufficient accuracy. Test data for Series A, test A 1 ans horage hardware to maintain elongation The jack we:s r+ are shown in Tacles 3.3-8 and 3.3-9 respectively.
- ted, blocked by means of a chair extension .oad, re appiied, more shims stacked. This cycle was continued to comp!e. The relationship of the actual properties of wire used for :
i of Phase i elongation, shown in Fig. 3.3-5 just after re. A, tests A 1 and A 2 as comparea to the minimum prop vai of the Stressing Ram. Tendon elongation during Phase I required by ASTM: A 421 can be seen on Fig. 3.3 8
; designed to be safely below tendon ultimate, but sufficient. stress strain curve shown has the same shape as aload-elong.
great that tendon ultimate force could be obtained within curve and is based on the 10 inch gauge length specifi ngle 8 inch stroke of the Stressing Ram attached to the east ASTM: A 421. It can be seen that the wire used has an ai, I for Phase 11 elongation. This was for the safety of person- yield point 16.9% greater, an averaos ultimate 3.1% greatet and protection of the test equipment. an average elongation 52.5% greater than corresponding mums specified by ASTM: A 421. I wire used for Series A tests was testea to determine mean - 2es of actual ultimate strength, yield strength, and elonga-i as shown in Tables 3.3-6 and 3.3 7 for tests A-1 and A-2 C b* * ' * ' # { "* '"' ' '"*** " [ m71 3ectively, 22P (280. 4 M. g, 3 om w. res, . re,.e. si
) lied load was recorded at intermediate values of applied =L igation. Elongation was measured by means of a stael tape, i~ '""***'**'"*'
urate to 0.01 inches, wnich measured Stressing Ram piston Jb rel. Loads were measured by both a calibrated hydraulic 0.2s* ow w,, a s e 0.04,1 a ge and by a calibrated hydraulic pressure transducer having #"* igital read-out. Calibrations were performed with a setup ilar to that shown in Fig. 3.31 except that a 1000 kip s . w A .0t .02 .= .0 , a, wie see rest a.: tNorvlount use esoren:ts s.,,ee, m ,ee5,e.ess,ee,c , e, e oe,= w .2% en.se, PIG. 3.3-8: Stress Strain curve showing meenanical propertie. no p me, af s7se te : 6, .e,w s.,e ,,.e, wire used in Senes A. tests A-1 and A-2, shown superimposet
- I Nummer SA sees (eiesse, e p ele) the theoretic 38 curve for awire having properties per ASTM: A specified rninimums.
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it 12.24 - - 3 37 it pe,t ee 12.00 10. m. 7.0 4.0 2M .; 0 300 0 35 12 f t.02 set isso 3% 0 so 1030 4.1 9.7 14 $15 2400 340 0 ?5 12.00 10.00 . S.4 405 2760 587 0 80 13 > t 12.05 10 45
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Msf A-2 oara f f 82 4.22%) aehoe . + 42u65. 4.s4M 3.,oe Leage s (2003. 4.oq se ag aos if8eerese aree : 38' 21 2. 'as es. in. er s aoM.'Je ser.e4 Pie. 2% GP46f seriel fee. 3929 $ e R. Seveer, .8. sesaser, A.si. sastes
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# FIG.119: 1.oed-Elongation curve of Series A, test A-1 and A.2 l malts, shown superimposed on the theorectical curve for a tena iloi saan tiss 2.00 m seso its3 2. 3o having properties per ASTM: A421 spe:sfied mhernutns, ia m aseo iss2 2.37 1 ea2o t eso I.ss i.5c3 a2 72.o isu 2.7s The summary and arsalysis of Series A test results are tabul
, ,y 2j in Table 3.310. Following procedures established in prior i is.e . '. 3.n Technical Reports, performance is rated by: 1) nominal "a"
o o $$ so, ion. en ee o. ciency - that is, performance of the tendon relative to
'882 8820 ' 87* s.io minimum guaranteed wire properties, and 2) actual efficier ivoo es o ise4 e.de in4 eeso ino p.ao ?.et is, performance of the tendon relatise to actual
" ner.hanical properties. Both nominal and actual efficiencie sie s.co 3,,w,,,,,..,,,,e,,,,,,,.,
ee. ces. io . shown for both ultimate load capacity and ultimate elonga Z Q isso esso ll'l inn s is io ao . using notations defined in T' able 3.3-10. These efficiencie theoretical, for comparison purposes, and cannot be consid . 8 4l 70 28 ' lC i3.00 an exact measure of performance of a multi-wire tendor several valid reasons. 25,so i "sio e 2022 i3.so 2006 M3a 2024 14.co
*"***'* '"' ' First,the mechanical properties of sample wires are determ
"." ijj so, ,,,, ,, e,, i . ,,,,e,,
vsro 2ms is.m by tests on a 10 inch gauge length per requirements of AS
ll,$ A 421. There is no valid correlation of performance base o o is.so s.eisesone.ie so. ==sem= = a 10 inch ge.o length to performance based on much to 96 o 2043 le.6o teme o 20s4 ir ao see.e - so,ie, eses. se : gauge iengths . 385 inchesin Series A. ,
f 971 o E G finse ese seere ==en emeled O r.e-
). '/
Second, a multi-wire tendon c nnot be assur.ied to perfor TABLE 139; Test date for Series A, test A.2. the sum of the individual wine performances due to indivi differences in the wires. since it is obviously impossible f multi wire tendon to be any stronger tr.an the sum of th a load-elongation data fcr Series A, tests A.1 and A.2 is dividual wires, it follows that it must be weaker, since to 1
. tied in Fig. 3.3 9. For clarity, only the resulting curve is equal strength would be a coincidence. Therefore, it is the.
>wn without intermediate points. A theoretical curve for a cally' impossible to have actual efficiency ratios (tRp and i don having mechanical properties per ASTM A 421 mini. greater than 1.0. It further follows the nominal eff. .ici ms is superimposed, as a se m e. I r e., e e c ,e 4 e. .,
s ui.. a ee . ,- w , , s ui.-e %,.e. seu t'e9 r,;e r .e. e j ea,
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a w. 2 2 2 2 2 2 l 4 1 Dess) tore 3 iF.&s 4.444 I .o3e 6.006 1.t io o7 e 6ies) 9.sco 42s . tie 00o5 0.o14 o.029 oI e WI 450 2.409 2.612 483 1.392 2.ei3 is.4 1 3e 68 6 2t o3.o 10.3so 4.199 f .351 1,040 f .l ff I .C toes.o ts.300 4.093 1. 021 S.964 1 323 2* I . 3e 6.sel Neees. P est. ee f A4La 3. 3-4 7 ander me faatt 3.3-9 3' ansee e rasta 3.3 4
't ted. ee f a4LE 3.3-7
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t il e ame.=ee seespymnea d eereen e ame ae4 esengenen of =we (I;) e go g. ie.ym, t; e 4. 0%. h tr e shuseeeeeed eensel eseagueseo ei seamen e eensei eieagenen of ==e (I;') e paese 'e gm. Per a.i.If e 1 (alongenem6 e e.9%i. M.eeedere Er e o.069 e 3as e 26.s1 :aese . per A.2, E;' e ! (eeengenen) e 3.23%7 fbe ease, tv e o Q528 e 384 e to 23 ineees. S' e t, e t;* / ta
! 12 e te e t;' / t;'
TA8t.E 131J; Sumrnery and Analyasof Series A Test Reetsits. 22
tios (nR, and nR E) can only be greater than 1.0 if the wire There is insufficient experimental data available to dr: actugily better than specified minimums. As it relates to valid conclusions as to the tneoretical true efficiency of t timate tendon elongation, General Atomic has taken this into don, that is the relationship of the tendon actual failus
- count by requiring an ultimate elongation of 3.5% for a 30 (or elongation) and the sum of the actual wire properties.
iot gauge length, thus requiring that nRg > 0.875. A would indicate: 1) a relatively high true efficiency for ultimate (tR, = 0.992 to 1.020) with a small variance, the mechanical properties are determined for each coil of a somewhat lower true efficiency for ultimate elongatio. ire in any given lot of material prior to selection (such as a = 0.627 to 0.863) with a large variance. Should this ho ill heat), the quantitative values of any property for all coils in all cases, then it could be expected that a 30 foot to ill follow a normai distribution curve similar to that shown in tendon fabricated from 170 wires having exactly mit
- 49. 3.3-10 for ultimate tensile strengtn. ASTM: A 421 requires properties (that is 11.78 k UST and 4% elongation) cc minimum ultimate tensile strength of 240 ksi (or 11.78 k) for at 1986.7 k and 2.5% elongation, but the confidence 250 inch diameter. Theoretically, all wire shipped could have accuracy of this expectation would be quite low, iis minimum tensile strength and no more. In actual practice 3 wever, this is impossible. In order to limit rejects, the steel From the above discussion, it is reasonable to conclude :
ills must aim to produce a product higner than the mini- Icog test tendons snould exhibit efficiency ratios of nR, ums, as shown by the horizontal position of the vertical line for tensile and nRE > 0.875 for elongation. It must be ex presenting mean tensile value ((}. Approximately all coils however that test tendons fabricated from wire havin
'9.7%) will have a tensile strength within the range of the mum properties would fall below these efficiency ratic ean tensile value plus or minus three standard deviations should be of no concern as the actual strength of all te ~ t 3 o), and therefore the variance of the product, as meas, both test tendons and those used in the structure, willi ed by o, determines how much higher the aimed for value the same frequency distribution as the wire itself.
[) must be over the specified minimum in order to limit jects. To aim for a f which is too high is to risk rejects for Series A tests show that the two tendons tested exceet
.her specified properties, e.g. coils having the highest tensile fication requirements for both ultimate load and elon-rength may be rejected due to low elongations. They also contribute significant information on the be of long multi-wire tendons loaded to ultimate, from s can be seen by reference to Tables 3.3-6 and 7, the variance, more exact criteria and code requirements can eventu. . measured by the coefficient of variation (v), is quite small derived.
er tensile strength, but is four to ten times greater for elonga-an. 3.3.5 TEST SERIES B AND C - GENERAL 'O
%/ In general, Series B tests were conducted to determin (honeycomo) shear ultimate both with split shims (Seri
_ SPE ClrlED MINIMUM TEMSILE and without split shims (Series 82); and Series C test conducted to determine shear ultimate load for the
- diameter thread, which couples the Washer to the Washe p,
g g gg pg y bott' with split shims (Series C1) and without split shims C2). Specific details which apply to each of the four serie TENSILE .5TRENGTH B2, C1 and C2), including discussion, objective, test proc test results and analysis, are presented separately for eact in succeeding sections. I In relation to the anchorage hardware components or blies, the term " outer face" is used to describe the surf which the wire heads bear, that is, the face on which tt is applied; and the term " inner face" is used to descri 7 opposite surface, that is, the face which has a reactive fr y the opposite direction to the applied load. JJ l Of d' The interior wellof the 4 million pound capacity test b. used to apply the test load as shown schematically i 3.3-11 and by photos in Fig. 3.3-12 a) thru c). Three REJECT ton stressing rams were attached to the inside east end x l test bed and were hydraulically interconnected to a st-power unit located on top of the bed. Ram force was tra ted thru a movable load block to the component being y R -5 e __ p K + 3 & _ Lad reaction was provided by a fixed spacer block re
" ~" "
against the inside west ena of the bed. O U ULTMATE TENSILE STREMGTH Redundant determination of the applied test load is pr by means of: 1) a Martin-Decker,12" dial 0-10.00t d s' of m 11 es o e am N:$4 re. hydraulic gauge measuring to 20 psig subdivisions the o sure being applied equally (in parallel) to three identice 2 0003 286
- 2) by a Transducers Inc. Model GP 46F 10,000-7103 hy- but no load cell even close to this capacity was available.
Jiic transducer attached to one ram and reading to 50 ever, failure load as determined by the mean of gauge c nd subdivisions on a Transducers Inc. Model ADX 38 Auto- mined load and transducer indicator determined load is ic Digital indicator. Both the gauge and ADX 38 were sidered to be accurate to at least 1.0 k 2 since; al all rarr anted on the hydraulic control-power unit installed on top identical, b) all rams are connected in parallel by equal le he test bed. Both the hydraulic gauge and the transducer- lines to the hydraulic pump, c) calibrations showed a i cator were calibrated to the capacity of a 1000 ton load relationship, and d) there is close correlation between 5 and transducer indicator calibrations.
) lied test load as measured by the gauge is determined by in order to determine the degree of uniformity throughou tiplying the gauge reading (corrected to the calibration section of the heat treated components, Composite Wa se) by the total effective area of the three rams (An = 3 x Serial No. 002 was sectioned after being tested to web .65 = 637.95 sq. in.). Applied test load as measured by the failure (Series B1, Test 1) and Rockwell C scale hardnes isducer-indicator is determined by multiplying the indicator measured at approximately 100 points across one face.
fing (corrected to the calibration curve) by three. Calibra- center section of a Composite Washer was chosen as bein i by means of an eight miluon pound capacity load cell most critical for uniform heat treat results due to this co alled in lieu of the test assembly would be more accurate i
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e o, e. ee _ :: N. - _ ,.o o.a. ..e se* S artiusa 4 J' '9 ' men eies puts 1.taf eeu.se 4e .. 1G'*s 98 twa 4 af 7ME'%eprTea ses,8 'W* a f Wa cou a e**aesmess, esessa Mascast eartE 10 erteP4 Ttef ORAssenes plG.1311: Schematie irsvang of Test Sed setuo for Sener. 8 and C tes s. showing general arrangernent and general detail of test fixture. Actual detaal of test fb .ure varies for eacn senes and is shown sacerately for each specific senes. i jf3.
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m-FIG. 3.312: Test setuo (a) for Series 8 and C tests snowing 1,0i
.y - rams, movable load block, and fixed soacer block in well of te-I- . '
Three rams and movaole load b6cck are snown enlarged in (c
!. Q.cw
.," ' movaoie load block, test soecimen and fixed spacer block sho rarged in (b).
."~ '
f.* h:' g,n.. .e. . ( t
. 4 .-
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- 1. \
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q .' I r . l i..- i , h*,rh kt; 7 ...
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4
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.. m int having the greatest dimensions and mass. The location of a section tested and iso-hardness lines are shown in Fig. 3.3-13.
nrdness distribution was approximately as expected. The oo s west hardness of RC 30 occurs in the center of mass of the I
/'
o3SS~0yjSg
.c a.
inulus outside the critical shear path at a location where Ak lEj50 *Goer!3E c Ak resses are low and ductility is of more importance than 1 A 0--- - A rength. Hardness along the shear path shows a mean value of 2:.f $7 E~# 2 c 38.2 for the same component where predicted ultimate ****^0*' as based on a value of RC 40.5 as measured on the outer face. {our:f3kogssy #
'gS isinteresting to note that the ratio of average measured hard- $0 0 iss along thr. shear patn to measured hardness on the outer ce (38.2 + 40.5 = 0.943) is quite close to the ratio of pre-cted ultimaa to actual ultimate (2509.7 + 2613 = 0.9605),
dicating that ariation in Fsu, as measured by hardness, ac-
- tunts for most of De small (3.9%) error in predicted ultimate. '
cu En rAc e lis is an example of the type of variable which is conven- [. ** b .*-. 4 i I '"* 9tly hancied by use of a recture factor (k,). 3 f *[, ns (I - i li f i summary of data and results for all Series B and C tests is [, - l !! l!w! t ,I l i l i ; j own in Table 3.311. It can be seen that actual failure loads ( ! ll 1 ll Q a ij i j erage 0.4% higner tnan those predicted by calculation in !g e t' ig / tction 3.2. This extremely small error gives considerable con- j' . - ;*
' i dence in the design and in the assumotions on which it was \ /
O Ised. it can also be seen that the safety factor of 1.5 x min. 2arariteed tendon stre gtn which was estaolished as a pre. 6 m E2 nG
'1: nary criteria for comoonent cesign, is met for all tests ED C N f\"Ak FIG. 3.313. Scaematic drawing of section cut from Codoosite scent Series C2. TNs is of no coocern since the condition sted bv Ser:es C2 Oces 9ot exist in the structure contem. Senat No. 002 after vaving oeen 'oaced to wee f a. lure in tut [
marones values were measured for accrou~stery 100 meints. ated and, :n any eveat, tae recucticn i9 ore iminary s . d.r.
. is .
tien of raaterias aarceen ta'eu;aout Se t-Oss section is as sn a 3ll, ne scnematic iso-narceoss liaea 0003 290
6 l se re., N.y n N. , Omer.e**a O ime e woner wowise Pe.e
<.. s .. ~. i cameo woner ' Ame l 1 e sei.e j hee.c..e l ames Lei I Las i,.e ! i, ,
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- 3 1.e7 - - 302 Ya
- 23Je 1 2483 -40 3332 1 374 36 2 2 $.8.e1 - - 30] vm 2377 3 2442 39 323e f . ele 81 3 3 t S.847 - = cas vs. 2344 0 25ae 09 3120 I 55 4 32 1 3 wee lh.or . * *.we M S.I .e1 - - 009 No 2434 4 2548 4. 5 30 5 1 330 32 2 e j $.I .e7 - - 005 No 2400 2 2518 33 3338 1 317 cl.1 a e- A.e. Sh , . wr.* %== 5. ' .e7 01 4 JOI . Yes 3430.5 3378 .t4 3373 1.e47 Cl.2 10 } S.347 01 1 002 . vn 3443.2 3541 - 2. 2 1548 1 773 C2-8 7 4' No.se Shee, . manue han $.2 47 01 0 005 . No 2903.3 2922 -0e 2922 I 45 9 1 C2 2 0 3 247 01 2 012 3 . N. 217 2 2930 38 2930 1 d43 C2 3 9 3 247 0:3 00] . No 2741. 0 274j 0.1 274$ l.371 N i. E, e (e ..,.eL.es.4.naLei.a. 4t.se,n ,.. - e - , .me+ a,., ..e.cee.
j 2. Gove.e6.ae fansen bisime.e L e e Asaned fee L.ee .e fe.Ivre . e, ; ( t, e 0 429 so , 5 . 81 I 3. se, %,., . w. f ute. . %- e. .,in . s.r a w. 6 re utn-.= . 5002.s. TABL! 1311: Series 81, B2. C1 and C2. Summary of Data and Results. 3.6 TEST SERIES B1 WEB SHEAR WITH SHIMS ab shear is a critical failure mode for both the Composite block and are not considered as part of the components ssher and the comparaole assembly of Washer. Washer Nut. tested except for the Bearing Plate - Split Shim Interfact addition to shear along the critical shear path, low order Section 3.2.2), which is an accurate duplication of actua j uural stresses exist due to bending, resulting in combined ditions. After application of an approximate 400 kip pr aar and flexural tension on the inner face of the washer. The to seat all parts of the loading train, the load is reduced t I fect of flexural tension will be less for the assembly of Washer. kips and any gap existing between the mandrel and Was isher Nut as tension cannot be transmitted in the radial measured and recorded as an indication of degree of ecce [ ection through the 6" thread connecting the components, ity of applied load application. l aulting in a shorter lever arm as compared to the single piece imposite Washer. Therefore the Composite Washer was select. Test results for the three tests of Series 81 are shown in for testing as representing the most critical condition. 3.315 thru 17, and are summarized and analyzed in 3.312 which shows the method of calculating values indic om the calculations (Section 3.2.7) and analysis, there is no The low coefficient of variation (v) for actual test results ; ison to assume any difference in ultimate strength of the web indicated consistency in both components and test procec ! tar failure mode for assemblies either with or without split The small error, .2.93% average, indicates that predicted ims. The principal reasons for testing three assemblies with (P",) are quite accurate but conservative since actua ! fit shimsin Series 81 were to: 1) verify the above assumption loads (P") are higher in all cases. This is probably comparison with results of tests conducted without split to the fact that predicted loads are based on nominal ims (Series 82), 2) establish a minimum ultimate strength shear area (A' = 22.C1 sq. in.) while the actual area (A") the bearing failure mode at the Split Shim . Composite be slightly higher. This is probably why the Rupture F isher interface as analized in Section 3.2.3, and 31 assist in (average kr = 0.971) is less than 1.0. Such a small var alysisof the effect of bearing on the ultimate strength of the between predicted and actual values does not indicati l thread tested with split shims (Series C1)- change in k, = 1.0 for use in identical calculations of si mechanisms designed in the future. e test fixture for Series 81 tests is shown schematically in
- 3. 3.314. The double bearing plates are used to transfer )
tar around the oversize (10 inch diameter) hole in the spacer l f=+ Ca***== *****. '*a** b=' *= *a. t=+ uts t uts tr.) un (P"3 Nye $ i I ... o.ra, %., s., ~ N.,. b, bee.UM m ..I e 31 4 23 0'.7 2509.7 2633 4.0 3.40 J0'5. 2 e . f*- r. . .l .2 2s77 , 2,77., u.2 19 04 1 m3 7 %y v65.. : f . sl.3 2544.0 25 4. 0 25ae - 0. 9 0 991 29 e.. i JN - j F , t e
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<>9 1.44 3197.9
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_ . - - 4 be.es besw. 8.eens. h. . r/s P*
- k. .* 5. ama.=we Eew w Tene a ufs. P" es, . css so F..
= essenes G'a -
p; p..) . (pa/ e.) (m.. p,, $ a, ec/ p, et se swa.a6
"" per ae.awee . sea, nameron m i, 4. 0.829 per amesekee sma.=en a, e 40. F. 109 had G.1114: FInture for Sertes 81 tests. Camconents being tested are 4. seawy km,, s7 85 >=.y Pg. . P. >=.y2002.8 ovvet shaded. Mendret Conforms to the04 illustrated in Fig.1$11 for tn 1 sheer failure mooe. TABLE 1312: Summary Analysis of Series 81 Test Results (I
ys i DL i i: t i i + >- l 26
ULf1 Mart LCae. rt373. PRCTOTYPt ANCMotAGk ompoNf NTS l ULIIMAII LCAId .IIII
- PacICTYPt ANCHCaAGt WMPCMNT1 as as-# ftst # ftst CAft , se a, 6, Stails t#-t ft37 ? TE17 Daft i +f., c ,
CR8MICN wee % . e# , em f.,,.o f Of3CaPflON woo hee, . W. #a .-.s wroNENT DATA COMPONENT C ATA
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C.=es .. *e.ae, i Nuuat 1 5Cata : aaAoiNo
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- n. ..ast fPeOC2 Dust Tt3T P'CCIDbat P4.ees se _ , 400 kipes ne em a m seees Pre 8eed se espremeneecy 400 hies; aonem se sever Moensee goe heemeen esmo.nesete end pemeng Meeere goe messe _ . _ .o one yonsag teketoe ame A,ee (As) e 3 s 212.43 e 637.95 sq. in.
Effechwe an= A=a (Ae) e 3 s 212.65
- 637.95 so. in.
4 0ATA LOAO CATA q I ?t3T CauGI*D 4 fBAN500CfrD ft37 GAUCE'D i Tts 4100Cfri' atA04No l LOApt' I acx.34 l LCAoq) r,ea i 6. es I staoshol Lee) aEMAars atADINo l LCAOt l A03 34 i LCA0Q) wa i &a 1. staciNoi sian) a MAars i i e.e4 ea, , i e p.e.een. ces e . .a si.. i ., iA ,,i i Po . cae e , ,,t.e. I a e, i s.. i r. . I ori ter i e s, i s,s it .. i ,*r a ro r i l,*.. t stos. t s.. s rsr i urr a mm. I stas i ., n ?n n i,re a s t es. i i o .c. n a. i rers o ver i n o t. e is**
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!.4. I t r .d. istniners i r. r i 7 t r. I t re, t es. t ani 15, r i 4,, I tar 4 t evr*: a r. i t. ,. t r. ., I.u. r es A,A. i 2:sr t Hr s ,,. i t i, ,. , , s . . . ,o i 6 l e i s. e e. i s. . , 4ts. I r s ,1, I ser,q etr e ta rr i r. as 1 4. . /. .e v: . I t (tsis', s A.a...o . - 6. a s ene v: . e ! f ( s e -> s s.aaue .a er ,.a o r* e i tsei . 1. sie I s,rs 6 are... 7a v.n er r* . ' tes t . to att i J s a, l ar...., ne, v. w, Neeene Q nye.ma;e f.e c.,,e nye. set fewesesse pie. ADM Digiene ammeeve y,e -
pensene { Mye, niefee one,e .es nye.ne66. f _ -pe. ADM ospese anse mesonemone W me eene 6ess. p e eenwee.w messeemeow et nie e e. ases.
@- Laos e fee, c e - 0.43e
@. Lees - fe o , asesi., s 0.aae
@ Less - Aom.as nessie, s 3 @ L-se e Aox.as ame.;eg n 3 L 1315; Test 81 1 Data FIG 1316: Test 812 cata ULitMAft LOAa. .IITS e PtoTCTYPt ANCHCAAG4 OMPONENTS 45 ee-3 ft31 3 ft37 OAft i **e v 4 9
- RinnoN w . n . eut st* a e '
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) l , iComeCNENT NAAet ,, _ 5t aiAt I NUMat t i $CALA l was0Nt11 i uf5T 4tADINo i %a 0003 292. e Commen. a.mme, ..A . e ,,s a t e J. 4 i s. j 's.. . w..e . _. i - . a,i . [ . ...
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Peeeese se eeemen.seeoly 400 hiaee acom se se , Mmewe goe twesen _ __ e aid ponene Efkenee som A (Ae) e 3 s 212.63 = 437.95 sq. Ia. 0OAfA l *t1TCAyGEM 4 faANTOUCtrD i atA04No 1 ACX-34 i LCA0@ l wa I LCA&l' &I itAoiNol si
- s. i almanus I t w.eemed I a c .i,,# i homes. Gee e a e,#ie.
a *s. e 2,. It..e ere ror i I gde i ,,99 ld..tf9fl ,t g r i I tete 1 799 14..if'd ,*4f I i f*a. I is** i s i t *ft tier e f *
- f. i tite taa is?is tr( e 1 eese i 1<<s i 0 9.* osts c*C e re .ae . :.e.e ' <- w ee j i 4 1 e ___ t s, , #,re t 4se ..w a. c 4 r e I i y ff s) I4',e u. a .e f..s e d Ass I r* e i 1f se a fa st, 1 Jet, eg , 74.,e us c%.. wir
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@ Less e fee On.g. amen.og a 6.638 l @ Lsee e A0x.38 asus.ng e 3 1,1317; Test 813 Cats:
27
l
*. l I
is test results do not prove that Shear Path 1 is more critical The final acceptance criteria established in Section 3.1..
.an Shear Path 2. in contradiction to predictions based on that the proof test load equal to minimum guaranteed ti talysis. since the mandrel used applied load to Shear Path 1 )
UTS must be 90% of the yield point of the weakest f i id therefore forced failure along this path. In f%, examina. mode predicted from statistical analysis of test results. an of the Composite Washers r.b. .'.;;wre indicates that shear is no well defined shear yield point and, in fact. shear ilure was trying to occur along Shear Path 2 in spite of the and shear ultimate probably conicide, so we may conserva ct that the mandrel applied the test load to Shear Path 1. In assume F,y = .9 Fsu. Therefore. final acceptance criteri; veral instances failure started along Shear Path 1 at the outer be expressed as: ce of the Composite Washer (directly under the mandrei), p at ended along Shear Path 2 at the inner face of the washer. 0.9 (f 3 a) x p'" > PPT = P'no = 2002.8 kips eference to Section 3.2.7 shows that minimum couivalent ' ndon UTS for failure along Shear Path 2 should be 0.964 ries that along Shear Path 1. Since the analysis of Section 2002 8 2.7 and the examination of components after failure both I 3 a >0.90 x 0.90
- licate tnat Shear Path 2 is critical, the average minimum
- uivalent tendon UTS of 3062.3 kips should be reduced to: The revised Py (min.) of 2815.0 is 1.14 times greater thai 2472.6 kips required for acceptance of test results,indic Revised P'y (min.) = 0.964 x 3062.3 = 2950.6 that the web shear failure with split shims exceeds requirerr nce the value of standard deviation is not effected by this This series also shows that the bearing at the Split Si irrection,it follows that the lowest equivalent tendon ultimate Bearing Plate interface and at the Solit Shim-Composite W ould be Revised Pi (min.) 3 a = 2950.6 3 x 45.19 Interface are not critical failure modes as both sustained 115.0 kips, thus giving a revised S.F. = 1.41. This correction as high as 2682 kips without failure, on tne conservative side since the actual failure mode is prob-dy along a composite of both shear paths. Figs. 3.3-18 and 19 are photos of the outer and inner respectively of both Composite Washer and Split Shims beingloaded to failure in test 813. They are typical for all l of Series 81 O,
k 5 4 x g
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+ , ba -
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- i. 1318: Outer face of Composite Washer and Sollt Shims of FIG. 3.319: Inner face of Composite Washer and Solit Shims -
813 after being loaded to ultimate. Note that sheer failure is 81-2 after being loaoed to ultimate. Note that shear fadure is tg Path 1 oniv. Photos here and in Fig. 3.319 are tyoical for a. Doth Paths 1 and 2. s in Series 81. 0003 293 28
ULiunart LOAo r*tts . Pec'soTY't ANcHoaAct coa 4PoPaNf1 3.7 TEST SERIES 82. WEB SHEAR WITHOUT SHIMS
" "" "# # " " * * ' " * ~
oucaimoN w si ,- s,..... f ,-r e Composite Washer without split shims could only be used
@ a " fixed end, that is a non-stressed end of a tendon. How-
-comeoNameoAra I ' "W 5 5 '
tr, even for a tendon which will only be stressed from one , " " , ' " ' " .. -.lMa^'
- f, there are advantages to using split shims at both ends. The
! [ ,"'N it shims distribute the force from the washer over a greater I - ----
'a of the bearing plate, thus reducing the flexural moment ,,o ,$i $ $ _ - - -- -
ri and stiffen the bearing plate, both of which permit use of hinner bearing plate than would be allowable without split
- , . , , , Ps.f ,,A,,g ;--} j y,.; je, T.w. 3.
.- , * ,,,"J,,,,,,
"~~ ' * * ' fa*
ms. Using solit shims at both ends of a tendon stressed from * # # '" ly one end allows both bearing plates to be of the same p...,,,w,,,,. --e.-. **'. 0 s"29
. s . 3.<4.A .
,,,,,,,,, 3. . i.0 ckness and further provides a convenient method of taking nnPiccsk *'***
slack in the tendon prior to stressing. 7.c e s. - , .00 6em w v. .
==
.. e *:
En u a A, (A.) .- 3 2it.as . 437.ps .s.. e purpose of the two tests in Series 82 was primarily to pro. LoAo oArA e additional test data on web shear strength and secondarily test cAuc@ ' t.ANsouct r D ! determine if the presence of split shims has a significant ect on web shear failure. acAoiNo W ' bal } toAot g 'E^D'"G'Aox.aitoAo@j'8^'^ars ***' i m ., i ...g i s e.c , ,,. . e test setup is shown schematically in Fig. 3.3-20. Test "* ' '** " * * ' ' * "
,", e r ms. , ,,s. u.. e nri o
>cedures were comparable to those used in Senes 81. Data u.. i , , , , .. ....,f.n ,, . s i each test is shown in Figs. 3.3 21 and 22 and is summarized a v i, isu. i ... . ..s, e r e f i J analized in Table 3.313. - ' - 'en'am ru, i s. ., . s~..
I i 8 1 1 %4. Aa .. P... I l I i i d. ,=u g 4e r;* . 4 I
.J (t fed i
_n v. . ssur- I o. nu ! .,. s r i en.
- n .,. m
.. **} h
.* N @ Mye.e,48. r.e G.,,. .no nye, 44e r. _ pei. Aox oigs e A es e __ w4.
. . ; .* @ L d . f e G.eg. A i., s 0. A3e
..,". p, y.; . . , . . @ L s . Aox.3s a ;.,a 3
/ / e
. 4j ; FIG. 3.321: Test 821 Data i
l d Byr i ULTimeArt toAo itsTs - PeofofYPE ANcMotAC4 ContPOMNfs
\ '
y $Etel 8 t. - 1. MST f. MST oam i A#w 6 7
; y ..:
.; +
% , ;., oncnimoN s si . v.... 6,-:
[ . coneow NroArA a .. , . . . _ , , i coaiPowNrNAw _. I Nua.arn i sc usi seAo.No i l _ , . , c-.- i..,- <._.s,. . 1 i
- 10. 3.3 20: Test setuo for me two tests in Series 82. ,,, M tij i _,,, _ -- - -
- . r. . '. * *; - ' :. ".. F .:. :w r.sa.3.:
-- , , ,/*.,,t, *' ?.i
. . T6 1. h A., . . 22.64 c.e s e ** . e A. e is , ee s.a y P' .
2s as . t , ee l e ( a. ,. 0.829 ( e ufs (P.) e.> yts (PD ufs (P*) NT h. T.is Urs P v., ..
- p 3 g- % . h. . 1.0 I to. u., ni Nu.,u , ,,,cc,r"r.. ., to . " s 2.. .... ..
2-2
.i-. i ..'
2400 2 1 2a00 2 1 2318 i .. i i . ., ,
.~o0 m., - .-
. 33 1 A13 3 fe73. . I t 44 4. e e.e a n o
' . 2; _ 2 2 2 2 2 2 (N n 9A **** Ar (A.) . 3 212.63 637.95.e. i..
%. i 2s27.3 ' 2627.3 2329 3 . 3. 9 t 039 2es2.a t . 43 Loao oArA
. 27 10 27.10 11.3 0 40 a. 006 0 Of 14.33 't!TCAucta i is ANsouctr D
. 1. 3 1.GI 0.45 13.34 0 58 0.58 0.70 aox.as
. ie 27w . 2,w . mo . 70 ., 2,,2. 0 . aEAoie.o w . l toAct o.> Ao No to.Acq) ..> ~~
*3e 25 6.0 254s.0 24 5.0 2.10 1.021 2st 2.7 I .0 ' '* e.d I ac%.i i N'*=.- G * * * *15a -
.?. i v1 e... i,*rt .ra r i ee+. I ersy a 4. i J.f l. ,, s s- i 1 C.6 4.e.e ufs ( P;) . 7, . A4 e w 4, p s., (.,) ? e .. I eva6 6 4.. IMi , .m f l
- 2. '
.e urs (P;) . P;/h g P; . (P . A4/h. g h, . l . 0 E,'.e'.'. ( P; . P1
#'#* ' #' ' * *. 8 7' O * ? 8 J' '
- 3. ,
100 P* >p 196. i tsts e edst 4 3,e t rii ec.....,. f.,,. v ., O4 s. w as e t 9".
- p. a i e
p.m.n n, . p;j P= 4 r uts.P= w 6.n.. ) * ( P'/ t.) . 6 . F 3 8. 4/ P., .d f
-7 6 P* . '
e e I i i i (t fisi i I s..., 4. , ;.,
.n a.. a. a. . O s29 r- . i7r,e
- I +.st. i r.ie i s. . . h, em
- s. aw- -e
. % = ,. . . m s., v._. . m .. v xn .
- 0. 7. . i o, s., N @n
- i. r.e c n,
w-i. r, , Aox os, e n BLE 1313: Summary Acolysis of Series 82 Test Resutts. [."* [ '** FIG. 3.3 22: Test S2,2 Cata 2' l 0003 294
utnmart toao rests - encrofvn anes.oaaca cowoNeNrs staits e t = # fts? 'r ftsf omft 't**.*6,
,s shown in Table 3.3-13, the minimum equivaler : tendon ,
ucaimoM c o o w A..- . .-< Itimate which would be expected from the statistirm' analysir f test results would be 2312.7. For reesons set forth in Section Co**ewNr oara
.3.6. this value should be reduced to give: Py (min.) = 0.964 x 812.7 = 2711.4 kips. This is 1.10 times greater than the min-he,,ey,, %
lNuEn ,,, .n
! $n. b num strength of 2472.6 required by the basic acceptance *=~~~' - ' *" 'C ' "'
@ P== fig. 3.2-e . . . . . . _ .. . . . . _
PuoiCito 848Was LoAo -- 46 = v f! + y comparing the valces of f for Py (min.) as given for Series r. - ' , ' '* ' - as.m s. . as.a > ,,4 a' s,.mi. 6. L8 1 and 82, we can see that the use of split shims gives a 7% 5. % 44,.s.ne.. s.n s .--+-' '],,'*", crease in equivalent ultimate strength, contrary to pre. test ea r; % - - t '. u 6i. s. e.g ,
<pectations. Comparision of Fig. 3.3-20 to 3.314 shows that, s. - h r; ithout shims, the moment arm is greater resulting in higher itst reoctoum exural stress which would redce the ultimate load of the '***d**
. u "."
6 :*=***r#6'".w***""*,"*' asher without shims due to the effect of combined shear and an a 4, u.) - a a 212.as - av.9s . i . nsion stresses. The components after being tested to ulti- , , , , iate were the same as shown in Fig. 3.318 and 19. ,,s, c,uc,m r,4u,oucir o
~
EaolNo l Lc40P aox-a l LoaoQ)
- 'an.) l Si.) RE4otNG l bi ) atmaars i.e e l a. i i P .+ s . G. n
.3.8 s e. 6 -1 .e.. ia-r, ro r s
! TEST SERIES C2 - 6 INCH THREAD WITHOUT SHIMS ,y, ; , , , , ,,,, i.,r ,,r ' i e... 4 ,, , r, i .. i nri r u r i tries C2 is reported out cf wouence, that is before Series C1, ste. iesc6 i=~ ' veri n er i order that C2 results at.d analysis may be used in analizing d'd* "' ' "* ' 'e ri t
- n e tries C1. Fnr the same reasons discussed in Section 3.3.7. the
.sembly of a Washer and Washer Nut would normally be used ,. , l*"*[*"",**~"*"**'
, f, . e , % ,, ,,. f, ,,,
. conjunction with split shims.
- i. i.
l N.sms @ per ee,6 Tm om g. .nd pop.sm d. f. _ p.m aan osg6=a a. i l 9e primary objective of Series C2 tests is to determine the r""""* d"*='*"'- timate shear capacity of the 6" O.0. threads (P'), isolated om the effect of additional load capacity resulting from bear-f [ ,* "'**'*',' l g on the split shims (P'er). The secondary objective is to FIG. 3.3 24: Test C210sta ,
' ovide relevant data for condition where it might be advan- '
geous to use a Wasner Wasner Nut assembly without split ,, , , , , , , , c, ,3 iims on the fixed (non stressing) and of a tendon. sEtits e t - t itsf d itsf omit t atew 67 9e test setup is shown schematically in Fig. 3.3-23, and test osscaimoN coon - w. . - h-i ocedures were similar to those described for Series B1. co.,ogu, o ,, sania6 i .asowss ist data for each of the three Scries C2 tests are shown component N4W I NUMaitI SCAW IE 4omG parately in Figs. 3.3 24 through 26. Summary and analysis test results is contained in Table 3.3-14 which also shows
" Z"5*'] , _
l } ' [ l l7[ 4 i . e method used to calculate tabulated values. @ '== 7's. 3. 2-+ - a==. I reoecTto sanwas Loao __ . . -. 4; . v. n . r; . f.4 A . ts.m e, . as.a e I m.. . s s et i ha. 6
% * % a 44.
8.79 F. . S.n n = -.- F,. *.T. hw
,;, , 3, ,, ,,
r . r, r - 1 sos s ki.. g, g,y , 6.= Fl 7, e
-. W ,. -
i
. x
[:. ' , Test reoctoust ei _ , a 6i., . , r n,
.. 7 -
. * .* ue s.n a, w p, w.
,'. "e
., tra A u)-
312.as i sv.'s . i.. U3 Ay /j
-N Lono cara
'/- N,' rest oavca+ i reausouctrP j aox 3e g / /
- /,
saomo () l Loaot as.) l staomoj tomoq) es. armaars i . a. . ee . G., . *
// % *s. t sat i s .1. .,r re, e
*. .. '/// < ion. i 11. . s o.o ! su s is e n i
.[ f ;
M' V.?:.%'.
< *'.v e ses, e ave.
391. s22ws i s.. q rs te..e,*rt i.o r e nor i
.t c~ .
x .'- .. e . .. i s,.s . .....,r! ,,,r i
", , y, . . . u .. .en, , . a,r i 2,. r v....., - r u . , W
, , , , i
- one,,g, .
.,,,,,,,,,, e- .. e n n .s ! A n.., - A. .
t , i~ i G.13 23: Test estuo for the three tests in Series C2. Nea" O w f* C=,5= * "r*e.u. r: p aan oi,=4 a. p m.o e - .i mm. e i e.
@ L a . fee ce,g. a s g s e.63e
@ L s e 4Dx=30 A n g a 3 h g ' :y(. nG. a.a.:s. Test c2 2 c.ta 30
As discussed in the analysis of Series 81 and B2, the i utrinars tomo rests . ewrotves anc>.oaaca ccaarows s i value for minimum equivalent UTS (P'r min.) represen minimum strength expected in a population of Was, er.V j set-1 fast 9 rest Dart 1 Add y 6 7 Nut assemblies as derived from a statistical analysis c airtioN 4* o O M es . V'~~' #"'"' results, is based on 90minal shear Jrea and is corrected I soNest oara shearstrength corresponding to the lowest value of Rc al j , i samA6 i so~m i urrP by the quality assurance provisions established for the 2.0
.courowNrNud I nuusta i scusi asaomo i s'
. 3 .e , . ,,s .A . i.. .o..., 170 W Post. Tensioning Sistem. The value of x . 3 a f (min.) is shown to be 2668.2 kips which .as 1.08 time
.. __ i .,, . c . ,.r , , , . ,
W '
! i 2472.6 kips established as the minimum by the basic c T* rie. 3.2-4 "'**
for acceptance. The mean value of 1.08 for the revised creo saawas Loao 44 v.93 m. m.
. ' ' N . 2s.m v 2s.3s e /,a. . ted h. . i . , Rupture Factor (kr. ) is quite close to k, = 1.1 used i
. " *e a;. . e .7, ,,, . e.7, . *= " '**'* 3 8 *'
. +
- u. . 3'*. <2 m . . dicting UTS. Photos of components after being tested in t r . at. . t m. mi ,,, 3,37 ,,, g,,;,,,,3 are shown in F. .ig s. 3.3-27 and 28.
- a. . ri 3.24 e.ociouse P eems .. . 400 bie., tea sees u: . e h, are n 4, ca.) 3 . Zi2.es e 437.'s . ..
3Dara
- ff17 CAUGIC fil4NtouCf rD acx.2e i Loao@
seaoiNo 4> sw I tomot =AoiNo l s. a umass: 6 he s I ac ,gi beamed. Gee . .e .
*&. I 4,* lie, ie.fi for
/see f es *9 (Jeb429tt.sser t e.e 6 r , g t, t6 ifort f. p f-e f 9 f. I t 1*h i A. 179f t tidf 1 E' , .
- I 1.F7 19,$" 1Tdf l#s,..ma '
. T essa .ne44 '.
t i u. da.se %eso. a * '~# t i t h
- r. . i vu .o I i (~ 1
-e .=_.
e ,. ,,e- . ~ r- - p . .i se .m lead. . ' N ; h.=12.t T t e . f.e o .' fO ( @ tow . anx.3e n.wa., a 3 a wi., e, o.43e
*.=
.3.326: Test C2 3 Cata FIG. 3.3 71: Outer face of Wasner Serial No. 010 and Wasner Serial No. 005 after being loaded to ultimate in Test C21.
c s aa e, a na . k.. . see ,
.. urs (P. ) urs (PO uts (P") N.,t .. uts (Pi d %
.., ni o.>. t m. > o.> re % t t o.,) . N r?
l-1 31'1 e 2,ro 3 2922 -Oe t op 2704 I 1.39 12 30't 2930 f 06 ' Nif 2 , 3.8 2877.2 f . e4 ** l.3 3 01 s. 27as '1 3 2745 l . 0.1 1.10 2770 4 I 38 ._ ' a 3 3 3 3 3 3 3 ,V f
'3 .s 3102.6 2 20 e 2Ses.7
- I .5 1.3 20t0 e t . 40 ~-
e 73.00 2.33 ee.38 45.39 1.44 . 01 7 47.45 02 ./^
. 2.33 2. M 109.3 t 37 9.69 I 87 3e 2321.4 30t t 7 31 21.3 4.42 f .13 2992.9 1. as 3e iss3.4 2421.4 2409 5 3.42 l.03 2ees.2 f .32 i ! %1 '
- l. ces e sb urs Ir..) . s issa,.0 e a; ..m e .# sh Leme,.
8 e th ). F won a) me .m .h ya h s . een,ed .eh e
.s e. . a; ,
27.92 ee. .
- 2. P s. e lh e ufs ( P'.) . F (.s.we4 e a;/h .. h, 1.1 f.es sese e ,
af 4er.eee m . g, l 3. I, , . ( P.' . P' }/ 8".. N., , sa .
- 3. epensee .h
- c ..w. .
4 to e Ib.or bem e P.s.ee (k .) . Pj.g P*. -
- s. Ae ms. E a e uts ( Pi )i. P" go e .e F. 75 6 .) . P" e 7 s.ed
/ P., i e ee) . P* . 109/ P., Isen O.
- e. se %.e, (s.r.) . et t .v .e e se. uts s P. rt WV O SLE 1314: Summary Analvsse of Series C2 Test Resuits. ,
. .w
'Aytfr,;
FIG 3.128: Inner face of comoonents snown in F^.g. 3.; Both pnotes are typical for Series C2 components after fa' , 0003 296
ULftmArt LOAo fif73. PECTOfvFt ANCMoaAGE COwCNENTS H.N5 C,.s 4 37 4 417 0An e u av .'
.3.9 TEST SERIES C1 6 INCH THREAD WITH SHIMS otsCairr,oM .. o o n . ./... f.... .
CoaneoMNT DATA he test setup for Series C1 is shown in Fig. 3.3 29. Test u ,,,, , ,,,,oy 3 3
.roCedures were sirnilar to those previously described. Data for IcewnwNr NAMt l NuMata i sCAtsi as,omG he two tests of Series C1 is contained in Fig's. 3.3-30 and 31: ' ' ' '
. . ~' ". . . " . ' .' .. ' . . " s.'
nd a summary analysis of the data is Contained in Table n,... s . - .,, . 4 1.3-15. O
- 8's. 3.2-+ ._. ..-. . - . . . - .,
,sonCTto ,uwas LcAo _ . - ... . A4 17.,2.
r.. '"*. ,s.s.,e,.
,s .s. ,,i. . e e.v. t kw b, - 3.1
, mt n, . , e n . ..rs ,. . e.rs a s, . . r. , .
g, w r . r,. n. . s 1.. I k.e ,_.,_,,,,,
,. . s rn.
rt37 PeOCf Due
.J e v' e p ", P es ee e +y .00 6;,., a, f Meow.o ese b.
e,.s.w.en:
.. N / (twn a A, (A.) . 3 s_ se 21 3 . 31. 95 eg. i..
. I e.
f
- LOAo oarA
*.* -(
- g
'ww _
.-..*- I rtstoAucem i r eausouct r i' s - - , i // -
; i uAomo i scAat' i Aox.,a t toAom
, '/ '
. r,. I s. > l nAomGI .i uMAus I .. . ___
! i ie d e Ae.w.s , P s. Gee . ..
vipg// N
- n. , ... ,=...c, c,r ,
m .e ' :' ! s a <. i,,.? ,...,t,rI ,,e r ,
, , f
.y 3 m '& ..
/x M , e ... i ,4., i ,... i .. ri , , s c t 3 , , . . , ,,,. ,,.c.,,,, ,,,c i o -
_A,,,
.,, f. , , . . i ,,,, ,,,,..,,,,, ,,,,
. , i
,,.._.,.z..,...
v , , ,, v., . .. . . < . j besses h Mye.e.66. Te.# G. age one h.ee f ph. ADN Os,. a se. l fe e.it .f li.e es le.d. slG.13 29; h L.ed . I f G ,e % a e.eae Te.t setuo for the two toets of Series C1. g w . ,,,,3, ,,,,,,,, , , FIG.13 30: Test C11 Data
)
ULinMAft LCAo filfs . PeCTCTYPE ANCHCaAG8 COMPONENTS j stads C,.t. Tilf te TE17 0 Aft J ef.v .7 I Ot3 Cat,f1ON .*oo % e. .,. [. . # CowowNf oArA l l M B, AL , sano~t s3 mes u,1 &ies i men.s .r.or se e a . no lose.P CowoNEp.f NAME I NUMata 8 5CAtti asaomo
~ a -
t m.,
- T I _,-r ,,.1 ,_ . ,r I u,,..-, ~
C. un _ ,
, ,,r < s., ~s _ .~., . _
. . , , . m . , , .
. . . , ,m , . , .
m. C,., c,., ,,,,,,.,.
- , , , . . , ; ,. m.., i . .. ...e. i ,,, im.,. .
s ,.. . - e , ,. .. .. . _. ... . _ . . . .
, _ ,z.
Cn. . . . . . _ . .y.,,,,. m., matLeA.~ , , , . , , , , . , ,. ., ,, ,. ,, . , , , , . . . . ,
. . .., .,, .. ,. ... ,,, . r. . ,_ . ~. . . m .. . . o . , . . . .o s.,..,.,
, . . . -,, a. . , , , . .,, ,, . ,,, . r, . n. . ,.., ,.,,,,,
, . . . ., .. m,. , , , . . . , , . ... ,,,
,. . m ,:,
filf peoC10ust
**"'.*. ' 8 e sheer utt , P ) i n.as . _ . . _ ,s es no. ,eae C,. ,8 p 6e.e e. es,, e, .00 b.a.s a.eu. se ,
M4 I be Ie.4e .. 4. * .h. .e of 8. tes.ws4 se . 6.s. p.ne.o Thes.sen.
.l .)
- 75IWe b 9 Ms.twe. goe b.e e. . . .c g d sh, P. * 'e 8 8. .eae.l/ l# gtw,. a A, u.) . 3 ,12. 3 . 37.95 eg. en .
e . C. e 4.
.. m s.e=e .me ..=e,,uit
. (M.I e , .e c. . ..8..~i (8.e . sh=.8 . e .a .,,..en. s.a .. m,., lea o.,A 3- P*M*% ril,GauGEO , f aANSD,JCf r O j
'~/ 2 "' L r.
. .. ... . . S'.~a j ulay l ,,a;=o l T.aM ,,,,,,, ,
, m...,.,,,. i - A- ' -.e. c- 4.
.. . . , , .... ...<, c, , ,
'A8d 3.315: SuWnelm of Series C1 Test Results. ,, , ,,,. i ,,y.fi ,,,c ,
, . , . , ,..A ....,,..i.... ,
,,.. , ,,.. . 4,<, ,,,, , -
c .. ,n...
.....,-c ,,.< ,
,,,..~.,c,,,< , Q
. ~ . , . , , . ,..<....c,,,, ,
e... ,,,n w.....,,... , c... - , . n . , Pe O m,e.ed.e f e Go., .e mee 6e fe.=sumer pese ADE Dig.=4 an.
. e., _. .
@ L e . r., o.,, a e6., a e.aas g,,,,,
@ . Aox.se - ;ea .,,a #.,
- J fo r , . ,
;GC : ..!. F,G. 3.3 31: Test C12 Data t\ i .' i < ' 32
ies C1 tests allow analysis of the total ultimate strength the design approach. However variance is relatively hiq
-) of an assembly of Washer. Washer Nut bearing on Split 12.03%) and in future designs of similar mechanisms, a rns, but, taken alone. give no informatien as P the relative C3, = 2.0 would seem both reasonable and conservative.
O- tion of the totalload taken by either shear in tM 6" threads
) or by bearing on the shims (P'o r). When compar,1 to re.
. The vafue for Py (min.) is derived from correcting the ts of Series C2, Series C1 allows the qualitative conctusion P; and P'n . to minimum values of Fu allowed by qualit) t shims increase the total load capacity (a conclusion further ance procedures. Thus, corrected P' , = P', x Fsu (actu.
.stantiated by design analysis) but still provide no accurate (min.), and correct P' o r = corrected Cer xFtu (min.) f.
ermination of the interaction between shear and bearing material x A;. In accordance with this procedure, Py(n ds. each test of Series C1 becomes: P'r (min) =(2913.7 x F,o (actual)71091 +(Ce, x Ftu (mir se assume that the ultimate shear strength of the 6" threads been setablished by Series C2 at 2810.6 kips (the f value As an example, for test C1 i ! Pi (min.) at F su = 109 ksi per Table 3.314,then this value be corrected to Fsu (acutal) for the components tested in Py(min.) = (2913.7 x 109/113) + (2.12 x 58 x 3.42) = 32 ies C1 and plugged for P;in Table 3.3-15. Continuing from We then arrive at Py (min.) for the system at E 3 a or :
; first premise, we can then assume that the actual ultimate kips which is 1.26 times the minimum value of 2472.6 De
- ring load (P"br) is the cifference between actual total load acceptance criteria. Numerical values derived above can
) and P',. The above premise is not precise since actual considered accurate as they are based on assumptions o ar ultimate (P';) for Series C1 is not necessarily the same as tionable quantitiative accuracy. This is of no concern a t established for Series C2. Still, there appears tc be no better inch thread with shims is not the critical failure mode roacn based on a limited series of tests and the error in event.
clusions so derived will be small. No real significance, how-r, should be attached to the actual numerical value of the Photos of the components after being tested to failt imate Bearing Strength Constant (Cor) derived from this shown in Fig's. 3.3-32 and 33. Due to the relative magr lysis. of theultimate thread shear force (P',) and 'he ultimate t load on the wilt shims (P'or), it car be assumed that th l :an be seen from Table 3.3-15 that the variance of test bearing failure which is clearly shown in Fig. 3.3-32 d alts, as measured by the coefficient of variation (u), is only occur until af ter thread shear failure. 4%, a small value which gives a relatively high confidence in values for total load (Py). The mean value for Co, of 2.41 j lose to the approximatevalue of 2.57 arrived at in the design lysis of Scction 3.2.6, which gives reasonable confidence in ,.
. 4.. -
w ,
~
*:y =d W
$ 5 5'*** .
ab mi== tss yp ' i i , eT M..? i .' i ,c q; . ? w i . ,. y , O - FIG. 3.3-32: Outer face of Waseer Serial No. 01a Washer Nut FIG. 3.3 33; Inner face of components shown in Fig. 3.3 Serial No. 008 and Solit Shims mer bemg toeced to ultimate as Note that inner face of Washer Washer Nut assembly beers an assemtHy in test C1 1. outer f ace of Solit Shims. 33
10 ANALYSIS OF FAILURE MODE FROM TESTS immary of failure loads for each mode of failure, based on of times which many components withstood actual tende is 8 and C test results, is contained in Table 3.3-16. Failure mate, without failure, gives increased confidence in the of the split shims at either the Bearing Pfate or the more criteria that the end anchorage be stronger than the t cal Composite Washer interface was not determined, but it which it anchors, t bein excessof the 3561 kip maximum load acolied during ten tests and must be due to bearing failurewhichis not a cal mode. 3.3.11
SUMMARY
CONCLUSIONS I hole web shear is shown to be Jightly more critical than The average error of predicted ultimate loads was - ( ir at the 6 inch diameter threads. Failure loads shown in varying from -4.0 to +4.5 maximum error; therefore, it n le 3.316 for both Wire Hole Web Shear and 6" Threads concluded that the design methods used are quite accura h shims) are mean values. give predictable results.
- ompared with the 6 inch threads of the same form, the The coefficient of variation of test results is small, ha 8 inch threads are subjected to a temporary load only, are mean value of 1.974% and varying between a low of 03 iaded in the structural condition, are subject to a maximum a high of 2.98%, indicating that the combined effect of I which is 20% less and have a nominal area which is 57% type production variables and testing variables is insigni ter. This thread is obviousiv not critical and was not tested. therefore it may be concluded that both production me and test procedures were satisfactory, provide uniformity in dimensions (to facilitate inspection, ping, field procedures etc.) it was decided to increase the All test results were over acceptance minimums based o kness of both the Composite Washer and ths Washer from servative basic critera; therefore it may be concluded th 6
4 inches to 4 inches matching the required thickness of the end anchorage hardware as designed and tested will not j her Nut. This increased thickness will provide additional weakest link in the tendon systern. l ngth for botn the Wire Hole Weo Shear and the 6 inch ead Shear failure modes of production components. The , %%u eased strength, computed by linear increase of prototype " l 'rw * ' " iconent test results is shown in Fig. 3.316. The increased i capacity'is not requ. ired for conformance to design or .n%
*1 3 m '
~
rptance criteria and hill'not substantially increase the fail- ** h w* * = > w w load for other (non-critical) modes of failure. TABLE 3.3-16: Summary of failure mode, type of failure and load for both prototype and production end anchorage hardware hould be noted that, by the tim's all test Series were com- on Series 8 and C tests. Summary is fx an anchorage consisti ed, several spucific components had been loaded several fy es to loads greater than the minimum guaranteed ultimate don strength of 2002.8 kips. As part of the overall test l gram, loads > 2002.8 kips were applied five times to Washer rial No. 002, seven times to Washer Nut Serial No. 004, a times to Composite Washer . Serial No. 007, and several es to other components. While this number of cycles cannot gggy gQ ( considered a fatigue test, the applied load is considerably VUUJ // I ier than will ever be applied in the structure, and the number i 1 1 0 34
Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Corrosion Protection 7.12 7 12.1 Describe the concrete cover provisions for reinrorcing steel and prestressing for the dome, base slab and cylinder. Include for comparison the minimum code require-ments. 7.12.2 To what extent will a water proofing compound or membrane be used for the containment base slab and lower cylinder l arest ' 7.12.3 Discuss the corrosion protection to be provided for the containment liner. In particular, what corrosion could develop at the liner cue to concrete structural shrinkage or other cracking at or below the water table? ANSWER 7.12.1 The concrete cover provisions for reinforcing steel and prestressing vill be as shown in Table 7.12.1. The minimum cover specified by ACI-318 is also tabulated for comparison. ! 7.12.2 The Reactor Building vill be surrounded by a circular i \~ retaining vall extending from top of foundation to l elevation 304 ft. The retaining vall is to provide access ' space for stressing and inspecting tendons. At the base i of the retaining vall between the vall and the Reactor l Building, a drainage system vill be provided to keep water ; from accumulating in the access space. Therefore water-proofing of the lover portion of the cylinder is not required. There vill be no water proofing or membrane i used on the base slab. l l 7.12.3 corrosion protection for the containment liner will be as specified in manended Section 5.1.2.8, " corrosion i Protection" and Appendix SE of the PSAR. The-retaining vall as described in the answer to Question 7.12.2 vill eliminate ground water corrosion problems on the liner. 0003 300 l l I 7.12-1 (Revised 11-6-67) i
O T A B L E 7.12 1. MINIMUM COVER LOCATION TYPE OF STEEL TO BE USED BY ACl.318 R EIN FORCING DOME OTHER5 2tiIN. IS IN. PRESTRES$1NG 6 IN. 14 IN. R EINFOR CING CYLIND ER OTH ERS . 2!$ IN. 14IN. ( l PRESTRES$1NG 6 IN. 14 IN, BOTTOM 78 & 145 3 lN. 3 IN. REINFORClHG OTHERS.3IN. 3 IN. BASE MAT TOP 18 & 145 2'.i IN. 2ti & 1% IN. REINFORCING 0 TH ERS . 2!1 IN. Ib IN. l 4 ., c. g;i, e' t: 0003 301 I O T.12-2
Docket 50-289 Supplement No. 1 0::tober 2,1967 QUESTION 8.0 Construction 8.1 General 8.1.1 Indicate where and to what extent ACI 301 standard practice for construction vill be equaled, exceeded, or not followed. 8.1.2 Describe the general construction procedures and sequence that win be used in construction of the containment to include excavation ground water control, base slab construc-tion, liner erection and testing, and concrete construction in cylinder and dome regions. ANSWER 8.1.1 All structural concrete work vill be performed in accordance with " Specifications for Structural Concrete for Buildings," ACI 301-66, except as modified in Section 6.3 of Appendix 53, " Design Program for Reactor Building," and Section 2.1 of Appendix SD, " Quality Control." 8.1.2 The general construction procedures for concrete construction, liner construction'and prestressing are specified in Appendix 5D of the PSAR and in the answer to Question 7.10, 7.12, 8.1.1, 8.2, 8.3, 8.h,and 8.5. The general ccustruction sequence for the Reactor Building vill be as follows:
- a. Immediately after excavation of the rock for the foundation and a devatering system has been established, a lean concrete fin vin be placed to seal the rock l to prevent weathering.
- b. After the foundation is poured, the knuckle and bottom plate of the liner vi n be installed and tested. j Concrete vill then be placed on top of the base of the l liner. '
l
- c. The cylindrical portion of the liner vin be erected !
and individual velds tested prior to the placing of ' reinforcement, tendon conduit,and concrete. Concrete work on the cylinder may proceed prior to completion of the cylindrical portion of the liner.
- d. The dome liner vill be erected and individual velds tested prior to the placing of dome reinforcement, tendon conduit,and concrete.
e. The prestressing tendons will be installed and stressed. l3 O 8.1-1 (Revised 11-6-67) 0003 302 l
Occhet 50-289 Supplement No. 1
-O Cetober 2, 1967 QUESTION Concrete 8.2 8.2.1 Describe the concrete =ixing, placing and curing precedures to be used.
8.2.2 Describe the procedures for bonding between lifts. 8.2.3 Indicate the manner in which concrete lifts will be placed and staggered. 8.2.h Indicate the amount of user check testing of cenent to be accc=plished. ANSWER 8.2.1 A concrete batch plant vill be utiliced which cc= plies in all respects including provisions for storage and precision of
=easurements with " Standard Specifications for Ready-Mixed Concrete," ASTM Chh-6h. As indicated in Appendix SD of the PSAR, the Testing Agency will =aintain an inspector at the batch plant to ensure that the =ix proportions ce= ply with those for the design =ixes with water content =odified as required by =easure=ents to be =nde of content of surface
=oisture on the aggregates. This inspecter vill test period-(
N ically all =ix ingredients and vill ensure that a ticket is provided for each batch docu=enting the ti=e loaded, actual proportions of the =ix, amount of concrete, concrete design strength, destination as to portion of structure, identifi- ~l cation of transit =ixer, and reading of revolution counter , at first addition of water. ! The ready-sixed concrete vill be =ixed and transported in ; accordance with " Specifications for Ready-Mixed Concrete," ' C
, ASTM C9h-65 The =ini=u= amount of =ixing in truck =ixers !
3 loaded to =axi=us capacity will be 70 revolutions of the I
'drmn' or blades after all of the ingredients, including vater, are in the mixer. The maximu= nu=ber of revolutions at
=1xing speed vill be 100. Records vill be =aintained as to the ti=e and reading of the revolution counter when concrete l
is discharged. l Requirements for placing and consolidating concrete vill be as detailed in ACI 301-66. Placing te=peratures vill be 1 li=ited per the require =ents for = ass concrete. Treatment of construction joints vill be as answered hereafter, i Curing of the contain=ent shell vill be in accordance with ACI 301-66 except that the =ethod vill be li=ited to ponding, continucus sprinkling, or =aintaining a centinuously vet I covering. (i) I e 2-t 0003 303 , l 1
8.2.2 Horizcatal and vertical constructicn joints in the Reactor Building shell vill be prepared for receiving the next pour by either sandblasting, air water jet, brush ha==ering, or other =eans tc re=ove all c atings, stains, debris, or other foreign =aterial. The horizontal joints shall be da=pened (but not saturated), then thoroughly covered with a coat of neat ce=ent =crtar of si=ilar proportions to the =ortar in *he concrete. The mortar shall be at least 1/2 in. thick and fresh concrete shall be placed before the =ortar has attained its initial set. The vertical joints will be dampened (but not saturated) before concrete is placed. 8.2.3 Vertical joints vill be placed at the center of each buttress to take advantage of the increased cc=pressive stresses pro-duced by the lapping of the hoop tendor.s. Horizontal joints will be at tt.e same elevation for each lift. Joints vill not be staggered. 8.2.k The ce=ent for the Reacter Building shell vill be supplied by one =anufacturer. The =anufacturer vill sut=it certified copies of =111 test reports showing che=ical cc= position and certifying that the cement cc= plies with the specifications. Whenever possible, the cement vill be frc= the same silo. When a new silo is used the ce=ent will be sa= pled and tested by the testing agency under the direction of the con-struction =anager to ascertain conformance with ASTM C150-6k, / type II. These tests will be audited by Met Ed or its ren-resentative. f - l l \
.i 3.2-2 (Fevised 1-8-c8)
i 1 Docket 50-289 Supplement No. 1 l October 2, 1967 l QUESTION Reinforcing Steel 8.3 8.3.1 Indicate the amount of user check testing of reinforcing steel for strength and ductility to be acccmplished. Include the statistical basis for the,pregram and the basis for reinforcing steel shipment rejection. 8.3.2 Indicate the attention that will be given to cadweld splice quality centrol and include operater qualification and procedural requirements. 8.3.3 Indicate the reinforcing bar velding precedures and quality l centrol to be used in perfor=ing reinforcing bar velds. Include bar preparation, user check testing of reinforcing steel composition, =aximu= per=issible alloy specifications, ; temperature control previsions, radiographic and strength j testing requirements, and the basis for velded splice rejec- { tien and cutout. l I ANSWER 8.3.1 Mill test reports covering mechanical and physical tests will te obtained from the supplier covering each heat of reinforcing steel. User tests vill be perfor=ed on rein- ' f forcing steel to confirm compliance with physical require-ments and verification of mill test results. The frequency , of testing vill be two specimens taken frca each heat of saterial in excess of ten tons and within one heat of material, a series of tests for each twenty-five tons of steel. The user tests vill determine yield and ultimate strength and elongation. If test results do not meet specification requ rements er deviate = ore than 10% fran the mill test results, further testing of that heat of material and an engineering investigation vill be required. It is to be noted that internediate grade reinforcing steel, which is the type =aterial used for this structure, is the lovest strength =aterial ecm=ccly used for con-struction. Furthersore, no relience is placed on special high strength properties and therefore any interchange of higher strength material vould not jecpardice the strength of the structure. Each bar is branded in the deforming process to carry identification as to the =anufacturer, sice, type and yield strength. Because of the identification system and
- the large quantity, the reinforcing vill be kept separated in the fabricator's yard. In addition, when leaded !ct mill shipment, all bars vill be properly separated and ,
tagged with the c;aufacturer's identification number. l O 8.3-1 o#- . o. m, . m 0003 305 i
t 8.3.2 Prior to *he production splicing of reinforcing bars, each operator or crev vill prepare and test a joint for each bar size and position (i.e. vertical, horizontal, side entry, top entry) to be used in.the production verk. To quality, the completed splices vill =eet the following acceptance standards for work =anship:
- 1. Sound, nonporous filler =aterial shall be visible at both ends of the splice sleeve and at the top hole in the center of the sleeve. Filler =aterial is usually recessed 1/k" fro = the end of the sleeve due to the packing =aterial, and is not considered a poor fill.
- 2. Splices which contain slag or porous =etal in the riser, top hole or at the ends of the sleeve shall be rejected.
A single shrinkage bubble present below the riser is not detrimental and should be distinguished frc= general porosity as described above.
- 3. There vill be evidence of filler =aterial between the sleeve and bar for the full 360 degrees; however, the j
splice sleeves need not te exactly concentric or oxially aligned with the bars. l t
- h. Tte strength of the cadveld splice shall be equal to or greater than the specified =intnen ulti= ate tensile strength of the bar.
l A manufacturer's representative, experienced in cadweld splicing of reinforcing bars, vill be present at the job- llh / site at the outset of the work to demonstrate the equipment and techniques used for =aking quality splices. He shall also be present for at least the first 25 production splicss to observe and verify that the equip =ent is being used correctly and that quality splices are being obtained. The following quality control procedures vill be followed to insure acceptable splices: l
- 1. The splice sleeve, powder and =cids shall be stored in a clean dry area with adequate protection frc= the ele =ents to prevent absorption of =oisture.
- 2. Each splice sleeve vill be visually examined i=nediately prior to use to insure the absence of rust and other foreign =ater;al on the inside surface.
l I
- 3. The = olds will be preheated to drive off moisture at 1
*he beginning of each shift when the = olds are cold or when a nev =oid is used.
- h. 3ar ends to be spliced shall be brushed to re=ove all l loose nlli scale, rust, cencrete and other foreign
=aterial. Pricr to brush ng all vater, grease and paint vill be re=oved by heating the bar ends with a torch.
b [.b ; 8 3-2 (Fevised 1-8-o8)
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l 4
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i
- 5. A permanent line vill be =arked frcm the end of each O bar for a reference point to ecnfir= that the bar ends V are prcperly c, entered in the splice sleeve.
- 6. Before the splice sleeve is placed into final position, the tar ends vill be exa=ined to insure that the surface is free frcs =oisture. If scisture is present, the bar ends vill be heated until dry.
- 7. Special attention vill be given to =aintaining the align.
ment of sleeve and guide tube to insure a proper till.
- 8. '4 hen the temperature is below freezing the splice sleeve shall be preheated after all =ateria3s and equipment are in position.
9 All ecmpleted splices shall be visually inspected at both ends of the splice sleeve and at the top hole in the center of the splice. 8.3.3 There vill be no are velding of reinforcing bars. l 0003 307 a 8.3-3
Docket 50-269 Supplement No. 1 Cetober 2, 1967 QUESTION Liner 8.4 8.k.1 Describe the general sequence of liner construction and testing in relationship to the backing structural concrete construction. 8.h.2 Indicate the liner plate dimensional construction controls to be employed for liner plate cut-of-roundness with respect to its influence on liner buckling. 8.h.3 Indicate the extent of user check testing of liner NDT properties, liner thickness, ductility, veldability, etc. 8.h.h Indicate the applicable ASME or API code sections that vill be adhered to in liner construction. 8.h.5 Indicate the procedures and criteria for control of seam veld porosity. 8.h.6 Indicate the requirements for and the control that vill be placed on seam veld ductility. 8.k.7 Describe the quality control procedures for liner angle and d stud velding. 8.h.8 Describe the quality control procedures and standards for field velding of liner plate to include velder qualifica-tions, velding procedures, postveld heat treatment, visual inspection, magnetic particle inspection, liquid penetrant inspection, radiographic inspection, and construction recort Justify, in riotail, the measures selected and in particular the amount or seam veld radiography. I ANSWER 8.h.1 The liner vill be designed so that it can be erected as j a free standing vessel. The liner velds vill be tested I as specified in Section 2.h of Appendix SD, " Quality Control." Deficient veld vill be corrected before concrete is placed adjacent to that portion of the liner. Proposed sequence of liner erection is included in the answer to Questica 8.1.2. 8.h.2 The erecticn tolerances for the liner are listed in Section j 6.0 of Appendix 5E, " Liner P.1. ate Specifications." l
.' .*i O 0003 308 t
- 8. u-:.
8.k.3 The liner plate vill be tested at the fabrication shop to meet those requirements enumerated in Appendix SE of the PSAR. ASIM standard test procedures vill be g employed to ascertain ccmpliance with ASTM Specifica-tions. Certified copies of min test reports describing the enemical and physical properties of the steel vin be submitted to the user for approval. Tests for quali-fying velding procedures and welding vill be performed by the fabricator and monitored by the user. These testa vill provide confirmation on veldability and veld ductility. Testa on nil ductility will be performed for materials used in the penetrations to the extent described in Appendix 5E. The user, or tis authorized representative, vill moniter shop test procedures at the fabrication shop and vill audit all records. The user win not duplicate tests perfomed by the steel supplier and the fabricator. I 8.k.h The applicable codes that vill be adhered to for liner construction are specified in Section 2.1 of Appendix SE, " Liner Plate Specifications." 8.k.5 To control porosity the following steps vill be followed:
- 1. Excessive currents win not be used.
- 2. Each layer of veld metal vill be completely free of slag and flux before the next pass is made.
l
- 3. Molten metal win be puddled long enough to allow entrapped gasses to escape. g 4
Porosity shall meet the required standards of Appendix IV l of Section VIII of the ASME Nuclear Pressure Vessel Code. l 8.k.6 The seam velds in the liner will meet the % quirements of l ASME Boiler and Pressure Vessel Code, Section IX. This requires tests for qualifying veldh:g procedures and velders which provide confirmation on seam veld ductility. l 8.k.7 The liner angle velds vill be tested by liquid penetrant I method at the same frequency and to the same standards for the liner plate. Welding procedures vill be per the ASME Boiler and Pressure Vessel Code, Section IX. 8.k.8 Quality control procedures and standards for field welding vill meet the requirements of Section VIII of the ASME Boiler and Pressure Vessel Code. Inspection of the liner velds vill be as follows:
- 1. 100% visual inspection
- 2. 20% liquid penetrant tes:
0003 '09 l
- 3. 25 radiographic inspection
- k. 100% vacuus box test b[ k((?',
8.k-2 (Revised 1 8)
i t 4 f Construction records vill be maintained at the job site.
^ ' ** '2 *" ' * "'"**" '" **""****" '**
i C~) vill be kept in the possession of Metropolitan Edison Company for the life of the plant, except for those records reouired by code to be maintained by the fabricator. 1 l 1 0003 310 i 4
'O i
l i O 8.k-3 (Revised 1 8 68)
Docket 50-289 O Supplement No. 1 October 2, 1967 QUESTION Prestressing System 8.5 8.5.1 Indicate the basis for the wire /buttonhead factory quality control requirements imposed to ensure production material meeting design requirements and specifications. Where a system other than BBRV is specified, provide these require-ments. Also indicate the extent to which anchorage hardware will (1) be periodically tested for hardness and threading and hole tolerances, and (2) have controlled NDT properties. 8.5.2 Describe the corrosion protection provisions that will be given to wire / strand at the factory, through transportation and in the structure prior to prestressing. ! 8.5.3 Indicate the corrosion protection attention that will be given to the tendor. ducting. 8.5.4 Describe the prestressing sequences, procedures and tendon stress verification that will be employed. 8.5.5 Describe the grouting procedures and controls that assure O proper tendon grouting. ANSWER 8.5.1 As indicated in answer to Question 7.11.1, the BBRV system will be utilized. Tests have been performed using damaged wire and deliberately deforming the buttonhead to establish what maximum imperfections can be colerated without influencing the ultimate capacity of the wire. Wires that have not met the requirements of the specifications will be tested to verify if the imperfections would have influenced the ultimate strength of the wire. Ten percent of the anchorage hardware will be selected at random to check hardness and hole colerances. There will l be no tests for NDT properties. Conclusion not requiring NDT tests are included in the answer to Question 7.10.4. 8.5.2 The prestressing wire will be kept dry and stored in a dry l Place at the fabricator's facility. During fabrication l the wire will be inspected to ensure that the surface is j free from any pitting visible to the naked eye. After j fabrication, each wire will be protected with a coating 1 of NO-0X-lD "CM" and enclosed in a container. The tendons I will be protected from damage during handling and shipping. ; l _ S.5.3 4 The inside surface of the tendon conduit will be coated ! [3 i 'linwich NO-0X-lD "CM" i= mediately after fabrication. The 0003 5 11 l i S.5-1 (Revised 11-6-6 7)
conduit will be capped at all times during construction and will be capped with the permanent enclosure after the tendons are installed and censioned. 8.5.4 The vertical tendons will be stressed first. The stressing operation will start at four positions along the circumferenc< of the cylinder. The sequence of stressing the vertical tendons will be determined after further analysis. After completion of the vertical tendons the dome and wall tendons shall be installed in a sequence so as to minimize stress concentration in the shell. This sequence will be developed after further investigation of shell stresses due to vertical, hoop, and dome tendons. Each tendon will be jacked to eighty percent (807.) of the minimum guaranteed ultimate capacity of the wires. The jacking force will then be reduced to seventy percent (707.) of ultimate capacity when locked off (shi=med in place). The stress-strain curves for the production lots used will be submitted to the Engineer along with the final gage readin. and elongation for each stressed tendon. If the loss of prestress force due to failure of wires or buttonheads exceeds one-half of one percent (1/27.), the engineer will be immediately so advised. Force and strain measurements will be made by measurement of elongation of the prestressing steel after taking up initial slack and comparing ic with the force indicated by the jack- ) dynamometer or pressure gauge. The gauge will indicate the pressure in the jack within plus or minus two percent. Force-jack pressure gauge or dynamometer combinations wt11 be calibrated against known precise standards just before application of prestressing forces begins and all calibration will be so certified prior to use. Pressure gauges and Jacks so calibrated will always be used together. During stressing records will be made of elongations as well as pressures obtained. Jack-dynamometers or gauge combinations will be checked against elongation of the tendons, and the cause of any discrepancy exceeding plus or minus 5 percent of that predicted by calculations (using average load elongation curves) will be corrected and, if caused by differences in load-elongation f rom averages , will be so documented. Calibration of the jack-dynamometer or pressure gauge combina tions will be maintained accurate within the above limits and, if requested by the Purchaser, will be recalibrated, or newly calibrated combinations substitued, during and at the end of the tensiening operations. . ! 8.5.5 The tendons will not be grouted. l 1 0003 H2
! : \
O, i1' ,Ui;!; 8.5-2 (Revised 11-6-67)
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0003 313 O 8.5-3 (Revised 11-6-67) 1
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O I l 8.5-4 (Revised 11-6-67)
l fs (,) Docket 50-289 Supplement No. October 2, 196' - QUESTICN Preoperational Testing 8.6
- 8.6.1 Describe, in detail, the =anner and extent to which all valving v111'be tested for leaktightness both individually and during preoperational integrated leak testing.
ANSWER 8.6.1 Systems having pipe lines penetrating the contain=ent shell vill be hydrostatically tested up to the isolation valve. The pressure of the hydrostatic test vill be 15 ti=es the design pressure of the system of which each line is a part. In all cases this hydrostatic test pressure vill exceed the initial integrated leak rate test pressure of 55 psig. In the event that there is = ore than one isolation valve in l a particular line, the outer isolation valve vill be leak l tested by pressuricing between the two isolation valves to the required hydrostatic pressure.
,s ,.
For these few systems having lines penetrating the contain-ment shell that vill not require a hydrostatic test, special provisions vill be =ade inside containment to
'pressurice a portion of the syste= with air to 55 psis and / perform individual leak rate tests of the valve (s) forming ' s/ the isolation boundary of each system. (The major items in the latter category are: a) the reactor building purge inlet and outlet valves, b) the fuel transfer tubes, and c) service air lines. )
In all the above cases, the isolation valves vill be closed by their norsal operator, as originally adjusted to manufacturer's specifications. In the event that leakage via any valve is unsatisfactory (from the standpoint of the performance of the hydrostatic or special air test, as the case may be), the valve and/or its operator vill be repaired or adjusted as required to insure setisfactory performance before completing the test. The initial integrated leak rate test vill be performed someti=e following the above described series of tests under the following conditions:
- a. all isolation val'm closed by their normal operators
- b. all closed systa=s within the reactor building vill be vented to the buildin6 atmosphere, except for those systams which might be required to be in operation during the test (i.e., a portion of the normal, or amergency, building air coolers in order to provide
) atmosphere te=perature control during the test).
8.6-1 ~0003 315
- c. all closed syste=s outside the reactor building which extend frc= an isolation valve vill be vented to at=cs-h phere as close as possible to the isolatica valve.
- d. there vill be no pressurization of any sleeves penetrating tne contain=ent shell and no fluid blocking of any fluid penetration during the test,
- c. the reactor building vill.be pressuri;es with air to 63 psig, initially, for structural design proof test, and then depressurized to 55 psig to conduct the pre-operatiena leak este test.
The pre-operational integrated leak rate test of the reactor building vill be conducted by the absolute pressure =ethod. 0003 316 O : O i
- t. * ( : '* g g,g
Docket 50-289 Supplement No. 1 October 2, 1967 [} QUESTION In-Service Surveillance S.7 8.7.1 Describe the surveillance capabilities provided by the containment design to facilitate periodic inspection of the steel liner, and monitoring and/or periodic structural testing of the containment. Since leak-rate testing is intended to be performed at reduced pressure, provide an evaluation of the minimum level of such tests that would also serve to verify continued structural integrity. Con-sider in the evaluation structural response and installed surveillance instrementation requirements. 8.7.2 Indicate the extent of long-term structural surveillance to be provided by test samples and in-place instrumentation of the containment. 8.7.3 To what extent will destructible test elements be used to verify the long tenn structural integrity of the prestressing tendons. (i.e. , what will take the place of frequent inspection of the tendons) ANSWER 8.7.1 The tendons will be unbonded, as described in Revised Sections g
-'s 5.1.2.8 " Corrosion Protection" and the answer to Question 8.5, in order to proviC more direct means for future surveillance. The program for future surveillance, to be included as a part of the operating procedures,is under study. Capability will be provided to do the following:
- a. A periodic visual inspection of tendon anchorages will be made to det ermine if wire breakage has occurred.
The tendon is under a stress of approximately 140,000 psi. Should a wire break, the buttonhead will noticeably project beyond the anchorages. The anchorages will be completely inspected immediately before and af ter the structural test. No less than 10 percent of the anchorage randomly selected could be inspected six months after the test, and annually thereafter, for a period of five years. Partial inspection will be conducted af ter this period.
- b. The prestress in the tendon will be reconfirmed by a jacking technique similar to that used in the origina.
stressing o;mration. This phase of the program will provide for obtaining a lift-off reading by using a hydraulic jack to simply lift the anchor head off the shims. This procedure will p (; * , ,' provide a determination of the stress level in the tendon, 4 0003 317
- 8. 7-1 (Ravised 11-6-67)
and will also be used to conform previously predicted stress losses including steel relaxation and concrete creep. No less than 1 percent of tha tendons randomly selected could be so jacked on a biannual basis until losses are stabilized.
- c. Removal of a wire, or wire specimen, will be made from a tendon environment at varying time intervals. This wire specimen would undergo inspection and tensile testing if required fcr evaluation.
A selected number of tendons will include an additional unstressed 1/4 inch diameter wire specimen obtained from a reel represented in the tendon. These specimens which consist of the same material as the tendon and located in the'same environment, could be removed on an annual basis for examination. For example, four specimens could be removed annually.
- d. All tendon anchorages will be accessible, and with the unborded system, the tendons can be removed and replaced.
Tendons will be removed for inspection only if a question exists on the well baing of the tendons, based upon the other surveillance methods. The building is designed to withstand structural proof testing at 115 percent of the de. sign pressure. As previously described in Appendix 5F of the PSAR, the primary memas for / ascertaining that the structural response is satisfactory will be by means of displacement measurements made while the vessel is pneumatically pressurized 6 to 115 percent of the design pressure prior to plant operation. These measurements will be made to ensure that the response is essentially as calculated. For a radial deflection calculated for this con-dirion of 0.22 in., an interument precision of 0.02 in. was deemed to be satisfactory. It can be concluded that these techniques for measuring displacement should be satisfactory down to a pneumatic pressure of approximately 50 percent of design pressure, where the calculated radial displacement would be 0.10 in. The design of the containment vessel provides a capability to pressurize following the structural est up to 115 percent of design pressure. No structural proof eests are contemolated af ter the initial test. 8.7.2 The extent of long-term structural surseillance, including the j use of test specimens,is described in S.7.1. The use of in-i place instrumentation for long-term measurements is .not l contemplated. 8.7.3 Wire specimens will be periodically removed and inspected as described in 8.7.1. l 0003 318 9
,,2
\! ,,,h,b
*O 8.7-2 (Revised 11-6-67)
O l j l DELETED ) s 0003 319 O 8.7-3 (Revised 11-6-67)
Docket 50-289 Supplement No. 1 Cetober 2, 1967 9.0 I_NSTF.t!ErATIcN AND conTRot QUESTION Does the design of your protection system conflict in any way with 9.1 the proposed IEEE Standard for Nuclear Power Plant Protection Syste: If so, please state reascus justifying your position. ANSWER The design of the Three Mile Island Nuclear Station does not confli with the proposed IZEZ Standard for Nuclear Power Plant Protection Systems. 0003 J20 ( 1 l l 9.1-1 ! l i I l
Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Provide a list of all electrical components (cabling included) lo-9.2 cated within containment whose operation during the design basis ac-cident is required for the preper functioning of the engineered safet) features. 9.2.1 Throughout what time intervals must each component operate? 9 2.2 What tests vill be performed to ensure that these ecmponents can, in fact, withstand the postulated accident environment and perform as required? ANSWER Electrical components within the centainment (reactor building) re-quired for proper functioning of the engineered safety features are as follows: hk l ;-
- a. Reactor coolant pressure transmitters.
- b. Electric motor isolation valves.
- c. Reactor building cooling fans.
( d. Instrument cables for pressure instruments.
- e. Power cables and limit switch cables for the valves.
- f. Power cables for the fan motors.
9 2.1 compenent oueration The reactor coolant pressure transmitters need to operate long enough to initiate action of the high pressure injection sys-tem at 1,8CO psig and the low pressure injection system at 200 psig. The valves must operate long enough to isolate the reac-tor building. The emergency cooling fans must operate to cool the reactor building environment following the accident. Cablt associated with each of the above equipment need to operate as 1cng as the equipment is required. All of this equipment vill be designed to perform its required function during the rea m 2 building design basis accident. It . s expected that the final analysia of operating require-r.:.s for engineered safeguards equipment vill indicate that the reactor building emergency cooling fans and associated cabling vill be the only equipment required to operate for an appreciable time in postaccident ambient conditions. O 0003 T21 9.2-1
9.2.2 Testing O The instruments will be of a type which have demonstrated ca-pability to perform under the environments specified. The cables vill have an insulation system proven by test to per-form under the environments specified above. The fan and valve motors vill have c. system of insulation and enclosure which has demonstrated capability to perform under the environments specifiad above. 0003 322 0 ,
.1, m o O
9.2-2
Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION One can postulate several "First faults" in the " trip" bus feeding 9.3 the red release mechanisms whose existence cannot be detected during routine testing (e.g., the connection of the positive side of a d.c. source to the bus). The bus can therefore be disabled by the first detectable fault. For this reason the design does not conform to the sindle failure criterion. Discuss any changes you =ay make to re.-ove this vulnerability. ANS'4ER Several possible alternates are under consideration for the d-c bus feeding the control red drive clutches in crder to give extra assur-ance that the clutches vill be de-energized. These include the fol-loving:
- a. Selit Bus Sections Thir arrangement vould divide the red clutches into two or =cre groups, each on its evn independent pcVer bus. Thus, should ther be a series of faults which held one bus energized, it vould in n vay inhibit trip action by the other group or groups.
- b. Split Bus Sections and Automatic Rundevn (Q- This arrangement would be similar to a above, but vould take cred; for autcmatic rundown of the drives simultanecusly with a trip.
Taking credit for the automatic rundown would allev greater free-dem in the selection of subgroup size and arrangement.
- c. Bus Shunt Switch This arrangement would provide a short circuit across the two cen-ductors of the clutch bus thereby reducing the voltage to zero.
The effect of this arrangement on clutch release ti=e (due to trapped magnetic flux) has not been determined. 0003 323 9.3-1
d O Docket 50-289 Supplement .%.1 October 2, 1967 QUFSTION The protection system is required, under some circumstances, to take 9.h action in response to coolant pump-monitor signals. Is there suffici-margin in the design of the pover/ flow protection system to allow for single failures within the monitors which give rise to false indicati, of pump operation? ANS'M The reactor coolant flow measuremenii at all times acts to protect the reactor from a loss-of-coolant-flov accident. Single failures in the pump nonitors cannot prevent protection of the reactor. 0003 324 O O j 9.h-1
l O cechet 50-289 Supplement No. 1 l October 2, 1967 l QUESTION Identify the sources of pcver to (1) the coils of trip circuit breake 95 in the protection system and (2) the rod drive clutch power supplies. ANSWER The source of power for the coils of " trip" circuit breakers in the reactor protection system is vital instrument power fed from (or throu6h) the reactor protection system cabinets. The control rod drive clutch power is fed from the Station batteries through trip ; circuit breakers and reverse power diodes. Both the trip coils and l the clutch coils de-energize to scram. See Figure 8-1 of the PSAR. j i l i
'O 0003 325 l
1 i i l O 9.5-1
i l Docket 50-289 Supple =ent No. 1 i October 2, 1967 i i QUE.STION F.xplain the purpose of separate dual icgic channels for reactor bui" j 9.6 ing spray pumps and valves. 4 i ANS*4ER The separate logic per=its testing the reacter building spray syster I without actually spraying water by starting the pumps with the valve closed and opening the valves with the pumps shut off. ' i , 1 l J 1 t i i i
.O
- 0003 326 i
1 , i 1 O a e 1 , 9.6-1
, _ . . _ - , . . . _ , - . . _ _ - . ._ . . , .,_m.,,. .
. Docket 50-289 Supplement No. 1 October 2, 1967 Q,UESTION Discuss the significance of the alarm function of the incore instru-97 mentation system. What are the consequences of its failure?
ANSWER The alarm function of the incere instrumentation which is incorporat in the Station computer serves to warn the operator or unusual power distributions developing within the core. Loss of the alarms would require that the operator monitor the incore detector readouts. Normal readouts will be provided on the computer, and a sufficient quantity of readouts vill be provided at an alternate location. Therefore, there ere no consequences to the failure of the alarm functions, except the operator action indicated above. I 0003 327 O 9.7-1
; Oceket 50-289 Supplement No.
Cetober 2, 1967 I QUESTICN Provide your criteria for the design of those sub-systems which l 9.8 control the operatien of Icad-shedding and 1 cad-connecting circuit l breakers under design basis accident conditions. ANS*4ER 1. No single failure, as defined by the four classes of failure, shall result in the icss of more than one safeguards switchgear j bus. (One bus is redundant at every voltage level).
- 2. No single failure shall prevent the connection of a power source to a sufficient number of buses to safely shut devn the reactor.
- 3. No single failure shall result in a total elapsed ti=e, startim at initiation by protection systems and ending at full speed of the high pressure injection and decay heat pu=ps, of =cre than 25 seconds.
- h. The subayatem for each safeguards h160 volt bus will be test-able both for load shedding and load connecting.
( 0003 (28 O 9.S-1
Dccket 50-269 Supplement No. 1 October 2, 1967 QUESTION To what extent are your engineered safety feature systems vulner-99 able to an accidental reversal of a three-phase voltage supply? What precautions will be taken to prevent such an occurrence? ANSWER It is not credible that the three phase voltage supply from the unit auxiliary or from the safeguards transformer will reverse accidentally or that an accident will occur while these sources are reversed due to inslvertent connection, since discovery would occur before, plant startup. Inadvertent reversal of two phases at a piece of equipment follow-ing maintenance or preceding startup, if not discovered, would
- result in reverse rotation of a motor. This contingency will be considered a single failure and its effects guarded against in the design of safeguards systems.
Reversal of a diesel-generator would reverse the phase rotation i of one group of safeguard switchgear buses. This again will be considered as a single failure; and redundant equipment, such as valves and pump motors, fed from different power sources so as to prevent any suon phase reversal from affecting safe shutdown. Phase reversals due to inadvertent misconnection will be found {( ) by tests, such as jogging motors, after any di.> srbance of connections. Reverse rotation of diesels wiim >< prevented by specifying air starting equipment capable of retsting in only one direction. 0003 .s29 1 l I O 9 9-1 i
Docket 50-289 Supplement No. 1 October 2,1%7 QUESTION Describe the proposed 250/125 vde system. In addition, provide the 9 10 fonowing: 9 10.1 The capacity at each battery and battery charger. 9 10.2 The emergency loads in each d-c bus section. Will each battery be capable of carrying full emergency load? 9 10 3 Tests to be performed on the batteries and the test frequencies. ANSWER The arrangement and number of batteries, chargers and d-c distri-bution panelboards are as shown on the acccmpanyinE Figure, 910-1, with the facility for feeding the output of spare battery chargers to either of the =ain battery distribution panels and provision for crocs connecting these *.wo distribution panels. The entire system, comprising both batteries, will satisfy single failure criteria. Each battery will hqve sufficient capacity to feed the connected essential load for 2 hours continuously and perform 3 complete cycles of safeguard breaker closures and subsequent tripping. 9 10.1 Each battery has been tentatively sised as follows, presum-t ing 120 cell batteries and discharge to 1.75 volts per can: Ampere Hours Amperes , l 8 Hours 3 Hours 1 Hcur 1 Minute 900 702 702 1014 Each battery charger has been tentatively sised at 75 ! amperes. This would allow a battery to be fully recharged l
?5i in 8 hours, presuming one battery has failed and 4 battery U;checers are connected to the remaining battery. The normal d-c load is also accounted for in this figure. All of the foregoing figures will be refined for a specific manufacture of battery and number of cells, total final voltage across the battery remaining at 210 volts after discharge.
9 10.2 The anticipated loads on each battery bus are as follows: D.C. Distribution panel 1A Emergency 011 Pu=p 1 ' l Boiler Feed Pump Emergency 011 Pump la Reactor Coolant Pump 1A Reactor Coolant Pump 13 0003 330 0 9 10-1
D.C. Distribution Panel 1C D.C. Distribution Panel lE h Emergency Lighting Feeder 1A Screen Ecuse D.C. Distribution Panel D.C. Distribution Panel 13 Emergency Seal Oil Pu=p 1 Boiler Feed Pump Emergency 011 Pu=p 13 Reactor Coolant Pu=p 1C Reactor Coolant Pump LD D.C. Distribution Panel 1D D.C. Distribution Panel 1F E=ergency Lighting Feeder 13 D.C. Distribution Panel 1C 6900 V. Switchgear lA hl60 V. Switchgear lA 4160 V. Switchgear 1C 480 V. Switchgear LA 480 V. Switchgear 1C Water Treating Control Panel 1 D.C. Distribution Panel 1D 6900 V. Switchgear 13 4160 V. Switchgear 13 480 V. Switchgear 13 / 480 V. Switchgear 1D Condenrate De=inerali::er Panel 1 D.C. Distribution Panel 1.E 4160 V. Switchgear ID 480 V. Switchgear lE
})}
Static. Inverter lA Static Inverter 13 Safeguards Protection Bus lA Reactor Control Rod Clutch Bus lA Relay Rack 1A D.C. Distribution Panel 1F 4160 V. Switchgear lE ',1,[ [ '
.,.it..,
. MO V. Switchgear IF
/
9 10-2 (Pevised 12-22-67)
O Static Inverter 1C Static Invertir ID Safeguards Protection Bus 13 Reactor Control Rod Clutch Bus 13 Relay Rack 13 Each battery will be capable of carrying the entire emergency load, shedding only the larger motors after their function is fulfilled, as discussed in 4.10. 9 10 3 Tests to be perfor=ed on the batteries will be as follows:
- a. Individual cell specific gravity and voltage will be checked once a month, using a least 2 pilot cells per battery.
- b. Water level will be visually checked at the same time.
- c. Once a yeer all inter-cell and inter-tier connectors will be checked for corrosion and tightness.
- d. Voltmeters and ammeters will be provided in the contre room, as well as battery ground indication, so that one battery may be checked against the other at glance
- e. Upon installation and periodically thereafter a dische rate for 1/2 hr will be imposed upon the battery. A check of battery voltage before and after the test wil prove whether the battery discharge characteristics remain on the predicted curve.
0003 332 O 9 10-3 (Revised 12-22-67)
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Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION What is the rating of each diesel generator unit? Assuming a 9 11 failed generator, what margin exists in terms of power requiremet for minimum engineered safety feature operation? 3 ANSWER The diesel generators are tentatively sized at 2,850 kv. Each diesel generator vill be capable of assuming the entire safe-6uards load connected to its respective kl60 volt and h80 volt buses without exceeding the nameplate rating. (See Question 18.5 for a more complete answer.) In addition, sufficient capacity will be available to feed certain turbine plant loads, such as bearing oil pumps. It should be noted that the engi-neered safeguards equipnent to be carried by one diesel geners-tor, assuming the other to have failed, constittmes 100% capacity 03 334 I l l l l 1 l 9 11-1 (Revised 12-22-67) i
Docket 50-289 O Supplement No. 1 Cetober 2, 1967 QUESTION Discuss the independence of the diesel generator uni'ts with 9 12 respect to (1) physical separation, (2) starting systems, (3) lubrication systems, (4) fuel supplies, fuel pumps, (5) cooling systems, (6) control signals, and (7) fire protection. ANSWER (1) The two diesel generators will be located in adjacent rooms. separated by Class I, fireproof walls. There win be no interconnecting piping, wiring, or ventilating ducts througr these walls. (2) Each diesel win be air started by its own independent system which win include tanks, piping, air compressor and valving. ' (3) Each diesel win have its own independent lubricatior syster (4) Each diesel win be provided with a fuel tank with sufficien capacity for approximately two hours full load operation. Level in each tank win be maintained automatically by a fuel transfer pump supplied with each unit. (5) The diesel generators will be air cooled using two indep cdent cooling systems. (6) Control signals to the diesel generators vill be in separate cables, routed through separate underground con-duits. This separation will be maintained at the switchgear and at control boards. (7) Each diesel generator will be protected by a fire fog system. A rate of temperature rise detector, which will bring up an alarm in the control room, and a deluge valve will be provided for each of these systems. The electricany operated deluge valves will be opened by means of separate pushbuttons in the control room. 0003 335 m V 9 12-1 (Revised 12-22-67)
Docket 50-289 O' Supplenent No. 1 October 2, 1967 QUESTION Assuming a total loss of external power coincident with a design 9 13 basis accident, provide a failure analysis to show that no single failure can prevent the actuation of sufficient engineered safety feature devices. Postulated failures should include, but not necessarily be limited to: (1) short circuit, (2) open circuit, (3) failed diesel generator, (k) failed engineered safety feature device, (5) malfunctioning circuit breaker (lead-shedding or
- connecting), (6) loss of one battery, and (7) faulted undervoltag monitor (at emergency bus).
ANUWER Postulating Loss of off site Power Combined With a Design Basis Accident Single Failure Analysis for the Engineered Safety System Component Malfunction Comments & Consecuences
- 1. Engineered Short Circuit a. (With the loss of externa Safeguards power each 4160 v bus wil 4160 Volt Bus be isolated and all break ID or IE tripped.) The correspond kCO volt safeguards svite gear and motor control ce vill be lost, however, th are redundant valves and auxiliaries connected to f
the remaining switchgear ( sections and motor contro centers for safe shutdown
- b. The a-c source for invent fed from one of the two batteries will be lost, a will be one pair of batte chargers. That battery w assume the load of two in verters and half of plant e, s d-c load.
(see answer to i' '
'J.lfi question 9 10.) The fail of an a-c feed to any inv-or battery charger is sn-nunciated, and each batte.
,' is si ed for full emergen icad.
- 2. Engineered Short Circuit a. The faulted bus will be Safeguards isolated by protective LSO Volt sus circuit breaker action so IE or IF that no kl6C volt auxilia.
l will be loat. O 0003 336 9 13 1 (Revised 12-22-67) l l i
- b. One 480 volt safeguards switchgear section and control center will be lost. The corresponding battery will be affected as in 13. Sufficient redundant auxiliaries will be fed from the remaining switchgear and motor control centers.
3 Motor Control Short circuit a. The faulted motor control Center Bus center will be isolated by protective circuit breaker action. The a-c feeds to 2 inverters and 2 battery chargers will be 1 cat and the corresponding battery will assume the load. No protective ' function will be lost and sufficiert redundant valves and auxiliaries will be operative for aafe shut-down.
- 4. Any bus or Open Circuit The consequences could, at feeder the most, result in the loss of one 4160 volt bus and the correspondin6 480 volt j bus and =otor control center.
Consequences to one battery would be as in lb. Suf-ficient redundant valves and auxiliaries would remain in service for safe shutdown. _ 0003 337
- 6. Diesel gen- Failure The consequences would be erator A or B identical to those of 1 above.
Sufficient redundant valves and auxiliaries would re=ain in service for safe shutdosr fed from the re=aining diesel generator. np: . ;.e 9 13-2 (Revised 12-22-67)
i O . DELETED
- 8. Any engineered Failure a. Separate undervoltage det.
safety feature tion, relaying and logic 4 device provided for each diesel generator and the corres-ponding 14160 volt and 1480 volt switchgear buses. The
- naximum result of a failu:
of any component would be that of loss of one diese: generator system. Sufficie auxiliaries would remain service to safely shutdowi the reactor. 9 Any load shed- Malfunction a.
; f ding or con- ; f necting circuit breaker DELETED l
Failure-to-trip the incomir4 circuit breal from one of the two auxil; transformers will result lockout of the correspond: , diesel generator,
- b. Malfunction of a load shet ding circuit breaker woulc result in the inclusion o:
that load in the first bic i of equipment started by tl l corresponding diesel gene: ator. Since each diesel generator is sized to car: the full safeguard connec-load of the bus to which is connected, the effect t the :nalfunction would be - O 0003 338 9.13-3 (Revised 12-22-67) 0.5 ? ' !.(
yield a lower voltage than usual for about one second g until the voltage regulator brought the terminal voltage back to nor=al. At the worst, the affected diesel-generator might be overloaded The other diesel gen-erator system would not be affected.
- c. Failure-to-close of one diesel-Benerator breaker would result in the loss of a redundant system, as described in 1 above.
- 10. Battery A Lc'ta of one a. The control power source or Battery of hl60 volt safeguards B bus #1D and k80 volt safe-guards bus #1E vould be lost. The remaining, redundant safeguards buses vould be unaffected.
- b. The d-c feeds to 2 inverters would be lost. If the loss h of the battery occurred durin( j the first 10 seconds after initiation of diesel engine starting, scme instrument circuits would be: lost during this period but would be recovered when the a-c buses reenergize at the end of 10 seconds. Reactor protection and Nuclear In-strumentation a-c circuits would initiate a scram which would have no effect since, under the conditions post-ulated scram would already have taken place.
l
- 11. Emergency bus Faulted ~ a. There are 3 voltage monitors undervoltage en each hl60 Volt cafeguards monitor on bus. Disagreement between l
Bus ID and voltage monitors vill be IE annunciated. Failure of one of the safeguards bus voltage scnitors to operato correctly will be taken care of as g g73 ;. , f 11 "*: W
/
9 13 15 (Revised 12-22-67) l 1 0003 339
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DELETED l
- 1. When 2 of the 3 voltag monitors on a bus indi l cate voltage failure, the emergency bus in question vill be clear j and the diesel generat !
circuit breaker closed ! when it reaches proper voltage and frequency. When 2 out of 3 of the ; monitors on the bus l indicate full voltage, l starting of safeguards l equipment will begin. ; 1 2- Logic for the remainin l safeguards bus will ac independently. O 0003 140 DELETED i i , 9 13-5 (P.evised 12-22-67) l i
i
- Docket 50-289 Supple =ent No. 1 Cetober 2, 1967 QUESTICN Provide justification for not utilizing radiation =enitoring system 10.1 signals to initiate isolation or interlock functions in the plant liquid and gas discharge lines.
ANSWIB Interlock functions will be provided using the Radiation Monitoring System signals as follows:
- a. The ra?iation detector in the Liquid Waste Discharge Line vill initiate closing of the liquid vaste discharge valve,
- b. The radiation detector in the Plant Liquid Effluent Line vill serve as backup to ites (a) and vill initiate closing of the liquid vaste discharge valve.
- c. The radio-gas detector in the Auxiliary Building Ventilation Duct will initiate closing of the discharge damper and tripping of the supply ran motor.
- d. The radio-gas detector in the Fuel Handling Building Ventilation Duct will initiate cicsing of the discharge damper and tripping of the supply fan motor.
f
- e. The radio-gas detector in the Waste Gas Decay Tank discharge
~/ line vill initiate closure of the gas discharge valve.
- f. The radio-gas detector in the Auxiliary and Fuel Handling Euildings Exhaust Duct will act as hackup to ite=s (c), (d),
and (e) and vill initiate closure or all da=pera, tripping , of all supply and exhaust fan =otors, and closure of the ! Waste Gas Decay Tank discharge valve.
- g. The radio-gas detecter in the Reactor Building Purge Duct vill initiate closure of the Reactor Building Purge Valves I and tripping of the Reactor Building Purge and Supply fan 1
- otors.
The interlocks will be tested periodically to deter =ine their operability. 00h03 341 l O 10.1-1 t
Docket 50-289 O Supplement No. 1 October 2, 1967 QUESTION Describe the system which automatically drains vastes to the vaste 10.2 batch tank when they are sufficiently concentrated, the means used to determine concentration of vastes and the consequences of failure of the automatic systa=. ANSWER The vaste evaporator package contemplated for Three Mile Island Station operates on a batch-type process. The basic units in the package consist of the evaporator, the process batch tank, the evaporator feed pumps and the distillate equipment. Rav vaste is pumped frcm the batch tank into the evaporator using the evaporator feed pumps. Part of this flow is diverted before entering the evaporator and passes through an eductor below the evaporator shell. The division of flow enables a constant amount of liquid to be extracted frca the evaporator by the eductor, and this stream passes back into the batch tank to complete a closed cycle. A given batch of rav waste is pumped around the cycle continuously until the desired concentration is reached. The concentration of the vastes is determined using a nitrogen gas-bubbler system. Two dip tubes are Lunersed at different levels in the liquid in the batch tank and are supplied with a constant nitrogen supply of at least 1 scfh through a filter and purge meter and allow a steady ( flow of bubbles to escape into the. batch tank solution. As the solution becomes more concentrated, the rate of escape is reduced due to the increased pressure of the column of concentrated solution Density is measured by the difference in pressure between the two dip tubes and is indicated directly on a panel gage via a differenti pressure transmitter. When the desired density level is obtained in the batch tank, vacuum is broken in the evaporator and the feed pumps are shut down. With the batch tank physically located teneath the evaporator, the concentrated vastes remaining in the evaporator vill drain by gravit; into the batch tank. However, in the event that the final design of the evaporator package and plant layout does not allow placement of the batch tank below the evaporator, the concentrated vastes vill be extracted from the evaporator by valving off the inlet line to the evaporator and allowing the evaporator feed pumps to continue pumping solution through the eductor thus drawing off the remaining liquid in the evaporator. The final design of the plant will incorporate either the autcmatic gravity drain feature or the operati controlled drain feature. In either event, failure to drain the evaporator of concentrated vastes will not be a safety problem since adequate shielding and rsdiation control measures will be provided to prevent exposures in excess of 10 CFR 20 limits even when a slug of contaminated vaste is residing in the evaporator. G' 0003 342 10.2-1
l ! i I i Docket 50-289 4 Supplement No. 1 October 2, 1967 IGN h is the release rate used in the vaste gas decay tank failure I j ANS'M The release rate assumed in the vaste gas tank failure analysis is 10 cubic feet per minute. 1 I i i s ! 0003 343
,0 4
k i j )O 10.3-1
O Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Provide a Radiation Monitoring System sche =atic similar to the 10.4 format of Figure 7-2 which indicates location, equipment type, power sourees, and interlock functions. The schematic should show the relationship of the area, vaste and gas disposal, ventilation, and site monitoring systems. The schematic should be accompanied by adequate description. ANSVER The Radiation Monitoring System vill consist of the Area Gamma Monitoring, Atmospheric Monitoring,and the Liquid Monitoring systems as shown on Figures 10.h-1 thru 10.k-5 All monitoring signals will be indicated and recorded in the control room. The Radiation Monitoring System will receive pove **~ *he battery backend inverter fed vital instrument buses,and each channel vill have a loss of power and channel failure alarm in addition to a high radiation alarm. The Radiation Monitoring System detector locations and sensitivities vill be as follows: A. Area Gamma Monitcring Detector Range Location O 1. .1 to 100 = rem /hr Control Room Radiochemical Laboratory Relay Rocm Machine Shop Auxiliary Building
- 2. .1 to 10,000 Sample Room mres/hr Decay Heat Pu=p Area j Reactor Coolant Waste Evap.
Pump Area , (y .,* ,'- Make-up Tank Area l
'+*,' Inter =ediate Ccoling Pump l Area l haar Fuel Handling Pool l 1
Reactor Bui2 g
- 3. .1 to 10,000 Near Fuel Handling Bridges mres/hr Near Personnel Access Hatch Incore Instrumentation Area l
1 rem /nr to 1,000,000 rem /hr Reactor Building Dome O 0003 344 ! l 10.k-1 (Revised 12-8-67)
B. At=cspheric Monitoring The recorders vill be located in the control room; the detectors vill be located as follows: Location Tyre of Measurement Tree of Monite Reactor Building P.G,I Fixed Purge Duct Auxiliary and Fuel P ,G ,I Fixed Handling Building Exhaust Duct Control Room Ven- P.G.I Fixed t11ation Duct Fuel Handling P,G,I Fixed Building Ventilation Duct Auxiliary Building P,G,I Fixed Ventilation Duct Reactor Building Air P,G,I Fixed Sample Line Site Monitors (2)
@ 2000' Site Boundary P,G,I Fixed Sample Rocm P G.I Fixed Waste Gas Decey Tank G. Fixed Discharge Condenser Vacuum Pump G. Fixed Exhaust ~j Radiochemical Labor- P,G,I Movable atory Spent Fuel Area P.G.I Movable C. Liquid Monitoring Primary Coolant Letdevn Intermediate Cooling Water Nuclear Service Closed Cooling Water Spent Fuel Cooling Water Plant Liquid Effluent Line Liquid Waste Discharge (Prior to Dilution) 0003 T 5 -
l 9 e .7. ., a; - - I 1 10.h-2 (
1 CONTROL ROGM i RADIOCHEMICAL LA8.
.1 10 2 mr/h 1 RELAY ROOM i
U i ---- 1 MACHINE 5NOP d. i SAMPLE ROOM ___ , AUXILIARY BUILDING DEC. 3 RECORDER 5 NEAT PUMP __ ,, AUXILIARY BUILDING R.C. EVAPORATOR FEED PUMP
~
~~ ~~
AUXILLARY BUILDING MAKI (O
.1 10d mr/h TANK ARE A
' --~~
AUXill ARY BUILDING INTE COOLING PUMP AREA l AUDl8LE ALARM ' ~~~~ AUXiLI ARY BUILDING FUE NANDLING POOL AREA
' ~~~~ '
REACTOR BUILDING FUEL l NANDLING BRIDGE ARE.A LEGEND , EA00W --~~ 1 REACTOR BUILDlHG HEW F NANDLING BRIDGE AREA Q SENSOR (ION CHAMBER) MlGN & Low LEVEL ALARM LAMP i ~-~~ REACTOR BUILDING NEAR PERSONNEL ACCESS NATC
<r
~~~~
REACTOR BUILDING IN-C0f
' INSTRUMENT TERM. AREA 1-100 Rih 1
REACTOR BUILDING DOME O . . . 0003 346 AREA GAMM A MOHlTORINi
, FIGUR E 10.41 AMEND. 4 (12 6-67)
TT FIXED A AUXtLIARY & FUEL HANDLING
~ ~ ~ P O I SUILDING EXHAUST DUCT h O Tr B
, ____ p g , CONTROL ROOM VENTILATION DUCT O
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0003 347
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- SU' M 1A IB IC 1 AREA GAMMA CONTROL ROOM X AREA GAMMA AUX. BUILDING DECAY HEAT PUMP X AREA GAMMA R. B. NEAR PERS. ACCESS HATCH X AREA GAMMA RELAY ROOM X AREA GAMMA WASTE EVAP. FEED. PUMP X AREA GAMMA R.B. IN CORE INST. AREA X AREA GAMMA NEAR R. B. FUEL HDLG. BRIDGE X AREA GAMMA NEAR R.B. FUEL HDLG. BRIDGE X AREA G AMMA MAKE.UP TANK AREA X AREA GAMMA INTER. COOLING PUMP X AREA GAMMA RADIOCHEMICAL LAB. X AREA GAMMA MACHINE SHOP )
AREA GAMMA SAMPLE ROOM } AREA GAMMA NEAR AUX BUILDING FUEL HDLG. POOL AREA GAMMA REACTOR BUILDING DOME : ATMOSPHERIC RADI ATION CONDENSER VAC. PUMP EXHAUST X ATMOSPHERIC RADI ATION AUX. BUILDING VENT DUCT X ATMOSPHERIC RAD? ATION SPENT FUEL AREA X ATMOSPHERIC RADI ATION REACTOR BUILDING PURGE DUCT X ATMOSPHERIC RACI ATION SITE MONITOR X ATMOSPHERIC RADI ATION AUX. & FUEL HDLG. VENT DUCT X ATMOSPHERIC RADI ATION SAMPLE ROOM X ATMOSPHERIC RADI ATION R.B. AIR SAMPLE LINE X ATMOSPHERIC RADIATION FUEL HDLG. VENT DUCT X ATMOSPHERIC RADI ATION SITE MONITOR X ATMOSPHERIC RADI ATION RACIOCHEMICAL LAB. ) ATMOSPHERIC RADI ATION GAS DECAY TANK DISCHARGE ) ATMOSPHERIC RADIATION CONTROL ROOM VENT DUCT ) LIQUID MONITOR PR! MARY COOLANT LETCOWN X LIQUID MONITOR NUC. SER. CLOSED COOLING WATER X LICulD MONITGR PLANT EFFLUENT LINE X LIQUID MONITOR SPENT FUEL COOLING WATER X LICUID MONITOR INTER. COOLING WATER ) LICUl0' MONI TOR LIQUID WASTE DISCHARGE > V) ! 0003 350 t R ADI ATION MONITORING SYST POWER SUPPLY SOURCE $ FIGUR E 10.4 5 AMENO. 4 (12 8 67) l
Docket 50-289 Supplement No. 1 October 2, 1967 QUESTION Discuss how the design basis accident releases and other 10.5 accidental releases relate to the radiatics monitoring system design including range, sensitivity and detector location. 1 ANSWER The answer to question 10.h discusses the rsdiation monitoring i system and cutlines the number, type, range, location, and sensitivity of an radiation monitors. The reactor building dme monitor is intended to indicate the radiation level inside the building following the design basis accident and has a range of 1 to 106 R/hr. The total curies I released into the reactor building following the design basis accident is estimated to be 17 x 106 curies corresponding to the release of the total gap activity. It is estimated that this activity would produce a radiation level en the order of 2h0,000 R/hr which can easily be read by the monitor. The concentration of I-131 at the 2000 ft exclusion distance is estimated to be about 5 x 10-7 uc/ce during the 2 hour period following the design basis accident. The site monitors, with a sensitivity of 10-11 Jac/cc for iodine, will record and alarm ; thisactivgtylevel. The site gas monitors with a s'ensitivity of 2 x 10- Jac/cc vin record and alarm an estimated Xe-133 concentration of 3.0 x lo 3c/cc. For particulates, assuming all Cs-137 released during the accident vin be available for
/ leakage from the reactor building, the conce tration of cs-137 at the exclusion distance is about 1.1 x 10 uc/ce. With a sensitivity of 10-11 Jac/cc the particulate monitor win record and alarm this activity level. An site monitors would thus be expected to give indication of the radiation levels at the site boundary following the design basis accident.
The control room and auxiliary building gamma monitors are also anticipated to remain on-scale following the design basis accident. For the control room, the shielding provided by the reactor builaing and control room valls and roof and the design of the control roca ventilating system vill prevent expcsures
- in the control reem exceeding 3 rem over 90 days follow:ng an MHA. The radiation level in the control room over the first 2 hours folleving an MHA is estimated to be less than 100 mR/hr.
The consequences of the cesign basis accident are less severe than the MHA bence the centrol room monitor with a range up to 100 mR/hr shculd provide level indication throughout the accident.
,17.e auxiliary building monitors, with a range up to 10R/hr, are halsoe'stimatedtoprovidelevelindicaticaintheauxiliary building following the accident.
In addition to the design basis accident, the effects of other accidents, dise assed in Section ik of the PSAR, upon the radiation monitoring system have been investigated. These include steam i generator tube leakage, stesa line failure, loss of electric 1 power, rod ejection accident, and a fuel handling accident.
! 0003 351 10.5-1 (Revised 12-6-67 )
)
Docket 50-289 Supplement No. 1 October 2, 1967 l k 1 A 1 gp1 steam generator tube leak releases 290 p:/cI of Xe-133 into the secondary system. With a steam flow of 10 lb/hr, this gas could be diluted by as large a factor as 1.7 x 10-6, ! This dilution factor vould produce a Xe-133 concentration of 1 l 5 x 10 4 pc/:: available for release from the condenser. The condenser exhaust monitor, with a sensitivity of 10-6 pe/ce, will easily record and alarm this activity. For the steam I line failure, with steam generator tube leakage, 1k.1.2.9.2 of the PSAR indicatss a thyroid dose of 0.88 rem at the site boundary yorresponding to an I-131 concentration of about 1.6 x 10- pc/cc. The site iodine monitor, vita a sensitivity l of 10-11 pc/ce, will record and alarm this activity level. Loss of electric power, discussed in 14.1.2.8.1, analyzes steam relief to the environment concurrent with steam generator leakage and gives a concentration of 0.05 MPO for iodine at l the site boundary. This is estimated to correspond to a con- ! centration of 5 x 10-12 p:/cc of iodine at the site boundary during the 2 minute period of steam relief, which is well i vithin 10 CFR 20 limits. 1 1 Section 14.2.2.2.3 of the PSAR discusses the rod ejection with loss of coolant accident. 177,000 curies of I-131 are released to the reactor building producing a radiation level of 6500 R/hr. The addttional amcunt of Xe-133 released is estimated l to be 5 3 x 100 curies which vill produce about 13,000 R/hr. Hence the overall radiation level inside the reactor building vill be on the order of 20,000 R/hr which vill be recorded and ! alarmed by the reactor building dome monitor. The dose to the thyroid at the site boundary is 2.65 rem ove pondingtoanI-131concentrationof2xlog30dayscorres-pc/cc which vill be recorded and alarmed by the site iodine monitors. The fuel handling accident is analyzed in Section ik.2.2.1.2 of the PSAR. The fuel handling ventilation system monitor and the exhaust duct monitor vill normally provide automatic iso-lation of the fuel handling building for the sudden release ( of 28.k curies of iodine and 2.79 x 104 euries cf noble gases. However, in the event that all this activity is released to the environment the resulting concentrations at the site boundary are estimated to be 2 x 10-7 pc/ec for iodine and 2 x 10-4 pc/cc for the acble gases. The site monitors will reecrd and alarm these concentratiens. i l 0003 .52 l i l 9
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- 3 '9 It 10.5-2 sae.ised 12-3-67)
Decket 50-289 Supple =ent No. 1 October 2, 1967 QUFSTICN What actions are initiated upon receipt of a high radiation 10.6 alar =. ANS'ER Upon receipt of a high radiatica alert ala = tce following action vill be taken:
- a. The channel that is alar =ing vill be checked to deter =ine that the alar = is being caused by radiation and not due to equip =ent
=alfunction.
- b. For Area Gam =a alar =s, the area vill be =cnitored with portable instru=ents to determine the =re=/hr reading. If a high level is found, the area vill be cordened off and properly posted as to dose rate and the necessity of fil= badges, etc.
- c. For high At=cspheric and liquid radiation alar =s, autc=atic shutdown of the affected syste= is provided to insure isolation of the syste=. The amount of system activity vill then be checked by sa=pling,and =easures vill be taken to locate and isolate the source.
O 0003 353 9
/
O'%s 10.6-1
Oceket 50-289 Supple =ent No. 1 Cetober 2, 1967 AUESTICN Discuss the reasca for not =enitcring the turbine building exhaust. ANS'ER Ventilation of the turbine building is accc=plished by drawing air into the building near ground elevation through icuvers and exhausting it by use of fans in the turbine building rcof. Unit heaters a.e provided for te=pering inec=ing air as required. Off gases frc= the main and auxiliary condensers are discharged via a vacuu: pump to the eff gas vent. The condenser off gas vent leads frc= the vacuu pu=p to the building exterior and ter=inates at an elevation near the tcp of the building. As stated in the revised Section 11 of the PSAR, the condenser ) vacuus pu=p discharge vill be =cnitored to detect pri=ary to secondary syste= leakage. This =ccitor will detect carryover of radicactive gases, and an alar = will alert the station operater. Since this is the only source of gaseous activity in the turbine building, it is not necessary to =enitor the turbine building exhaust. ) i 1 l i. r 0003 354 1 1 i 4 l l l 1 O 1 1 i 1
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E s Docket 50-289 ( Supplement No. 1 October 2, 1967 QUESTION Describe the operation of the control reem heating and ventilation 10.8 systems during normal and emergency operations. ANS'4ER A flow diagram of the control rocm heating and ventilation system is shown on Figure 9-12. Equipment redundancy is provided. During normal operation, the system vill function in a conventional
=anner firut mixing a =ini=u= a=ount of outside air with recirculated air. The air mixture is then du'at filtered and heated or cooled as required to maintain the space temperature. The nuclear services cooling water system, which will remain in operatica during an emergency, serves the ec= denser of the air conditioning unit.
During emergency conditions, the use of cutside air vill be disecc-tinued by means of automatic dampers. Recirculated air vill bypass the r.ormal dust filter and pass through the emergency dust filter, HEPA filter, and C::arcoal filter before being heated or cooled and delivered to the conditiened space. 1 l \' 0003 355 l O 10.5-1
N s,) Docket 50-269 Supplement No. 2 November 6, 1967 QUESTION THERMAL SHOCK ON REACTOR VISSEL 11.0 With regard to thermal shock on the primary system co=ponents, induc by operation of the e=ergency core ecoling systen (ECCS), provide t following: 11.1 Details of an analysis which indicates that the reactor vessel can accommodate without failure the rapid temperature change at the end of its design life. The analysis should consider both the ductile yielding and the brittle fracture modes of failure, and should in-clude the following specific infor=ation: (See these further speci items following the answer to Question 11.1. ) ANSWER The state of stress in the reactor vessel during the loss-of-coolan accident has been evaluated for an initial vessel te=perature of 6C The inside of the vessel vall is rapidly subjected to 90 F injectic water at the maximum flow rate obtainable. The results of this ana ysis show that the integrity of the vessel is not violated. The assumed modes of failure are ductile yielding and brittle fract The modes of failure are considered separately as follows: ( a. Ductile Yielding The criterion for this = ode of failure is that there shall be n gross yielding across the vessel vall using the minimum specifi yield strength in the ASME Code, Section III. The analysis con sidered the =aximum co=bined thermal and pressure stresses thro the vessel vall thickness as a function of time during the safe injection. Comparison af calculated stresses to the =aterial yield stress indiceted that local yielding =ay occur in the inn 1k.7 per cent of the vessel vall thickness.
- b. Brittle Fracture Since the reactor vessel vall in the core region is subjected t neutron flux resulting in embrittlement of tne steel, this area was analy::ed from both a transition te. perature and a fracture mechanics approach. The results of the two methods of analysis compare faverably and show that pressure vessel integrity is no lost.
The criterion used in the transitien te=rersture analysis is th a crack cannot propagate beyond any point where the applied str is below when thethe threshold etress stress f9r 121agk initiation (5-8 ksi) er is co=pressive.tl)c This approach involves making the very conservative assumption that all of the vessel (} =aterial could propagate a crack by a lov energy absorption or
.ae s:;a.m 11.1-1 0003 356 15 ..u
cleavage mode. End-of-life vessel conditions were assumed. The crack arrest temperature through the thickness of the vall was developed en a stress-temperature coordinate system. The actual quench-induced, stress-temperature condition through the thickness of the vall at several times during the quench was developed and plotted (Figures 11.1 h and 11.1-5). The maximum depth at which I the material in the vessel vall vould be in tension or at vLich j the stress in the material vould be in excess of the threshold stress for crack initiation (5-8 ksi) was deter =ined by ec=parison of the plots. Cc=parison shows that a crack could propagate only l through the inner 35 per cent of the vall thickness if a crack initiation threshold of 5-8 ksi is applicable, and further that a crack could propagate only through the inner h3 per cent of the vall thickness if a crack initiation threshold of zero vere as-sumed. The foregoing method of analysis is essentially a stress analysis approach which assumes the vorst conceivable material properties and a flav size large enough to initiate a crack. Actually, the outer 83 per cent of the vessel vall is at a temperature above the RTT (NDTT + 60 F) when credit is taken for the neutron shield-ing, and for the original RTT profile through the vall thickness. The analysis is conservative in that it dces not deny that cracks can be initiated, and in that it assu=ed a crack frc= 1 to 2-ft long to exist in the vessel vall at the time of the accident. Therefore, it can be concluded that, if a crack were present in the vorst location and orientation (such as a circumferential1y oriented crack on the inside of the vessel vall), it could not propagate through the vessel vall. / A fra cture mechanics analysis was conducted which assumed a con-tinuous surface flaw to exist on che inside surface of the vessel vall. The criterion used for the analysis is that a crack cannot propagate when the stress intensity at the tip of the crack is be-lov the critical crack stress 1itensity factor (KIC). Using cen-servative values of KIC (for 'ully irradiated cold 302-Grade 3 steel KIC equals 30,000 psi)(3) and the method of E=ery(h) to cal-culate stress intensity factors, K I, in the variable thermal tran-sient stress field, it was found that the crack propagating energy is below that required for crack propagation when the crack reaches a depth of less than 3 in. or 35 per cent of the vall thickness. i 11.1.1 The gec=etry of the plate and the cooling =ethod assumed in the anal-ysis, l ANSWER The analysis assu=ed a long cylindrical section which was insulated l on the outside and subjected to a unifor flev of constant te= era-l ture (90 F) cold water flowing past the inner vall of the reacI;or
- vessel and cuter vall of the ther=al shield. For general di=e.sier.s l of the thermal =cdel and flow patch description, refer to Figure 11.1-1.
l The ecoling =ethod assu=ed in the analysis is as follows:
- a. The metal in the vessel vall and thermal shield is cooled by con- O duction. '
l w' ' .7 ' 11.1-2
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- b. The heat transferred to the fluid is by forced convection.
11.1.2 The heat transfer coefficient used, its experimental basis, and the degree of conservatism involved, ANSWER The analysis used a water film heat transfer coefficient of 3,000 Btu /hr-ft 2 -F. Using the classical (text book) approach,(5) the wate film heat transfer coefficient was calculated to be about 900 Etu/hr ft 2 -F. However, when the water film heat transfer coefficient react a value of 2,000 to 3,000 stu/hr-ft2 -F or more, the heat transfer properties of the metal, i.e., the metal conductivity, vill govern the heat transfer rate, and consequently the shape and variation of the temperature profile through the thickness of the vessel vall vit ti:pe (reference Figure 11.1-2) . The experimental basis and degree of conservatism for the use of a water film heat transfer ccefficient of 3,000 stu/hr-ft2 -F is as fel lows:
- a. The most severe condition that could possibly be postulated woul be to quench the cylindrical portion of the vessel in a quench tank. Much experimental work has been done to determine the wat film heat transfer coefficient for this cendition.(6)(7)(8)
Using Reference 6, the water filc heat transfer coefficient (f) calculated as follows: f = 2Hg k f << 2 x h x 277 f = 2,216 Stu/hr-ft2 -F vhere: f = vater film heat transfer coefficient, Btu /hr-ft2 -F Hg = Giossman's Severity of Quench (= 4 in violently agitated water) k = thermal cenductivity of the material, Btu /hr-ft 2-F/in. (= 277 for SA302GB)
- b. The comparison of our assumed water film heat transfer coefficie:
to the coefficient as calculated by Reference 5 yields a conser-vative ratio of 3.32, and a comparison to the water film heat transfer coefficient, calculated by Reference 6, yields a conser-vative ratio of 1.35. 11.1.3 The initial temperature of the vessel as a function of time delay in injecting the cold water, O v M W 0003 358 11.1-3 9
ANS'4ER The reactor vessel van is prxeted against radiation heating from the hot reactor core by three ;;*.id barriars: (a) the core shroud, (b) the core support barrel, ani (c) the ther=al shield. Each of these barriers is separated by a steam gap so that the reactor vessel is in a sense insulated frcm the hot core. In additica, the core barrel assembly and the thermal shield have considerable mass, i.e. , 63,750 lbs and 47,500 lbs respectively, that =ust be heated before the reactor vessel van is affected. The arrangement of these bar-riers is shown on Figures 3 k5 and 3 h6 of the PSAR. Calculations show that the reactor vessel vall temperature vill not increase as a function of time during the first several hundred seconds of an LOCA. The various ccmponent te=peratures at 500 see and at 1,k00 see are: Temperature Temperature Cc=cenent (at 500 sec), F (at ih00 sec), F Core shroud 731 1,770 Core support barrel 579 770 Thermal shield 576 582 Reactor vessel 576 576 n.1.4 The effect of axial te=perature gradient in the vessel, during fill-ing with cold water, on the total stress intensity, ANS'4ER Figure 11.1-3 shows the temperature profile through the vessel v al ) when the core flooding vater impinges on a section of the vessel vall considering an abrupt line of de=arcation between fluid and steam. The use of such an abrupt line of demarcation between fluid and steam is conservative. The conduction of heat through the vessel produces the gradual temperature change as shown on the isotherm plot on Fig-ure 11.1-3. This temperature digt ibution has been analyzed using the Seal Shen Ccmputer Program,\ 9{e and the results of this analysis are shown as a stress profile en Figure 11.1-3. This stress profile shows that the vorst stress condition is remote from the line of demarcation between fluid and steam, and that the axial conduction has scre than offset any adverse influence of tne uncooled portien of the vessel vall. Therefore, the original analysis , assuming a l long cylinder subjected to a unito, m quench, has presented the vorst
- condition because the effect of tbt axial gradient vill locany de-l crease tb tress produced by ECCS operation in the LOCA.
11.1.5 The errect, of a circumferentiany nonuniform ecoling of the vessel I shell, by the cold water entering the vessel through the injection no::les, en the stresses and distortion in the vessel, l l ANS*4ER The large curvature of the vessel vall results in approximately flat l plate characteristics when censidering local effects of circumferential nonunifers cooling. Again, as in the answer to Questica n.l.h, the O ' l .s a . ,u , 0003 359 11.1 h
1 I I (N g) circumferential gradients vill be =ade gradual and not abrupt by con-l uaction, and vill = ore than offset any influence of the uncooled por-tion of the vessel vall. Thus, it can be concluded here, as in the answer to question ll.l.L. that *.he effect of the nonunifor= gradient vill locally decrease the stress produced by ECCS cperation in the LOCA. 11.1.6 The te=perature profiles and the calculated ther=al stress profiles through the thickness of the plate for several times during the cold water injection transient, ANSWER Figures 11.1-k and 11.1-5 depict the te=perature and stress profiles through the thickness for four times during the LOCA. 11.1.7 The ti=e at which the stress intensities in the base metc1 are maxi-
=u=,
ANSWER The time at which the stress intensities are =axi=um in the base = eta' is approximately h09 see after the start of the injection of the 90 F core fldeding water.- 11.1.8 The value of residual stresses assumed in the base =etal and the veld are as , ANSWER The residual stresses assu=ed were () Ductile =cde: no residual stress. Brittle mode: 12,000 psi residual stress. 11.1.9 The =agnitude of the axial dead lead stresses in the vessel, ANSWER The axial dead load stress was calculated to be h00 psi ec=pression. 11.1.10 The magnitude of the stresses in the vessel shell due to potential st=ultaneous seismic loading, - ANSWER The reactor coolant syste= is designed to withstand the ner al de-sign lead in audition to the design earthquake without exceeding yielt stress in critical cc=penents. These ec=ponents are also designed to sustain no=:a1 leads plus twice the design earthquake without loss of function. Since the earthquake loadir.g vill not produce a loss-of-coolant accident, it is concluded that less-of-coolant accident lead-- ings vill not occur simultaneously with the earthquake condition. t Ecvever, the seismic loading is only a fraction of the dead lead stresses. As shown in the answer to Question 11.1.9 these are neg-ligible. 11.1.11 The value of the yield stress used as the f ailure criterien in the ductile yielding analysis. 5 {d 0003 360 11.1-5
ANSTER The analysis used the mini =us yield strength values as a function of temperature, as listed in Table N h24 of Section III of the ASIE Code. The values of yield stren6th for SA 302, Grade 3, are as follows: Tem erature, F Stress, esi 100 50,000 200 LT,150 300 h5,250 h00 hh,500 500 43,200 600 h2,000 0003 361 O, 9 l l l m'k. .A . .$ aa u 11.1-6 .
O e - czS d (1) Pellini, W. S. and Puzak, P. P., Practical Ccnsiderations in Applying Lab. oratory Fracture Test Criteria to Fracture Safe Design of Pressure Vessel. NRL 6030. (2) Pellini, W. S. and Pu:ak, P. P. Fracture Analysis Diagram Procedures for Fracture Safe Engineering Design of Steel Structures, NRL 5920 (3) Lande man, E., Yanichko, S. E., and Hazelton, W. S., An Evaluation of Rai tion Damage to Reactor Vessel Steels Using both the Transitien Temperature and Fracture Mechanics ApproachesjdAFD. - (h) Emer/, A. F. , " Stress Intensity Factors for Thermal Stresses in Thick Hollow Cylinders," Journal of Basic Engineering, March 1966. (5) Hsu, S Engineering Heat Transfer, Van Nostrand,1963, pg. 301.' (6) Grossman, M. A. , Elements of Hardenabi.11ty. (7) Austin, J. B., Heat Flow in Metals, ASM Publication. (8) Russell, T. F., Russell's Tables. (9) WAPD-TM-398. . (O 0003 362 4 11.1-7
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i i i . 4 6 i e i l 0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 DISTANCE THRU WALL, INCHES C,; ; . L . i h l.- 0004 003 O, PERATURE PROFILES OUE TO ECCS OPER ATION FIGURE 11.1-4
O CURVE TIME FROM START OF ECCS OPERATION h 234 SEC 90 -'j h 409 SEC y 80 - ta-h 643 SEC $ ]
/ 4 O
h 1870 SEC $ w* c w
'o -
m I> e m 2 / 2 2 60 ! h I h 50 - G x 40 - N W .0 - G g 20 -
/
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r0 8
=
10 - /
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0 , , , , f, , j , , , , 1.0 2.0 3.0 4.0 5.0 / 6.0 7.0 8.0 9.0 i
- to - /
* / /
/
l / j l
~ 30 -
i
/
DISTANCE THRU WALL, INCHES 0004 004 1 O TH ERMAL STR ESS DdE TO ECCS OPER FIGURE I1.1-5 l __}}