ML19309C544
| ML19309C544 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 05/01/1967 |
| From: | JERSEY CENTRAL POWER & LIGHT CO., METROPOLITAN EDISON CO. |
| To: | |
| References | |
| NUDOCS 8004080737 | |
| Download: ML19309C544 (57) | |
Text
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METROPOLITAN EDISON COMPANY THREE MILE ISLAND ""L*"* E l
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UNIT 1
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0001 076
TABLE OF CCNTENTS Section Pr 1 INTRODUCTION AND
SUMMARY
. . Volume 1 . . . Tao 1 . . . 1-
1.1 INTRODUCTION
. . . . . . . . . . . . . . . . l-1.2 DESIGN HIGHLIGHTS . . . . . . . . . . . . . . . 1-1.2.1 SITE CHARACTERISTICS . . . . . . . . . . . . . 1-1.2.2 ,
POWER LEVEL . . . . . . . . . . . . . . . 1-1.2.3 PEAK SPECIFIC POWER LEVEL . . . . . . . . . . . l-1.2.k REACTOR BUILDING SYSTEM . . . . . . . . . . . . l-1.2.5 ENGINEERED SAFEGUARDS . . . . . . . . . . . . . 1-1.2.6 ELECTRICAL SYSTEMS AND EMERGENCY POWER . . . . . . . 1-1.2.7 ONCE-TEROUGH STEAM GENERATORS . . . . . . . . . . 1-13 TABULAR CHARACTERISTICS . . . . . . . . . . . . . l-1.k PRINCIPAL DESIGN CRITERIA . . . . . . . . . . . . l-1.k.1 CRITERION 1 . . . . . . . . . . . . . . . l-1.k.2 CRITERION 2 . . . . . . . . . . . . . . . l-1.k.3 CRITERION 3 . . . . . . . . . . . . . . - 1-1.k.h CRITERION k . . . . . . . . . . . . . . . l-1.k.5 CRITERION 5 . . . . . . . . . . . . . . . 1-1.k.6 CRITERION 6 . . . . . . . . . . . . . . . 1-1.k.7 CRITERION 7 . . . . . . . . . . . . . . . l-1.k.8 CRITER'ON 8 . . . . . . . . . . . . . . . 1-1.k.9 CRITERION 9 . . . . . . . . . . . . . . . 1-1.k.10 CRITERION 10 . . . . . . . . . . . . . . . 1-O 1.k.11 1.k.12 CRITERION 11 CRITERION 12 1-1-
1.k.13 CRITERION 13 . . . . . . . . . . . . . . . 1-1.k. k CRITERION lh . . . . . . . . . . . . . . . 1-1.k.15 CRITERION 15 . . . . . . . . . . . . . . . 1-1.k.16 CRITERION 16 . . . . . . . . . . . . . . . 1-1.k.17 CRITERION 17 . . . . . . . . . . . . . . . 1-1.k.13 CRITERION 18 . . . . . . . . . . . . . . . 1-1.k.19 CRITERION 19 . . . . . . . . . . . . . . . 1-1.k.20 CRITERION 20 . . . . . . . . . . . . . . . 1-1.k.21 CRITERION 21 . . . . . . . . . . . . . . . 1-1.k.22 CRITERION 22 . . . . . . . . . . . . . . . 1-1.k.23 CRITERION 23 . . . . . . . . . . . . . . . ?-
l 1.k.2h CRITERION 24 . . . . . . . . . . . . . . . '.
l 1.k.25 CRITERION 25 . . . . . . . . . . . . . . . 1-1.k.26 CRITERION 26 . . . . . . . . . . . . . . . 1-1.k.27 CRITERION 27 . . . . . . . . . . . . . . . 1-1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS . . . . . . . . 1-1.5.1 CNCE-THRCUGH STEAM GENERATOR TEST . . . . . . . . . 1-1.5 2 CONTROL ROD DRIVE LINE TEST . . . . . . . . . . . 1-153 SELF-PCWERED DETECTOR TESTS . . . . . . . . . . . 1-1.5.h THERMAL AND HYDRAULIC PROJRAMS . . . . . . . . . . 1-l 1.6 IDENTIFICATION OF AGENTS AND CONTRACTORS. . . . . . . 1-
1.7 CONCLUSION
S . . . . . . .i. . . . . .. . . . 1-0 1
0001 077
Section 2 SITT AND ENVIRONMENT . . .
Volume 1 * *
- Tab 2 . . .
2.1 GENERAL DESCRIPTION .. . . . . . . . . . ....
2.2 LOCATICN, POPULATION,AND LAND USE . . . . . ....
2.2.1 LOCATION . . . . . . . . . . . . . ....
2.2.2 POPULATION . . . . . . . . . . . . ....
2.2 3 LAND USE . . . . . . . . . . . ....
23 METEOROLOGY , . . . . . . . . . . . ....
2.3.1
SUMMARY
232 SEVERE WEATHER . . . . . . . . . . . ....
2.3 3 AVERAGE AIMOSPHERIC DISPERSICN . . . . . . ....
2.3.h ATMOSPHERIC DIFWSION FOR ASSESSING ACCIDENTS . ....
2.h HYDROLOGY AND GROUNDWATER . . . . . . . . ....
2.4.1 CHARACTERISTICS OF STREAMS IN VICINITY , . . ....
2.h.2 OTHER PCWER PROJECTS IN VICINITY . . . . . ....
2.h.3 LOW FLOW STUDIES . . . . . . . . . . ....
2.h.h FLOOD FLCW STUDIES . . . . . . . . . . . ....
2.h.5 DESIGN '0F PROPOSED DAMS AND SPILLWAYS . . . . . . .
2.4.6 GROUNDWATER . . . . . . . . . . . . . . . .
2.5 GEOLOGY . . . . . . . . . . . . . . . . .
2.6 SEISMICITY . . . . . . . . . . . . . . . . .
2.6.1 SEISMICITY . . . . . . . . . . . . . . . .
2.6.2 RESPONSE SPECTRA . . . . . . . . . . . . . .
2.7 REFERENCES
pJ 3 REACTOR . . . . . . . . Volume 1 . . Tab 3 . . -
31 DESIGN BASES . . . . . . . . . . . . . . . .
3 1.1 PERFORMANCE OBJECTIVES . . . . . . . . . . . .
3 1.2 LIMITS . . . . . . . . . . . . . . . . .
32 REACTOR DESIGN . . . . . . . . . . . . . . . .
3 2.1 GENERAL
SUMMARY
3.2.2 NUCLEAR DESIGN AND EVALUA7CN . . . . . . . . . .
3 2.3 THERMAL AND HYDRAULIC DESIGN AND EVAIJJATION . . . . .
3 2.h MECHANICAL DESIGN LAYOUT . . . . . . . . . . . .
3.3 TESTS AND INSPECTIONS . . . . . . . . . . . . .
331 NUCLEAR TESTS AND INSPECTION . . . . . . . . . .
3.3.2 THERMAL AND HYDRAULIC TESTS AND INSPECTION . . . . . . ;
3 3.3 NEL ASSEMBLY, CONTROL ROD ASSEMBLY, AND CONTROL RCD 1 DRIVE MECHANICAL TESTS AND INSPECTION . . . . . .
3 3.h INTERNALS TISTS AND INSPECTIONS . . . . . . . . . l 3.h REFERENCES . . . . . . . . . . . . . . . . .
h Volume.1 .
REACTOR COOLANT SYSTEM . . . . Tab h . . . . I h.1 DESIGN BASES . . . . . . . . . . . . ...I h l.1 PERFORMANCE OBJECTIVES . . . . . . . . . . . !
k.l.2 DESIGN CHARACTERISTICS . . . . . . . . . . . . I h.1 3 EXPECTED OPERATING CONDITIONS . . . . . . . . . . I h.l.h SERVICE LIFE 1 O
J V h.l.5 CODES AND CLASSIFICATIONS . . . . . . . .- . . . 1 u 0'001 078
I l
Section k p l
h REACTOR COOLANT SYSTEM (CONTINUED)
. . Volume 1 . . Tab.h .
k.2 SYSTEM DESCRIPTION AND OPERATICN . . . . . . . . . . k k.2.1 GENERE DESCRIPTION . . . . . . . . . . . . . k 4.2.2 MAJOR COMPONENTS . . . . . . . . . . . . . . h h.2 3 PRESSURE-RELIEVING DEVICES . . . . . . . . . . . k 4.2.4 ENVIRONMENTE PROTECTION . . . . . . . . . . . . h k.2.5 MATERIES OF CONSTRUCTION . . . . . . . . . . . k l h.2.6 MAXIMUM HEATING AND COOLING RATES . . . . . . . . . k h.2.7 LEAK DETECTION . . . . . . . . . . . . . . . k 4.3 SYSTEM DESIGN EVALUATION . . . . . . . . . . . . k h.3 1 SAFETY FACTORS . . . . . . .' . . . . . . . . k l k.3.2 RELIANCE ON INTERCONNECTED SYSTEMS . . . . . . . . k 4.3 3 SYSTEM INTEGRITY . . . . . . . . . . . . . . k I k.3.4 PRESSURE BELT 7F . . . . . . . . . . . . . . . k l 4.3 5 REDUNDANCY . . . . . . . . . . . . . . . . k l h.3.6 SAFETY ANALYSIS . . . . h l
h.3 7 OPERATICNE LIMITS . . . . . . . . . . . . . . b k.4 TESTS AND INSPECTIONS . . . . . . . . . . . . . k h.k.1 COMPONENT IN-SERVICE INSPECTION . . . . . . . . . h h.k.2 REACTOR COOLANT SYSTEM TESTS AND INSPECTIONS . . . . . k h.k.3 MATERIE IRRADIATION SURVEILLANCE . . . . . . . . . k h.5 REFERENCES . . . . . . . . . . . . . . . . . k 5 CONTAIIDENT SYSTEM . . . . Volume 1 . . . Tab 5 . . 5 51 O' 5.1.1 REACTOR BUILDING .
DESIGN BASES .
5 5
5 1.2 STRUCTURE DESIGN . . . . . . . . . . . . . . 5 52 ISOLATION SYSTEM . . . . . . . . . . . . . . . 5 5.2.1 DESIGN BASES . . . . . . . . . . . . . . . 5 5.2.2 SYSTEM DESIGN . . . . . . . . . . . . . . . 5 5.3 VENTILATION SYSTEM . . . . . . . . . . . . . . 5 531 DESIGN BASES . . . . . . . . . . . . . . . 5 532 SYSTEM DESIGN . . . . . . . . . . . . . . . 5 5.h LEAKAGE MONITORING SYSTEM , . . . . . . . . . . . 5 55 SYSTEM DESIGN EVALUATION . . . . . . . . . . . . 5 5.6 TESTS AND INSPECTION . . . . . . . . . . . . . . 5 5.6.1 PREOPERATIONAL TESTING AND INSPECTICN . . . . . . . 5 5.6.2 POSTOPERATIONAL LEAK MCNITORING . . . . . . . . . 5 6 ENGINEERED 3FEGUARDS . . . Volume 1 . . . Tab 6 . . . 6 6.1 EMERGENCY INJECTION . . . . . . . . . . . . . . 6 6.1.1 DESIGN BASES . . . . . . . . . . . . . . . 6 6.
1.2 DESCRIPTION
. . . . . . . . . . . . . . . . 6 6.1 3 DESIGN EVALUATION . . . . . . . . . . . . . . 6 6.1.4 TESTS AND INSPECTIONS . . . . . . . . . . . . . 6 6.2 REACTOR BUILDING ATMOSPHERE COOLING AND WASHING . . .6 6.2.1 DESIGN BASES . . . . . . . . . . . .. . . . 6 6.
2.2 DESCRIPTION
. . . . . . . . . . . . 6 a
m 0001 079
Section 6 ENGINEERED SAFEGUARDS (CONTINUED) . . Volume 1 . . Tab 6 6.2.3 DESIGN EVEUATICN . . . . . . . . . . . . . .
6.2.4 TESTS AND INSPECTIONS . . . . . . . . . . . . .
6.3 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS . . . . . . . . . . . . . . .
6.31 Ista0 DUCTION . . . . . . . . . . . . . . .
6.3 2
SUMMARY
OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERATIONS . . . . . . . . . . . .
6.3 3 LEAKAGE ASSUMPTICNS . . . . . . . . . . . . .
6.3.h DESIGN BASIS LEAKAGE . . . . . . . . . . . . .
6.3.5 LEAKAGE ANALYSIS CONCLUSIONS . . . . . . . . . .
7 INSTRUMENTATICN AND CCNTROL . Volume.2 . . . Tab 7 . . . .
7.1 PROTEC"' ION SYSTEMS . .. . . . . . . . . . . . .
7.1.1 DESIGN BASES . . . . . . . . . . . . . . .
7.1.2 SYSTEM. DESIGN . . . . . . . . . . . . . . .
7.1.3 SYSTEMS EVEUATICN . . . . . . . . . . . . . .
7.2 REGULATING SYSTEMS . . . . . . . . . . . . . .
7.2.1 DESIGN BASES . . . . . . . . . . . . . . . .
7.2.2 SYSTEM DESIGN . . . . . . . . . . . . . . .
7.2.3 SYSTEM EVALUATICN . . . . . . . . . . . . . .
73 INSTRUMENTATION . . . . . . . . . . . . . . .
7.3.1 NUCLEAR INSTRUMENTATION . . . . . . . . . . . .
732. NONNUCLEAR PROCESS INSTRUMENTAL'ON O
7.3.3 zuccas son roa no SrSrzx . . . . . . . . . . .
7.h OPERATING CONTROL STATICNS . . . . . . . . . . . .
7.k.1 GENERE IAYCUT . . . . . . . . . . . . . . .
7.k.2 INFORMATION DISPLAY AND CONTROL FUNCTION . . . . . .
7.k.3
SUMMARY
CF EARMS . . . . . . . . . . . . . .
7.k.h COMMUNICATION . . . . . . . . . . . . . . .
7.k.5 OCCUPANCY . . . . . . . . . . . . . . . .
7.h.6 AUXILIARY CONTROL STATIONS . . . . . . . . . . .
7.k.7 SAFETY FEATURES . . . . . . . . . . . . . . .
8 ELECTRICE SYSTEMS . . . . Volume 2 . . . Tab 8 . . .
8.1 DESIGN BASES .. . . . . . . . . . . . . . . .
8.2 ELE ~rRICAL SYSTEM DESIGN . . . . . . . . . . . .
8.2.1 NETWORK DPIERCONNECTICNS . . . . . . . . . . . .
8.2.2 STATION DISTF.IBUTION SYSTP.M . . . . . . . . . . .
8.2 3 EMERGENCY PCWER . . . . . . . . . . . . . . .
8.3 TESTS AND INSPECTIONS . . . . . . . . . . ...
6 e
0001 ogg L.r
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l l
Section y 9 AUXILIARY AND EMERGENCY SYSTEMS . . Volume 2 . . Tab 9 9 O
l l 91 MAKEUP AND PURIFICATICN SYSTEM . . . . . . . . . . 9 9.1.1 DESIGN BASES . . . . . . . . . . . . . . . 9 9 1.2 SYSTEM DESCRIm0N AND EVEUAUCN . . . . . . . . . 9 92 CHEMICE ADDm0N AND SAMPLING SYSTEM . . . . . . . . 9
, 9 2.1 DESIGN BASES . . . . . . . . . . . . . . . 9 9 2.2 SYSTEM DESCRIPHON AND EVALUATION . . . . . . . . . 9 93 INTERMEDIATE COOLING SYSTEM . . . . . . . . . . . 9 931 DESIGN BASES . . . . . . . . . . . . . . . 9 9 3.2 SYSTEM DESCRIPTION AND EVALUATICN . . . . . . . . . 9 9.h SPENT NEL COOLING SYSTEM . . . . . . . . . . . . 9 9.h.1 DESIGN BASES . .7. . . . . . . . . . . . 9 9.4.2 SYSTEM DESCRIPHON AND EVALUATION . . . . . . . . . 9 95 DECAY HEAT REMOVAL SYS'"EM . . . . . . . . . . . . 9 9 5.1 DESIGN 3ASES . . . . . . . . . . . . . . . 9 9.5.2 SYSTEM DESCRIPTION AND EVEUAUCN . . . . . . . . . 9 9.6 COOLING WATER SYSTEMS . . . . . . . . . . . . . 9 9.6.1 DESIGN BASES . . . . . . . . . . . . . . . 9 9.6.2 SYSTEM DESCRIPHCN AND EVALUAHON . . . . . . . . . 9 9.7 FUEL HANDLING SYSTEM . . . . . . . . . . . . . . 9 9 7.1 DESIGN BASES . . . . . . . . . . . . . . . 9 9 7.2 SYSTEM DESCRIPTION AND EVALUATION . . . . . . . . . 9 9.8 STATION VENE LATION SYSTEMS . . . . . . . . . . . 9 9 8.1 DESIGN BASES . . . . . . . . . . . . . . . 9 9.8.2 SYSTEM DESCRIPTION AND EVEUAHON . . . . . . . . . 9 10 STEAM AND PCWER CONVERSION SYSTEM . Volume 2 . . Tab 10 . . 1 10.1 DESIGN BASES . . . . . . . . . . . . . . . . 1 10.1.1 OPE ATING AND PERFORMANCE REQUIREMENTS . . . . . . . 1 10.1.2 ELECTRICE SYSTEM CHARACTERISTICS . . . . . . . . . 1 10.1 3 FUNCTIONE LIMITATIONS . . . . . . . . . . . . 1 10.1.h SECONDARY FUNCTIONS . . . . . . . . . . . . . 1 10.2 SYSTEM DESIGN AND OPERATION . . . . . . . . . . . 1 10.2.1 SCHEMATIC FLCW DIAGRAM . . . . . . . . . . . . 1 10.2.2 CODES AND STANDARDS . . . . . . . . . . . . . 1 10.2.3 DESIGN FEATURES . . . . . . . . . . . . . . 1 lo.2.k SHIELDING . . . . . . . . . . . . . . . 1 10.2 5 CO?ROSION PROTECTION . . . . . . . . . . . . . 1 10.2.6 IMPURITIES CONTROL . . . . . . . . . . . . . . 1 10.2.7 RADICACHVITY . . . . . . . . . . . . . . . 1 3.03 SYSTEM ANEYSIS . . . . . . . . . . . . . . . 1 l 10 3.1 TRIPS, AUTCMATIC CONTROL ACTIONS, AND EARMS . . . . . 1 10 3.2 TRANSIENT CONDITIONS . . . . . . . . . . . . . 1 10 3 3 MEMNCTICNS . . . . . . . . . . . . . . . 1 10 3.h OVERPRESSURE PROTECTION . . . . . . . . . . . . 1 10 3 3 INTERACTIONS . . . . . . . . . . . . . . . 1 10 3.6 OPERATICNAL LIMITS . . . . . . . . . . ... 1 10.h TESTS AND INSPECTIONS . . . . . . . . . . . . . 1 0 ,
' 0001 081
Section
.P 11 O RADIOACTIVE WASTES AND RADIATION PROTECTION . . . .. . . . . Volume 2 . . Tab 11 . . 1 11.1 RADI0 ACTIVE WASTES . . . . . . . . . . . . . . 1 11.1.1 DESIGN BASES . . . . . . . . . . . . . . . . 1 11.1.2 SYSTEM DESIGN . . . . . . . . . . . . . . . 1 11.1 3 TESTS AND INSPECTIONS . . . . . . . . . . . . . L 11.2 RADIATICN SHIELDING . . . . . . . . . . . . . . I 11.2.1 PRIMARY, SECCNDARY, REACTOR BUILDING,AND AUXILIARY SHIELDING . . . . . . . . . . . . . 1 11.2.2 AREA RADIATION MONITORING SYSTEM . . . . . . . . . 1 11.2 3 HEETH PHYSICS . . . . . . . . . . . . . . . 1 11 3 REFERENCES . . . . . . . . . . . . . . . . . 7 12 ,
CONDUCT OF OPERATICNS . . . . . Volume 2 . . Tab 12 . . L 12.1 ORGANIZATION AND RESPONSIBILITY . . . . . . . . . . L 12.1.1 FUNCTIONE DESCRIPTION . . . . . . . . . . . . 1.
12.1.2 QUALIFICATIONS . . . . . . . . . . . . . . 1:
12.1.3 ORGANIZATION DIAGRAM . . . . . . . . . . . . . L 12.2 TRAINING . . . . . . . . . . . . . . . . . L 12.2.1 STATION STAFF . . . . . . . . . . . . . . . L 12.2.2 REPLACEMENT PERSONNEL . . . . . . . . . . . . . .1.
12.2.3 ON-THE-JOB TRAINING . . . . . . . . . . . . . 1:
12.2.h ENERGENCY DRILLS . . . . . . . . . . . . . . 1:
12.3 WRITTEN PROCEDURES 1:
O 12 ' arcoats - - - - - - - - - - - - - - - - - - t:
12 5 ADMINISTRATIVE CONTROL . . . . . . . . . . . . . 1:
13 INITIAL TESTS AND OPERATICN . . . Volume.2 . . Tab 13 . . 1 13.1 TESTS PRIOR TO REACTOR FUELING . . . . . . . . . . 1 13 2 INITIAL CKITICALITY . . . . . . . . . . . . . . l !
13 3 POSTCRITICALITY TESG . . . . . . . . . . . . . 1 lk SAFETY ANALYSIS
. . . . . Volume 2 . . Tab lh . . li ik.1 CCRE AND COOLANT BCUNDARY PROTECTION ANEYSIS . . . . 1-lk.l.1 ABNORMALITIES . . . . . . . . . . . . . . . 1- i 1h.1.2 ANALYSIS OF EFFECTS AND CONSEQUENCES . . . . . . . 1 lk.2 STANDBY SAFEGUARDS ANALYSIS . . . . . . . . . . . 11 lb.2.1 SITUATICNS ANALYZED AND CAUSES . . . . . . . . 1 lk.2.2 ACCIDENT ANALYSES . . . . . . . . . . . . . . 11 l
14.3 REFERENCES
. . . . . . . . . . . . . . . . . li 15 TECENICAL SPECIFICATIONS . . . . Volume 2 . . Tab 15 . l' O
<~
0001 082 1
i
TABLE OF APPE:iDICES Accendix 1A TEC'ai! CAL QUALIFICATIONS . . . . .. . Volu=e 3 . . . Tab 1A 2A ENGINEERING GECLOGY AND FOUNDATION CONSIDERATIONS . . . . .. . . .. . Volume 3 . . . Tab 2A 23 SEISMOLOGY AND METEOROLOGY . . . .. . Volume 3 . . . Tab 23 2C GROUND *JATER EYDROLOGY , ... . . . . Volume 3 . . . Tab 2C 2D GEOLOGY . . . . ... . ... .. . . Volume 3 . . . Tab 2D 5A STRUCTURAL DESIGN 3ASEC. .. . . .. . Volume 3 . . . Tab 5A 53 DESIGN PROGRAM FOR REACTOR SUILDING. . Volume 3 . . . Tab 53 SC DESIGN CRISIA FOR REACTOR BUILDING . Volu=e 3 . . . Tab 5C 5D QUALII'l CONTROL. . . . . ...... . Volu=e 3 . . . Tab SD 5E LINER PLATE SPECIFICATION . . .. . . Volu=e 3 . . . Tab 52 5F REACTOR SUILDING INSTRUMENTATION . . . Volume 3 . . . Tab 5F Surele=ent
- 1. . . . . . . . . . . . . . .. .. . . Volume 4 . . Supplement No.1
- 2. . . . . . . . . . . . . . . . . . . . Volume k . . Supple =ent No. 2 s . 3 . . . . . . . . . . . . . . . . . . . . Volume 5 . . Supple =ent No. 3 h . . . . . . . . . . . . . . . . . . . . Volu=e 5 . . S upplement No . k 5 . . . . . . . . . . . . . . . . . . . . Volu=e 5 . . S upple=e nt No . 5
)
(
vi:. (Revised c-23-63) 001 083
. l TABLE OF CONTriTS Section Page 7 E STRGffiTATION AND CONTROL i 7-1 i
)
7.1 PROTECTION SYSTET) 7-1 l 1
7 1.1 DEIGN BASES 7-1 !
7 1.1.1 Vital Functions 7-1 _
7 1.1.2 Principles of Design T-2
\
?.l.1 3 Functional Requirements 7-3 7 1.1.4 Environmental Considerations 7-k 7 1.2 SYSTEM DESIGN 7h 7 1.2.1 System Description - Reactor Protection System 7-4 7 1.2.2 Description - Safeguards Actuation System 7-6 I
7 1.2 3 Design Features 7-7 O 7 1.2.k Summary of Protective Actions 7-10 7 1.2 5 Relationship to Safety Limits 7-11 713 SYSTEMS EVALUATION 7-11 7131 Functional Capability - Reac*.or Protection System 7-11 7132 Functicnal Capability - Safe =uards Actuation System 7-12 7133 Preoperational Tests 7-13 7134 Ccmponent Failure Considerations 7-13 7135 Operational Tests 7-ik l 72 REULATHIG SYSTEG 7-15 7 2.1 DESIGN BASES 7-15 7 2 1.1 Compensatics Considerations 7-15 l 7 2.1.2 Safety consideratiens 7-16
.r .
0'001 084 7-1
COIiTENTS (Cont'd)
O Section Page 7 2.1 3 Startup Considerations 7-16 7 2.2 SYSTEM DESIGN 7-17
) 7 2.2.1 Description of Reactivity Control 7-17 7 2.2.2 Integrated Control System 7-20 723 SYSTEM EVALUATION 7-22 7231 System Failure Considerations 7-22 7232 Interlocking 7-23 7233 Emergency Considerations 7-23 7234 Loss-of-Load Considerations 7-23 73 INSTRUMENTATION 7-25 731 NUCLEAR USTRUMENTATICN 7-25 7 3 1.1 Design 7-25 7 3 1.2 Evaluation .
7-26 732 NONNUCLEAR FROCESS USTRUMENTATION 7-27 7 3 2.1 System Design 7-27 7 3 2.2 System Evaluation 7-28 733 INCORE MONITOR UG SYSTEM 7-29 7331 Design Basis 7-29 7332 System Design 7-29 7333 System Evaluation 7-30 74 OPERATUG CONTROL STATIONS 7-31 7.h.1 GLTZRAL IAYOUP 7-31 I 7.k.2 INFORMATICN DISPIAY AND CCNTROL FUNCTION 7-31 7.4 3 SUNMARY OF AIARIE -
7-32 7.k.4 CCMMUNICATION -
7-32
~
7.u '0001 085
COIITE:ITS (Cont'd)
Section Page 745 cCCUPANCY 7-32 7.k.6 AUXILIARY CCHTEOL STATIONS 7-33 747 SAFEU FEATLE 7-34 O
1 1
O 1 0001 086 7 111 l
l L
LIST OF FIGURES (At rear of Section)
- Figure :To. Title 7-1 Reactor Protection System 31ock Diagram 7-2 !iuclear Instrumentation and Protection Systems 7-3 Typical Control Circuits for Enstneered Safeguards Equipment 7-k Reactor Power Measure =ent . %rs and Control Limits 7-5 Reactor and Steam Temperatures versus Reactor Power 7-6 Reactor Control Dia6:am - Integrated Control System 7-7 Autematic Control Rod Groups - Typical rth Curie versus Distance Withdrawn 7-8 Steam Generator and Turbine Control Diagram -
Integrated Control System 7-9 !Iuclear Instrumentation Flux Ranges 7-10 Iluclear Instrumentation Detector Locations 7-11 tiennuclear Instru=entation Schematic 7-12 Incore Detector Incations 7-13 Typical Arrangement - Incore Instrumentation Channel i
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. 1 0001 087 -
7-iv l
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7 INSTRt:MDTATION AND CCIT"ROL T
71 PROTECTION SYSTEMS The protection systems, which consist of the Reactor Protection System and the Safeguards Actuation System, perform the = cst i=portant control and safety fune:1cns. The protection syste=s extend from the sensin6 instru=ents to the final actuating devices, such as trip circuit breakers and pump or valve =otor contactors.
7 1.1 DESIGN RASES The Reactor Prctection System menitors parameters related to safe opera-tion and trips the reactor to protect the reactor core against fuel rod cladding damage caused by departure from nucleate boiling ( DNB), and to protect' a6ainst reactor coolant system damage caused by high system pres-sure. The Safeguards Actuation System monitors parameters to detect fail-ure of the reactor coolant system and initiates reacter buildin6 isolation and engineered safeguards operntion to :entain radioacti re fission prod-ucts in the reactor buildin6 7 1.1.1 Vital Furctions The Reactor Protection System automatically trips the reactor to protect the reactor core under these conditiens.
- a. The reactor power, as =easured by neutron firx, reaches the 11=1t set by the reactor coolant flow. The resetor coolant flow is determined by the number of operating reactor coolant pu=ps.
- b. The reactor outlet temperature rea:hes en estaclished =ari=um ;
11:15. '
- c. The rsactor pressure reaches an established mini =u= limit.
The Beactor Protection System automati: ally t 1ps the reactor to protect the reactor coolant system under this condition:
- a. The reactor pressure reaches an estaclishei =an=um limit.
The Safeguards Actuation System automatically perfom the following vital functions:
- a. Commands operation of injection *=ergency cora ecoling upon de-tection of attormally low reseter :colant presrurs. *his condi-tien is indicative of a loss-of- .colant accident.
- b. Commands operaticn of the reactor building :coling syste=s upon detection of an abno mally high rea: tor building pressure.
p c. Commands, closing of the reactor building isolation valves upon
\J detection of an abnor= ally high reacter buildin$ pressure.
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- x. 0001 ogg
The core flooding system is a passive system and does not require Safeguards Actuation System action.
g 7 1.1.1.1 Nonvital Functions The Reactor Protection System provides an anticipatory reactor trip vaen the reactor startup rate reaches specified limits.
7 1.1.2 Princieles of Desi.m The protection syste=s are designed to meet the require =ents of the IEEE pro-posed " Standard for Nuclear Power Plant Protection Syste=s" dated Septe=ber 13, 1966. Prototype and final equipment vill be subject to qualificatica I tests as required by the subject standard. The tests vill establish the ad-equacy of equipment performance in both nor=al and accident enviren=ents.
The =ajor design criteria are su==ariced as follows:
7 1.1.2.1 Single Failure
- a. No single component failure shall prevent the protection syste=s from fulfilling their protective functions when action is required.
- b. No single component failure shall initiate unnecessary protection j
system action, provided i=ple=entation does not conflict with the criterion above.
7 1.1.2.2 Redundancy h
All protection system functions shall be i=ple=ented by redundant sensors, in-stru=ent strings, logic, and action devices that combine to form the protection channels.
7 1.1.2 3 Independence j Redundant protection channels and their associated ele =ents shall be electri-i I
cally independent and packaged to provide physical separation.
Separate detectors and instr.= ent strings are not, in general, e= ployed for pro-tection system functions and regulation or control. Sharing instru=entation I for protection and control functions is acco=plished within the framework of the stated criteria by the e= ploy =ent of isolation techniques to the multiple
' outputs of various instru=ent strings. This may be stated as a corollary to the design criteria, i.e., a direct short, open circuit, ground fault, or bridg-ing of any two points at the output te: sinals of an instru=ent string having I =ultiple outputs shall not result in a significant disturbance within more than l one output.
l l 7 1.1.2.4 Loss of Power
- a. A loss of power in the Reactor Protection System shall cause the af-fected channel to trip.
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',o 7-2 (Revised 7-21-67) 0001 089
- b. Availability of power to the Safe 6uards Actuation System shall be continuously indicated. The loss of instrument power, i.e., vital O bus power, to the instrument strin6s and bistables vill initiate a trip in the affected channels. System actuation requires control power from one of the tvc engineered safeguards de power busses so that loss of this power does not actuate the system. The system equipment is divided between the redundant engineered O
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O
-2a (Revised 7-21-67)
safeguards channels in such a vay that the loss of one of the de power busses does not inhibit the system's intended safeguards functions.
7 1.1.2 5 Manual Trip Each protection system shall have a manual actuating switch or switches in the control room which shall be independent of the automatic trip instru-mentation. The manual switch and circuitry shall be s1=ple, direct-actin 6s and electrically connected close to the final actuating device.
7 1.1.2.6 Equipment Removal The Peactor Protection System shall initiate a trip of the channel in-volved when modules, equipment, or subassemblies are removed. Safeguards Actuation System channels shall be designed to provide for servicing a single channel without affecting integrity of the other redundant chan-nels or without compromisin6 the criterion that no single failure shall prevent actuation.
7 1.1.2 7 Testing Manual testin6 facilities shall be built into the protection systems to provide for
- a. Preoperational testing to give assurance that the protection systems can fulfill their required functions.
- b. On-line testing to prove operability and to demonstrate reli-ability.
7 1.1 3 Functional Requirements The functional requirements of the protection systems are those specified under vital functions together with interlocking functions.
The functional requirements of the Reactor Protection System ae to trip the reactor when I
- a. The reactor power, as measured by neutron flux, reaches an al- '
lovable limit set by the number of operating reactor coolant '
pumps. i
- b. The reactor outlet temperature reaches a preset maximum limit. l 1
- c. The reactor coolant pressure reaches a preset maximu:n limit. I
- d. The reactor coolant pressure reaches a preset minimum limit.
- e. The reactor startup rate reaches a maximum limit while operating below a preset pover level. -
O O Interlocking functions of the Reactor Protection System are to to
.. 0001 091 7-3 i
- a. Eypass the startup rate trip when the reactor power reaches a preset value.
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- b. Inhibit control red withdrawal on the occurrence of a predeter-i =ined startup rate, slover than the rate at which reactor trip
! is initiated.
l l The functional require =ents of the Safeguards Actuation System are to
- a. Start operation of high pressure injection upon detection of a low reactor cooLut system pressure.
I b. Start operation of low pressure injection upon detection of a very low reactor coolant' system pressure.
- c. Operate the reactor building isolation valves upon detection of a moderately high reactor building pressure.
I
- d. Start the reactor building e=ergency cooling units upon detec-tion of a =oderately high reactor building pressure.
- e. Start the reactor bu11 din 6 spray system upon detection of a hi6h
- reactor building pressure.
7 1.1.4 Environ = ental Considerations The operating environment for equipment within the reactor building vill nomally be controlled to less than approximately 110 F. The Reactor Pro-h tection System instru=entation within the reactor building is des 1 ned 6 for continuous operation in an environment of 1ho F, 60 psig, and 100 per cent relative humidity, and vill function with less accuracy at the accident temperature.
The environment for the neutron detectors vill be limited to 150 F vith a relative humidity of less than 90 per cent. The detectors are designed for continuous operation in an environ =ent of 175 F, 90 per cent relative hu=idity, and 150 psig.
The Safeguards Actuation System equip =ent inside the reactor building vill be designed to operate under the accident environ =ent of a steam-air mixture.
! Protective equipment outside of the reactor bu11 din 6, control room, and re-lay room is desi6ned for continuous operation in an ambient of 120 F and 90 per cent relative humidity. The centrol room and relay room a=bients vill be =aintained at the personnel comfort level; however, protective equipment in the control room and relay room v111 operate within design l tolerance up to an ambient wuerature of 110 F.
l l 7 1.2 SYSTEM DESIGN 7 1.2.1 System Description - Reactor protection System g
Figure 7-1 is a block diaGra= of the Reactor Protection System. The sys-tem consists of four identical protection channels, each te=1nating in P -
0001 092 ru
a noninverting bistable and reactor trip relay. In the nor=al untripped f
state, each channel functions as an AND gate, passing current to the ter=inating bistable and holding the reactor trip relay energi:ed only if all channel inputs are in the nor=al energized (untripped) state.
Should any one or more inputs to a channel becc=e deenergized (tripped),
the terminating bistable in that channel trips, deenergizing the reactor trip relay. Thus , for trip signals each channel becc=es an CR gate.
Contacts frc= the four reactor trip releys are arranged into two identical 2-out-of h coincidence networks. Sach of these coincidence networks con-trols the power to one of the two identical control rod drive power sup-plies.
The reactor trip circuits are shown in =cre detail en Figure 7-2, which is an overall diagra= sheving the Nuclear Instrumentation Syste= (7-2A),
Reactor Protection Syste= (7-23), and the Safeguards Actuation Syste=
(7-2C). Figure 7-23 shows the circuit breakers centrolling input pcver to each control red drive clutch asse=bly and the canner in which the re-actor trip relays trip.these circuit breakers.
Reactor trip is accc=plished by interrupting all input power to each drive clutch asse=bly. Each centrol red drive clutch power cupply receives its input pcVer through two circuit breakers in series so that opening of either interrupts that source of power. The two control red drive clutch pcVer supplies operate in parallel so that both must be deenergized for the centrol rods to trip. Circuit breakers No. 1 and No. 2 centrol pri-
=ary power to one clutch asse=bly pcver supply, and circuit breakers No.
O d 3 and No. L control power to the other. Thus, reactor trip is acec=plished by tripping one circuit breaker in each of these pairs.
The control rod drive clutch holding coil power supply circuit breakers are equipped with undervoltage coils which =ust be energi:ed for the cir-cuit breaker to be closed or to re=ain closed. The holding voltage for the undervoltage coil of each circuit breaker is taken frc= the vital bus.
1 Referring to circuit breakers Nos. 1 and 3, the undervoltage coils are energi:ed through contacts of trip relays RS1, RS2, RS3, and RSh under i nor-C conditions with all trip relays energised. If trip relays RS1 l and RS2, RS1 and RSh, RS2 and RS3, or RS3 and RSk becc=e deenergized, circuit breakers Nos. I and 3 undervoltage coils vill be deenergized, and the circuit breakers vill open. The trip relays that vill cause circuit breakers Nos. 2 and h to open are RS1 and RS3, RS1 and RSh, RS2 and RS3, or RS2 and RSh. Thus any 2-cut-of h trip relays vill cause either cir-cult breakers Nos. 1 and 3 or circuit breakers Nos. 2 and h to open, re-
=oving power.
The trip circuits and devices are redundant and independent. Each breaker is independent of each other breaker, so a single failure within one trip l circuit does not affect any other trip circuit or prevent trip. By.this arrange =ent each breaker =ay be tested independently by the =anual test switch. One seg=ent of the =anual reacter trip switch is included in )
l each of the circuit breaker trip circuits to i=ple=ent the " direct action in the final device" cri:erion.
I T-5 (Revised 7-21-67) l l
(DIlITED)
The pcVer/flev =enitor icgic details are also shewn en Figure 7-2. There gre four identical sets of power /flev =enitor icgic, one associated with llh each protective channel. Each set of logic receives an independent total reactor coolant flev signal (IF), a " number of pu=p =cters in operation" signal (Pn), and three isolated reacter pcuer level signals (c).
The pcVer/flev =eni .or continuously ec= pares the ratio of the reactor neutron pcVer to the reactor coolant flev. Shculd the reactor power as nessured by the linear power range ;hannels exceed 1.07 times the total reactor ecolant flow, a reacter trip is initiated. All =easure=ents are in terms of per cent full flev er full (rated) pcuer. When the reacter is operating above a predetermined neutron power, X% FP, a reactor trip is initiated i==ediately upon the loss of a single pu=p. Selev this power level a reacter trip is initiated when the reacter power to reacter ecolant flow ratio exceeds 1.07. Thus belev a predeter=ined reactor power there is opportunity for the centrol system to reduce the reactor pcuer to an acceptable level without a reactor trip.
There are four ce=binations of logic functions within the pcVer/flev
=onit r which =ay lead to a reactor tript refer to Figure T-2.
The purpose of ( A1) is to ec= pare the total reactor coolant flow against the nu=ber of operating pu=p notors, P n. Nor-C ly, the loss of a pu=p vill cause an instantaneous decrease in Pn with the flov signal lagging.
Should the reverse ever occur, as might be indicative of a lost pu=p rotor, ( A1) vill initiate a reactor trip if the reactor power is greater than a predetermined value, X% FF (E1).
Belov X% FF, the flux-flev cc=parator (D1) vill trip the reactor when the flux to flev ratio exceeds 1.07 The (31) ce=parator cc= pares the reactor coolant flew against the nu=ber of operating pu=ps to deter =ine that not = ore than one pu=p has been coin-cidentally lost. Should (31) detect the coincident loss of more than one pu=p, the legic is required to deternine that the ratios of reactor pcVer to operating pu=ps (Cl) and reacter power to reacter coolant flov (D1) are both lass than 1.07 If either of these conditions is not satisfied, a reactor trip results.
The (Cl) cc. parator continuously ec= pares the nu=ber of operating pu=p
=ctors against the reacter power. A reacter trip is i==ediately initiated upon loss of a pu=p when the reactor pcuer is above a predetermined value, X% FP (E1). Selev this pcVer level, (Cl) vill not actuate a trip unless the (31) ce=parator detects the 1 css of =cre than ene pu=p.
7.1.2.2 Descrittien - Safeeuards Actuation Syste=
Figure 7-2C shows the action initiating senscrs, bistables, and logic for the Safeguards Actuatien System. The =ajor differences between this sys-tem and the Reactor Protection Syste= are:
- a. Each protective action is ini:iated by either of two channels h with 2-cut-of-3 coincidence logic between input signals.
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" 0001 094 7-6 (Revised 7-21-67)
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- b. Either of the two channels is independently capable of initiat- '
, -s ing the desired ;rctective actica through redundant safeguards equip =ent.
- c. Protective action is initiated by the application of pcuer to the terminating centrol relays through the coincidence logic.
There are three independent sensors for each input variable. Each sensor terminates in a bistable device. The cutputs of the three bistables asso-ciated with each variable are for=ed into two identical and independent 2-out-of-3 coincident logie networks or channels. Safeguards action is initiated when either of the channels associated with a variable beco=es energized through the coincident trip action of the associated bistables.
The engineered safeguards equip =ent is divided between redundant actuation channels as shown in Figure 7-2C. The division of equip =ent between chan-nels is based upon the redundancy of equip =en*; and functions. Where two active safeguards valves are connected in redundant -aeer, each valve vill be centrolled by a separate engineered safeguards channel as shown in Figure 7-2C. When active and passive (check valve) safeguards valves are used redundantly, the active valve vill be equipped with two CR con-trol ele =ents, each driven by cne of the safeguards channels. Redundant safeguards pu=ps vill be centro 11ed in tt e sa=e =anner as redundant ac-tive valves. Figure 7-2C shows a typical control sche =e for both safe-guards valves and pu=ps.
Figure T-3 shows typical centrol circuits for equipcent serving safeguards functions. Each circuit provides for nor=al start-stop centrol by the Station operator as well as auto =atic actuation. Nor=al starting and Ox stopping are initiated by =c=entary contact pushbuttons or control switches.
The centrol circuit shown for a decay heat re= oval pu=p is typical of the centrciler of a large pu=p started by switchgear. There are three decay heat re= oval pu=ps, two are equipped with single centrol relays povered frc= separate safeguards actuation channels. The third pump is equipped with two centrol relays , CR1 and CR2, each of which is powered frc= separate safeguards actuatica channels. Energising the centrol re-lays through their associated safeguards actuation channel, energizes the pu=p circuit breaker closing coi1 and starts the pu=p.
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N.
7-ca (Revised 7-21-67
The centrol circuit for a reactor building isolatica valve is typical of a motor-operated valve which is required to close as its engineered safeguar:is action. If the valve is e= ployed as one of two active redun-dant valves, then it is controlled by a single safeguards actuation channel to CRl. If the valve is empicyed with a passive red"M ant check valve, then the =otor-cperated valve is centrolled by two safeguards ac-tuation channels with CR1 and CR2 connected in an "CR" configuration.
De centrol relays, when energized by their associated safeguards actua-l tien channels, close the valve through contacts which duplicate the man-ual CLOSE pushbutton and at the sa=e ti=e override any existing signal calling for the valve to open. A valve limit switch opens just before the valve seats to pe mit torque closing.
Air-operated engineered safeguards valves automatically go to their engi-neered safeguards position upon loss of centrol air. Valves used with active redundant valves are equipped with a single electrical actuator for centrol by a single engineered safeguards channel as shcvn in Figure 7-2C. Valves used with-redundant passive valves are equipped with two electrical actuators, each centrolle:1 by a single safeguards channel operating in an CR configuration. Engineered safeguards actica is ini-tiated when power is applied to the electrical actuator.
The control of the reactor building spray pumps is by =eans of single control relays in each pump controller. Each pu=p is centrolled by sep-arate engineered safeguards channels. Safeguards action is initiated g when the pu=p centrol relay is energized by its associated engineered safeguards channel.
7 1.2 3 Design Features 7 1.2 3 1 Redundancy he Rea,; tor Protection System is redundant for all vital inputs and func-tions. Redundancy begins with the sensor. Each power range input varia-ble is measured four times by fcur independent and identical instraent strings. Only one of the four is associated with any one protective channel. Se total and ecmplete removal of one protective channel and its associated vital instrment strings vould not i= pair the function of any other instraent or protective channel.
There are two source range channels and two inte=ediate range channels, each with its evn independent sensor. )
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Be Safeguards Actuatica System is also redundant for all vital inputs l and functions. Each input variable is measured by three independent and l identical instru=ent strings. Ihe total removal of any cne instraent I string vill not prevent the system frem performing its intended functions.
7 1.2 3 2 Independence .,
/% The redundancy, as described above, is extended to provide independence in the Reactor Protection System. Each instrment string feeding into e ,. . .
I 7-; (Revised 7-2'.-6 0 0001 096
one protective channel is operationally and electrically independent of every other instn=:ent string. Each protective channel is likewise func-tionally and electrically independent of ever/ other channel.
Only in the coincidence output are the channels brouGh t into any kind of co==on relationship. Independence is preserted in the coincidence cir-cuits throu6h insulation resistance and physical separation of the coin-cidence networks and their switchin6 elements.
The Safeguards Actuation System instru=entation and control have elec, trically and physically independent inst: :nent strings. The output of each bistable is electrically independent of every other bistable. In-dependence is preserved in the coincidence networks throuGh insulation resistance and physical separation of the svitchin6 ele =ents.
7 1.2 3 3 uss of Power The Reactor Protection System initiates trip action upon loss of power.
All bistables operate in a nor= ally energized state and go to a deener-gized state to initiate action. Loss of power thus automatically forces the bistables into the tripped state. F16ure 7-23 shows the system in a deener61 :ed state.
The Safeguards Actuation System instru=entation strings ter=inate in bi-stable trip ele =enta similar to those in the Reactor Protection System.
Loss of instru=ent power up to and includin6 the bistables forces the bi-stables into the tripped state initiating safeguards action. The logic netvorks and the equipment control elements are povered from the Engineered SafeguaMs D-C Power Bus 1 and 2. Electrical safeguards equipment is pov-ered from one of the En61neered Safe 6uards A-C Power busses. Loss of en-Eineered safeguads power to the logic netvorks or to the safeguards equip-
=ent does not initiate safe 6uards action as described in 7 1.1.2.k.
7 1.2 3.k Manual System Trip The =anual actuating devices in the protection systems are independent of the automatic trip circuitry and are not subject to failures that =ake the automatic circuitry inoperable. The =anual trip devices are independent control switches for each power controller.
7 1.2 3 5 Equipment Removal The removal of =cdules or subasse=blies from vitt.1 sections of the Reac-tor Protection System vill initiate the trip no::: ally associated with that portion of the system. The removal criterion is imple=ented in two ways:
(1) advantase is taken of the inherent characteristics of a normally en-ergi:ed system, and (2) interlocks are provided.
An inherent characteristic is illustrated by considerin6 the power supply
. for one of the reactor protective channels. Removal or this power supply auto =atically results in +, rip action by virtue of the resulting loss of power. No interlock is required in such cases. Other instances require a syste= of intericeks built into the equipment to insure trip action upon removal of a portion of the equip =ent.
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The Safeguards Actuation System prevides Ier servicing without affecting the integrity of the redundant channels.
O 7.1.2 3 6 re time The protection systems vill meet the testing criterion and its objectives.
The test circuits win take advantage of the systems' redundancy, indepen-dence, and coincidence features which =ake it possible to initiate trip si6nals manually in any part of one protective channel without affecting the other channels.
This test f'esture vill allow the operator to interrc6 ate the systems from the, input of any bistable up to the final actuating device at any time during reactor operation without disconnecting permanently installed equip-ment.
The test of a bistable consists of inserting an analog input and varying the input until the bistable trip point is reached. The value of the in-serted test signal represents the true value of the bistable trip point.
Thus the test verifies not only that the bistable functions but that the trip point is correctly set.
Prestartup testing vill follow the same precedure as the on-line testing except that calibration of the analog instrument strings may be checked with less restraint then during reactor operation.
As shown in Figure 7-23, the povtr breakers in the reactor trip circuit may also be manually tested during operation. The only limitation is
(~ that not more than one power supply may be interrupted at a time without causing a reactor trip.
7 1.2 3 7 Physical Isolation The physical arrangement of all elements asscelated with the protection systems will reduce the probability of a single physical event impairin6 the vital f.a. 'tions of the system. For example, pressure measurements of reactor coa. ant pressure vill be divided between fcur redundant pres-sure taps un as to reduce the probability of collective dama6e to all sensors by ingle accident.
System equipment vill be distributed between instrument cabiners so as to reduce the probability of damage to the total system by some single event.
Wiring between vital elements of the system outside of equipment housing vill be routed and protected within the unit so as to maintain tne true redundancy of the systema with respect to physical hazards.
71.238 Primary Pever Scuree The primary source of centrol power for tne Reactor Protection System is the vital busses described in 6.2.2 7 The source of power for tne mea-curin6 elements in tne Safeguards Actuation System is also from the vital busses. Ccmmand circuits frem tne Safeguards Actuation System coincidence S -
@ 7-9 0001 098 1
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logic that extend to Engineered Safeguards Equipment centrollers are pcVered from the Engineered Safeguards d-c busses. Engineered Safeguards equip =ent such as pu=p and otor cperators and their starting contacters are pcuered frcs the Engineered Safeguards a-c busses. g 7.1.2.3.9 Reliability Oesign criteria for the Reactor Protection System and the Safeguards Actuation Syste= have been formulated to produce reliable systems. System design prac-tices, such as redundant equip =ent, redundant-channels, and coincidence arrange-nents permitting in-service testing, have been e= ployed to imple=ent reliability of protective action. The beet grades of ec=mercially available ec=penents will ce used in fabrication. A system fault analysis will be =ade considering the nodes of failure and determining their effect en the system vital functions.
Acceptance testing and periodic testing vill be designed to insure the quality and reliability of the ccepleted systems.
7.1.2.3.10 Instrumentation for E=ergency Core Cooling Initiation The instrumentation system =akes use of both physical and electrical isolation.
The high pressure and 1cv pressure injection systems are activated by low pres-sure signals from three pressure transmitters =easuring the reactor coolant sys-tem pressure, as shown in Figure T-11. Two transmitters are connected to one reactor cutlet pipe; the third transmitter is connected to the other reactor I
- utlet pipe. Each transmitter has a separata tap on the reactor coolant piping inside the secondary shield areas.
Che precsure transmitters are physically separated frem each other and located utside the secondary shield inside the rsactor building. Tne electrical out-put of each transmitter leaves the reactor building through a separate penetra-
- ion. The output of each transmitter provides isolated signals to its asso-lll
- iated bistable trip devices. The bistable trip devices of a given icgic func-
- ion are physically separated by cabinet barriers. Each pressure transmitter tnd its associated bistable trips are pcuered from separate battery-backed rital bus pcuer sources, the sa=e power sources which power the reactor protec-
- ien channels.
Two, isolated 125 volt d-c engineered safeguards control power sources are used for the poser to the engineered safeguards channels and logic, is shown in Figure T-2. Eac h major function is , therefore, activated from two
.ndependent sources of cont 1 power.
Se operatien of the engineered safeguards channels and the trip relays ferming due system logic ia described in 7.1.2.2.
The high order of system redundancy assures compliance with the single failure
- riteria of 7 ..l.2.1 without actuation by diversified =eans.
- 1.2.h
. Su==ary of Prctective Acticns 7:ur abncr=al condi:icns which initiate a reactor trip are listed belev:
.., 7-10 (Revised 11-6-67) i s},n
Steady State Trip Value or Trip Variable No. of Sensers Normal Range Condition for Trip Neytron Flux h 0-100% ' 107.5% of full (rated) pcver Neutron; Flux / Reactor k' Flux 1 to k pumps (1) Number of oper-Coolant Flev 16 Reactor Coolant ating coolant pump Pump Monitors motors exceeds total 2 Flev Tubes ecolant flow, and reactor power exceec predeter=ined level.
(2) Ratio of reac-tor power to total reactor coolant flos exceeds 1.07.
(3) More than one reactor coolant pu=p motor is lost, and reactor power exceeds re=aining pump capability by more than 107%.
(k) Reactor power eauseus number of operating pump motor and the reactor pove exceeds predetermine level.
Startup Rate 2 0-2 Decades / min 5 Decades / min Reactor Coolant 4 2,120-2,250 2,350 psig Pressure psig 2,050 psig Reactor Outlet 4 520-603 F 610 F Temperature The reactor trip functions of the power /flev monitor logic are suma-1::ed as follows:
Trip 7criable No. of Sensors Neutron Flux = $ 4 Reactor Coolant h Flev = IF .
No. of Operatin6 Pumps = P 1b n
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T-10a (Revised 11-6-67) 0001 100
- eactor Trie a) () > 1.07P n) and ($ > X%) h b) ($ > 1.07Pn ) and (IF - P )n = L ss of more than ene pu=p c) ($ > X%)* and (P n - IF) = Abnor=al relation of P n > IF d) ($ > 1.07 IF)
Predetermined neutron power level to be specified during detail design.
etions initiated by the Safeguards Actuation System are as follows:
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.w 7-10b (" "1= u u-e-er, g997 ,0I 9 , .. . . . .s.
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Steedy State Action Trip Condition Nor=al Value Trip Point Hish Pressure Low Reactor 2,120-2,250 psig 1,800 psig Injection Coolant Pressure Low Pressure Very Lov Reactor 2,120-2,250 psig 200 psig Injection Pressure Start Reactor H16h Reactor At=ospheric 4 psig Buildin6 E=er- Building Pres-6ency Coolin6 sure Unit and Reactor Building Iso-lation Reactor Building High Reactor Atmospheric 10 psig Spray Building Pres-sure 7 1.2 5 Relationship to safety L1=1ts Trip setpoints tabulated in 7 1.2.4 are consistent with the safety li=1ts that have been established from the analyses described in Section 14. The set point for each input, which must initiate a trip of the Reactor Pro-tection System, has been established at a level that v111 insure that con-trol rods are inserted in sufficient time to protect the reactor core. l
,m Likewise, the set points for parameters initiating a trip of the Safeguards V Actuation System are established at levels that vill insure that corrective i action is in progress in sufficient time to prevent an unsafe condition.
Factors such as the rate at which the sensed variable can change, instru-mentation and calibration inaccuracies, bistable trip times, circuit breaker I trip times, control rod travel times, valve travel ti=es, and pump starting ti=es have been considered in establishing the targin between the trip set points and the safety limits that have been derived.
The flux trip set point of 107 5 per cent is based upon the tolerances and error bands shown in Figure 7-4. The incident flux error is the sum of the errors at the output of the measuring channel resulting from rod motion, and instrument drift during the interval between heat balance checks of nuclear instrumentation calibration.
713 SYSTEMS EVALUATION 7131 Functional Capability - Reactor Protection System The Reactor Protection System has been designed to limit the reactor power to a level within the design capability of the reactor core. In all acci-dent evaluations the time response of the sensors and the protective chan-nels are considered. Maximum trip times of the protection channels are '
listed below.
O f*' ! l 7-11 '(Revised T-21-67) . 02
e
- a. Te=perature - 5 see
- b. Pressure - 0 5 see O
- c. Flux - 0.3 sec
- d. Pu=p =cnitor - 1.0 see Since all uncertainties are censidered as cu=ulative in deriving these times , the actual times =ay be only one-half as long in =cs: cases. Even these maximum times, when added to control red drop times, provide conser-vative protective action.
The Beacter Protection Syste vill 11:1t the power that might result frc=
an unexpected reactivity change. Any change of this nature vill be de-tected and arrested by high reactor coolant te=perature, high reactor ecolant pressure, or high neutron flux protective action.
An uncentrolled rod withdrawal frc= startup will be detected by the ab-non: ally fast startup rate in the intermediate channels and high neutro 1 flux in the pcver range channels. A ctartup rate trip frc= the inter-mediate-range channels is incorporated in the Reactor Protection System.
A rod withdrawal accident at power vill b:=ediately result in a high neu-tren flux trip.
Reduced reactor coolant flow results in a reduced allevable reactor power.
The reactor coolant pump =enitor operates to set the appropriate reactor power limit by adjusting the pcVer level trip point. A total loss of flev results in a direct reactor trip, independent of reactor power level.
Two =ajor =easure=ents feed the pover/flev =cnitor: (a) reactor coolant ficv, and (b) neutron power level. The flow tubes which provide the re-actor coolant flov =easure=ent vill exhibit no change during the reactor life. A periodic calibration of the flew trans=itters will be =ade. The neutron power level signal vill be recalibrated by ec=parison with a reu-tine heat balance. The power range channels use detectors arranged to effectively average the =easurement over the length of the core as de-scribed in 7.3.1.1.2. Therefore, their output is expected to be within k per cent of the calibrated value during nor=al regulating red group I position changes and the need for additional calibration thereby eliminated.
l l A less of reactor coolant vill result in a reduction of reactor coolant pressure. The low pressure trip serves to trip the reactor for such an occurrence.
A significant turbine-side steam line rupture is reflected in a drop of reactor ecolant pressure. The low reacter pressure trip shuts devn the Station for such an occurrence.
O cp
- ~0001 103 T-12 (Revised 7-21-oT)
E 7.1.3.2 Functional Carability - Saferuards Actuatien System The Safeguards Actuation System is a draded protection system. The pro-gressive actions of the injection syste=s as initiated by the Safeguards Actuatien Systes provide sufficient reacter ecclant under all conditions while =intnizing the possibility of setting the entire syste= in operation inadvertently.
The key variable associated with the less of reae cr coolant is reacter pressure. In a less-of-reacter-ccolant accident, the reactor pressure vill fall, starting high pressure inje; tion at 1,800 psig. If hizh pres-sure ir.jection does not arrest the pressure drop, then low pressure in-jectica starts upon a signal of 200 psig.
l l l
l I
a l
l I
o I
r 1
., 0001 !04. :
7-122 (.:.evised 7-21-c7)
s The key variable in the detection of an accident that could endanger re-3 actor building integrity is reacter building pressure. A reactor building pressure of k psig initiates operation of the reactor building emergency cooling unit and isolation of the building while a higher pressure of 10 psig initiates operation of the reactor building sprays.
7.1 3 3 Preoperational Tests Valid testing of analog sensing ele =ents associated with the protection systems will be acec=plished through the actual =anipulation of the measured variable and comparison of the results against a standard.
Routine preoperational . tests will be perfor=ed by tne substitution of a calibrattng signal for the senser. Simulated neutron signals may be sub-stituted,in each of the source, inter =ediate, and pcuer range channels to check the operation of each channel. Si=ulated pressure, te=perature, and level signals =ay be used in a similar fashion. This type of testing is valid for all elements of the syste= except the sensors. The sensors should be calibrated ~against standards during shutdowns for refueling, or whenever the true status of any measured variable cannot be assessed because of lack of agreement a=eng the redundant measurements.
The final defense against sensor failure during operation will be the Station operator. The redundancy of measurements provides more than ade-quate opportunity for cc=parative readings. In addition, the redundancy of the systems reduces tne consequences of a single sensor failure.
O) k s- 7.1.'3. 4 Cc=ponent Failure Considerations
!1e effects of failure can be understood through Figure 7-23. In the R9 actor Protection Syste=, the failure of any single input in the " tripped" direction places the syste= in a 1-cut-of-3 =ede of operation for all vari-ables. Failure of any single input in the "cannot trip" direction places the syste= in a 2-out-of-3 = ode of operation for the variatie involved, but leaves all other variables in the normal 2-cut-of L coincidence mode.
If the fault were of the " tripped", open circuit =cde, then the system would be able to tolerate a =inimum of two "cannot trip", short circuit failures within the same =easured variable before cc=plete safety pro-tection of the variable were lost. With one " tripped", open circuit fault, a second identical fault within the same variable would trip the reactor.
A similar fault relationship exists between channels as a result of the 2-out-of 4 coincidence output. One " trip" faulted channel places the syste= in a 1-out-of-3 or single-channel mode. A "cannot trip" faulted channel places the syste= in a 2-cut-of-3 = ode.
At the final device, a " trip" faulted power breaker does not affect the protective channel =cde of operatien, reacter trip being dependent upon one of two breakers in the unaffected primary power supply Oc the control
,_ rod drives. A breaker faulted in the "cannot trip" =ede leaves the sys-( ) tem dependent upon the second breaker in the affected pri=ary power sup-ply.
5 i '
0001 105 7-13 (Revised 7-21-o7)
The Safeguards Actuation System is a 2-out-of-3 input type of system. It can tolerate one fault of the "cannot trip" variety in each of the coin-cidence networks. yor this type of fault, all re=aining inputs must func-h tion correctly. A " tripped" input fault allows any cne of the two re=ain-ing inputs to initiate action.
Pri=ary power input to both protection systems has been arranged to mini-
=1:e the possibility of loss of power to ei*Jaer protection system. Each channel of the protection system vill be supplied from one of the four vital busses described in 8.2.2 7 The operator can initiate a reactor trip independent of the aute=atic protection action.
The engineered safeguards have been connected to multiple busses to mini-mi:e total loss of safeguard capability. The individurd parts of the Safeguards Actuation System can be placed in operation through =anual op-erator controls independent of the auto =atic protection equip =ent.
7135 operational Tests The protection systems are designed and have the facilities for routine manual operatienal testing.
Most inputs to the protection systems or161nate from an analog =easure-
=ent of a particular variable. Every input of this type is equipped with a continuous readout device. A routine check by the operator of each reading as ec= pared to the other redundant readin6s available for each variable vill uncover =easurecent faults. These ele =ents plus the bi-stables and relays of the protection syste=s require a periodic dynamic test. Each system provides for routine testing. Each bistable =ay be manually tripped, and the results of that trip traced through the system logic and visually indicated to the operator. The trip point setting of each bistable =ay be verified by the application of an analog signal pro-portional to the =easured variable, and that signal =ay be varied until the bistable ele =ent trips.
O
. 0001 106 tb' ..a .
7-lk
l 72 REGUUTING SYSTEMS 7.2.1 DESIGN 3ASES 7.2.1.1 Compensation Considerations Reactor regulation is based upon the use of movable poison (control rods) and chemical neutron poison (boric acid) dissolved in the reactor coolant.
Relatively fast reactivity effects including Doppler, xenon, and moderator temperature are controlled by tne control rods, which are capable of rapid compensation. Relatively slow reactivity effects, such as fuel burnup, fission product buildup, samarium buildup, and hot-to-cold moderator def-icit, are controlled by soluble poison.
It is possible to chan6e the reactor coolant system boric acid concentra-tion to " follow" xenon transients over approximately 70 per cent of each core cycle without centrol rod operation. However, to reduce waste han-dling requirements resulting from chemical shim operation, control rods are used throughout core life for xenon transient associated with normal power changes. Chemical shim is used in conjunction with control rods to compensate for equilibrium xenon conditions.
At the beginning of first core life when the mcderator temperature re-activity coefficient =ay be zero or slightly positive, the control rod drive response is faster than necessary to maintain the power error with-in the allowed deadband. Analog computer analysis shows that the only O~ es ee t= co=tro1 resro se vae= vositive =oderator coefricie== or re-activity exists is an increased frequency of control rod motion.
The reactor controls are designed to maintain a constant average reactor coolant temperature over the load range from 15 to 100 per cent of rated power. The steam system operates on constant pressure at all loads. The average reactor coolant temperature decreases over the ran6e from 15 per cent load to cero load. Figure 7-5 shows the reactor coolant and steam temperatures over the antire load range.
Input signals to the reactor controls include reactor coolant average temperature, megavett demand, and reactor power as indicated by out-of-core neutron detectors. The soluble poison dilution is initiated manu-ally and terminated automatically or manually. Manual rod control is used belev 15 per cent of rated power. Autc=atic or =anual red centrol
=ay be used above 15 per cent of rated power.
Increasing power transients between 20 and 90 per cent pcver are IL=ited to rs=p changes of 105/ sin and step increases of 10 per cent. pcver in-creases frc= 15-20 per cent and above 90 per cent are limited to 5%/ min.
Decreasing pcver transients between 100 and 15 per cent pcuer ar 'd-d ed to rs=p changes of 10%/=in and step decreases of 10 per cent. The tur-bine bypass systes per=its a load drop of h0 per cent or a turbine trip fres k0 per cent 1 cad vittcut safety valve operation. The turbine bypass i system and safety valves permit a 100 per cent lead drop without turbine
,) trip to satisfy "blackcut" requirements as described in ik.1.2.3.2.
J .
p~ '*\'
>s 7-12 (Revised 7-21-o7) l07
7.2.1.2 Safety considerations 7 2.1.2.1 Shutdown Margin .
The centrol rods are provided in sufficient number to allow a hot shutdown that is greater than 1 per cent suberitical with the rod of greatest worth fully withdrawn and a typical level of soluble poison (F16ure 3-1).
7.2.1.2.2 Reactivity Rate Limits The maxi =um average rate of change of reactd vity that can be inserted by any group of rods does not exceed 5.8 x 10 ~3 ak/k/sec. (The accidental l
vithdrawal of the rod group of greatest worth is discussed in 14.1.2.2 and 14.1.2 3.)
The maxi =um nor=al rate of pure water addd . ion does not change reactivity vorth more than 3 x 10-6 dc/k/sec. React aity control may be exchan6ed l between rods and soluble poison consistent with the design bases listed above.
7 2.1.2 3 Power Peaking Limits l
The nominal reactivity available to a power regulating control rod group is limited so that established radial and axial flux-peaking li=its are
- not exceeded with the rod group in any position at power levels up to 100 l per cent power.
l 7 2.1.2.4 Power Level Limits O The reactor automatic controlis incorporate a high limit and a lov limit of power level demand to the reactor. Limits are imposed on reactor mega-vatt demand by lack of feedvater flow capability and reactor coolant sys-tem flow capability.
7.2.1 3 Startup considerations Over the life of the nuclear unit, startup will occur at various tempera-ture levels and after varying periods of downtime. Examples of regulating system design requirecents as related to startup are l a. Control rod and/or control rod group "vithdraw inhibit" on hi 6h l startup rate (short period) in the source range and intermediate range.
- b. Reactor trip on high startup rate in the intermediate ran6e.
- c. Startup control = ode. This mode prevents auto =atic rod with-drawal below 15 per cent power.
- d. In startup control = ode, the controls are arran6ed so that the steam system follows reactor power rather than turbine system power demand. The controls vill limit steam du=p to the con-denser when condenser vacuum is inadequate. ,
wi .1 r 0001 108 7-16
- e. Sufficient control rod vorth is provided to cverride peak xenon and return to power folleving a hot shutdown er het stardby.
During cold shutdevn it vill be necessary to increase boron concentratien to =aintain shutdcyn margin. Folleving a cold shutdown, boron concentration chan6es vill be =ade durin6 start-up. A number of rods (or Broups), sufficient to provide 1 per cent shutdevn =argin during startup, are required to be with-drawn before a dilution cycle.
- f. Minimum pressurizer water level conditions =ust be =et before and during startup.
7 2.2 SYSTEM DESIGN 7 2.2.1 Description of Reactivity Control 7 2.2.1.1 General Description The reactor controls move centrol reds to regulate the pcVer cutput of the reactor and maintain.censtant reactor coolant average temperature above 15 per cent rated pcVer. As shown in Figure T-6, the =egawatt de=and sig-nal is added to the reactor coolant avera6e temperature error to form a reactor power level demand signal. The reactor power lavel demand signal is compared to the reactor power level measured by a pcVer range detector in the nuclear instru=entation. "nen the resulting reactor power level error signal exceeds the deadband, the output signal is a control red drive "withdrav" or " insert" ce==and to the centrollins rod 6rcup. For reactivity control limits see 3.1.2.2.
O 7 2.2.1.2 Reactivity Control Reactivity control is =aintained by movable control rods and by soluble poison (boric acid) dissolved in the reactor ecolant. The moderator tem-1 perature coefficient (cold to hot critical), as well as long-tem reactiv-ity changes caused by fuel burnup and fission product poisoning, are con-trolled by adjusting soluble poison concentration. Short-tem reactivity changes caused by power change, xenon poisoning, and moderator tempera-ture change frem 0 to 15 per cent power are controlled by control rods.
First-cycle values for the reactivity ecmponents and control distribution are listed in Tables 3 h and 3-5 Twenty-cne of the 69 control rods are assigned to automatic control of re-actor power level during the first core cycle. Thereafter, 25 reds are used. These control rods are arranged in four sy= metrical groups which operate in sequence. The position of one automatic group is used as an index to soluble poison dilution. Soluble poisen adjustment is initiated manually and teminated autcmatically. The position of this group acts as a "pemissive" to restrict the start of dilution to a " safe" rod posi-tion pattern. The position of the same grcup teminates dilution auto-matically.
p During reactor startup, control rods are withdrawn in a predetem ined
- ( sequence in symmetrical groups of fcur or more rods. The group size is preset, and individual control red assignments to a group are made at a i .; .
7-17 (Revised 7-21-67)
- ' 09
control rod grouping panel. However, the operator can select any individual control red and any control rod group for motion as required. g A typical control rod group withdrawal scheme is as follows:
First Cycle Equilibrium Cycle Group 1 16 rods 12 rods Group 2 12 rods 12 rods Group 3 12 rods 12 reds Group 4 8 rods 8 rods G.-cup 5 h rods' '8 rods
- "E # 8 9 r ds Group 7 5 rods > Regulating Groups h rods Group 8 h rods, h rods (DELETED)
An automatic sequence logic unit is used for reactor control with four regu-lating rod groups in the power range. Thus unit allows operation of no more than one control rod group simultaneously except over the last 25 per cent travel of one group and the first 25 per cent travel of the next group when overlapping motion of two groups is permitted. This tends to linearize the reactivity insertion from group to group as shown in Figure 7-7.
As fuel burnup progresses, dilution of the soluble poison is controlled as fol-lows:
When the partially withdrawn active control rod group reaches the fully with-drawn point, interlock circuitry permits setting up a flow path from a demin-erslized water tank, in lieu of the normal flow path of borated makeup, to the reactor coolant system. Demineralized water is fed to the reactor coolant sye-tem, and borated reactor coolant is removed.
The reactor controls insert the active regulating group to compensate for the reduction in poison concentration. When the control group has been inserted to the 75 per cent withdrawn position, the dilution flow is automatically b Lcked. The dilution cycle is also terminated automatically by a preset tim-ing device, which is independent of rod position. Normally, a dilution cycle is required every several days.
l l 7.2.2.1.3 Reactivity Worth l The maximum verth of any group of the four autcmatic control groups is approx-l I
imately 1.25 ak/k. At design speed, a group requires approximately 6 minutes to travel full stroke. This rate of control rod group travel results in a re-activity rate of 5.8 x 10-5 ak/k/sec.
O 7-18 (Revised 11-6-67) g, ; ,,, ,,,
The =axi=u= rate of reactivity additica with the soluble poison system, i.e. ,
s jecting unborated water frem the =akeup system at To gym =ax1=um, is 3.0 x 10-ak/k/sec.
Table 3-5 snows a shutdevn reactivity analysis. ~he rod worth provided gives shutdown =argin of 5.1". ak/k or = ore under normal conditions , and a margin in excess of 15 ak/k with the red of greatest verth stuck in the withdrawn posit Under conditions where cocidevn to reactor building ambient conditions is re-quired, concentrated soluble poison vill be added to the reacter ecolant to pr duce a shutdevn =argin of at least 1% ak/k. The reactivity changes from hot
- ero power to a cold conditien, and the corresponding increases in beric acid concentration, are listed in Table 3-6.
7.2.2.1.k Reacter Centrol D.e reactor ccatrol is =ade up of analog ec=puting equip =ent with inputs of =.
vatt demand, core power, and reactor coolant average te=perature. The output of the centroller is an error signal that causes the centrol rod drive to be positioned until the e rcr signal is within a deadband. A block diagram of the reactor control is shown in Figure 7-6.
Firrt, reactor power level demand (:I d ) is cc=puted as a function of the meg demand (W d ) and the reactor coolant system average temperature deviation (g AT frc= the set point, according to the following equation:
lid"E#d+K2 l ( a7 + [afdt)
O Megawatt demand is introduced as a part of the demand signal through a propor tional unit having an adjustable gain factor (K 1). The temperature deviation is introduced as a part of the demand signal after proportional plus reset (i:
tagral) action is applied. For the temperature deviation, K2 is the adjustab c & and e is the adjuststle integration factor.
The reactor power level demand (?!3) is then ec= pared with the average reactor power level signal (ii i) which is_ derived from the nuclear instrumentation.
The resultant error signal (Il d - fl i) is the reactor pcuer level error signal (7p)-
When the reactor pcVer level error signal (E )p exceeds the deadband settings ,
the control rod drive receives a command that withdraws or inserts rods depent ing upon the polarity of the pcVer error signal.
The following additienal features are provided with the reactor power control.'
- a. An adjustable low limit on the megawatt de=and signal (W d ) to cut (
the automatic reactor centrol action.
- b. A high limit on reactor power level damand (:tq).
(a ,t ) ,<3 ,,
.<>.,g.
7-19 (Revised 11-6-67) 000l JJ;
- c. An adjustable icv limit on reactor pcVer level demand (ig ).
ieparate fec=, but related to, the autc=atic reacter centrol syste= is the re-tetor ecolant flov signal system. Pcver to each reacter coolant pu=p motor is
- cnitored as an indication of reactor coolant flev. Lcgie units conti::uously
- c= pare the nu=ber of energi:ed pu=ps to the =easured reactor power to sense
. hat tne flew is adequate for the operating pcVer level. If the flow is icv,
.he reactor power level demand is reduced by the Integrated Control System.
.2.2.2 Integrated Centrol System he Integrated Control System maintains constant average reactor ecolant tem-erature and constant steam pressure in the nuclear unit during steady state
.nd transient operation between 15 and 100 per cent full power. Figures 7-6
.nd 7-8 show the overall system. The system is based on the Integrated Boiler-
'urbine concept videly used in fossil-fuel-fired utility plants. It combines he stability of a turbine-follev!ng system with the fast response of a boiler-
'olleving system. Opti=um overall Station performance is maintained by limit-ng steam pressure variations; by limiting the unbalance that can exist among he steam generator, turbine, and the reactor; and by limiting the total Sta-ion load demand upon less of capability of the steam generator feed system, he reactor, or the turbine generator.
igure 7-6 shcus the reactor control portion of the Integrated Control System escribed in 7.2.2.1.h. Figure 7-d shows the steam generator and turbine con-rol portion of the Integrated Control System. This control receives inputs f megawatt demand, system frequency, and steam pressure, and supplies output ignals to the turbine bypass valve, turbine speed changer, and steam generator eedvater flow controls with changing operating conditions. g he turbine and steam generator are capable of autcmatic control from :ero over to full power with optional manual control. The reactor controls are esigned for manual operation belev 15 per cent full power and for automatic r manual operation above 15 per cent full power.
he turbine is operated as a turbine-folleving unit with the turbine control alve pressure set point varied in proportion to megawatt error. The steam enerator is operated as a boiler-folleving system in which the feedvater flow emand to the steam generator is a su=mation of the megawatt demand and the team pressure error.
ae Integrated Control System obtains a lead demand signal from the system dis-atch center or from the operator. A frequency loop is added to match the speed rop of the turbine speed controls. The lead demand is restrained by a maximum
- ad limiter, a minimum load limiter, a rate limiter, and a runback limiter.
a nor=al operation the limits vould be set as follows:
Maximum lead limit 102%
l Mini =u= lead limit 15%
Rate limit 10%/ min O
O '
['i 7-20 ( Revised ll-e-67) 0001 !12
1 Se runbacks act to runback and/or limit the ict.d demand on any of the folleving conditions:
- a. One or more reacter coolant pumps are incperative.
I
- b. Total feedvater flev lags total feedwater da-a-d by =cre than i 5 per cent.
l'
- c. The four shim / safety red groups are not fully withdrawn. j
- d. Asy= metric rod withdrawal patterns exist.
- e. The generator separates frem the 500 kv bus.
- f. The generator loses stator cooling.
I
.j
'l 8
0001 !13 T-2Ca (.2.a'tisel 7-21-c7) ,
1 (DELETED)
The output of the limiters is a megawatt demand si$ nal which is applied to the turbine control, steam generator control, and reactor control in parallel. The reactor control responds to the megawatt demand signal as described in 7 2.2.1.h.
7 2.2.2.1 Turbine Control The megawatt demand is co= pared with the generator megawatt output, and the resulting megawatt error signal is used to change the steam pressure set point. The turbine valves then change position to control steam pres-sure. As the megawatt error reduces to zero, the steam pressure set point is returned to the steady state value. By limiting the effect of mega-vatt error on the steam pressure set point, the system can be adjusted to pemit controlled variations in steam pressure to achieve any desired rate of turbine response to megawatt demand.
7 2.2.2.2 Steam Generator Control Control of the steam generator is based on matching feedvater flow to megawatt demand with bias provided by the error between steam pressure set point and steam pressure. The pressure error increases the feedvater flow demand if the pressure is lov. It decreases the fegvater flow de.-
mand if the pressure is hish.
The basic control actions for parallel steam generator operation are
- a. Megawatt demand converted to feedvater demand.
- b. Steam pressure compared to set pressure, and the pressure error .
converted to feedvater demand.
- c. Total feedvater demand computed from sum of a and b.
- d. Total feedvater flow demand split into feedvater flow demand for each steam generator.
- e. Feedvater demand compared to feedvater flov for each steam gen-i erator. The resultins error signals position the feedvater flow controls to match feedvater flow to feedvater demand 'or each steam generator.
- f. The generator loses stater cooling.
For operation belov 15 per cent load, the steam generator control acts to
=aintain a preset minimum downecmer water level. The conversion to level control is autcmatic and is introduced into the feedvater control train tnrough an auctioneer. At low loads below 15 per cent, the turbine bypass valves vill operate to limit steam pressure rise.
p The steam generator centrol also provides ratio, limit, and runback actions d as shown in Figure 7-8, which include
((; ,' ,;3, , 7-21 (Revised 10-2-o7) 4 0001 !14
- a. Steam Generator Lead Ratio Control Under normal conditions the steam generators will each produce one-half of the total load. Steam generator load ratio control is provided to balance reactor inlet coolant te=peratures durin6 1 operation with more reactor coolant pu=ps in one loop than in the other,
- b. Rate Limits Rate limiters restrict loading or unloading rates to those that are ecmpatible with the turbine and/or the steam Generator,
- c. Water Level Limits A maximum water level limit prevents gross overpu= ping of feed-water and insures superheated steam under all operating condi-tions.
A minimum water level limit is provided for low load control.
- d. Reactor Coolant Pump Limiters These limiters restrict feedwater demand to match reactor cool-ant pumping capability. For example, if one reactor ecolant pump is not operating, the maximum feedwater demand to the steam generator in the loop with the inoperative pump is limited to approximately one-half normal. lll
- e. Reactor Outlet and Feedwater Low Temperature Limits These limiters reduce feedwater demand when the reactor outlet temperature or the feedwater temperature is low,
- f. Feedwater Pump Capability A feedwater pump capability runback signal limits the megawatt demand signal whenever total feedwater flow lags total feed-water demand by 5 per cent.
3 SYSTEM EVALUATION 31 System Failure Considerations
( undant sensors are available to the Integrated Control System. The rator can select any of the redundant sensors from the control room.
l
( ual zwsetivity control is available at all power levels.
l l s c- electrical power to the automatic controller reverts reactor control
! tP; manual mode.
- i, O
0001 !15 bIE *'ic 7-22 ' (Revised T-21-oT)
~~
7232 Interlocking
() Control rod withdrawal is prevented on the occurrence of a positive short pe-riod below 10 per cent power.
The automatic sequence logic sets a predete::ined inse tion and withdrawal pat tern of the four regulating red Groups.
Control circuitry allows =anually selected operation of any single control rod or control rod group tnroughout the power range.
An interlock vill prevent actuation of both withdrawal and insertion of con-trol rods simultaneously with the insertion signal overriding the withdrawal.
Control rod drive switching circuits allow witndrawal of no more than a single control rod group in the manual = ode.
The autocatic sequence lo61 c limits regulating rod =otion to one group out of four at one ti=e except at the upper and lover 25 per cent of stroke where operation of two groups is per=itted to linearize reactivity versus stroke.
Maximum and minimum limits on the reactor power level demand signal (:Td ) P#'~
vent the esctor controls from initiatin6 undesired power excursions.
Maxi =u= and mini =um levels on the =egawatt demand signal (!G d ) prevent the re.
actor controls from initiating undesired power excursions.
() 7233 E=ergency Considerations Loss of power to the control rod drive ma6netic clutch initiates a reactor trip.
'4 hen e=ergency conditions arise that exceed the capability of the control sys-tem, the operator can revert to the manual control = ode.
7 2 3.h Loss-et-Load considerations The nuclear unit is designed to accept 10 per cent step load rejection withr.t.'
safety valve action or turbine bypass valve action. The combined actions of the control system and the turbine bypass valve permit a LO per cent load re-duction or a turbine trip from k0 per cent load without safety valve action.
The controls will limit steam dump to the cendenser when condenser vacuum is inadequate, in which case the safety valves may operate. The combined action.
of the control system, the turbine bypass valve, and the safety valves per=1t a 100 per cent load rejection without turbine trip. This permits the unit to ride through a " blackout" condition, i.e., sudden rejection of electrical loa down to auxiliary load without turbine trip. (The " blackout" provisions are discussed in 14.1.2.8.2.)
The features that permit continued operation under load rejection conditions include O
V '
'. . t ...
7-23 (Revised 7-21-o7) 000t il6 '
L.
- a. Integrated Control System During nor=al operation the Integrated Control System (see Fig-ure 7-8), controls the Station load in r- pense to load d::and from the c, . dispatch center or from the operator. During nor=al load changes and small frequency changes, turbine con-trol is through the speed changer to =aintain constant steam pressure.
Durin6 large load and frequency upsets, the turbine governor takes control to regulate frequency. For these upset conditions, frequency error at the input to the inte6 rated control system becomes more i=portant In providing load =atchin6
- b. 100 Per Cent Relief Capacity in the Steam System This provision acts to reduce the effect of large lead drops
. on the reactor system.
Consider, for example, a sudden load rejection greater than 10 per cent. When the turbine generator starts acceleratin6, the l
Governor valves and the intercept valves begin to close to =ain-l tain set frequency. At the same ti=e the cegavatt demand signM is reduced, which reduces the governor speed changer setting, l feedvater flow de=and, and reactor power level de=and. As the governor valves close, the steam pressure rises and acts throu6h the control system to reinforce the feedvater flow de=and re-duction already initiated by the reduced =eSavatt demand signal.
In addition, when the load rejection is of sufficient magnitude, h
the turbine bypass valves open to reject excess steam to th7 condenser, and the safety valves open to exhaust steam to the atmosphere. '!he rise in steam pressure and the reduction in feedvater flow cause the average reactor coolant temperature to rise which reinforces the reactor power level de=and re-duction, already established by reduced =egawatt de=and, to restore reactor coolant temperature to set value.
As the turbine generator returns to set frequency, the turbine controls revert to steam pressure control rather than frequency control. This feature holds steam pressure within relatively narrow limits and prevents further large steam pressure chan6es which could impose additional load chan6es of opposite sign on the reactor coolant system. As a result, the reactor, the re-actor coolant system, and the steam system run back rapidly and smoothly to the new load level.
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l 73 DISTRLNENTATION 731 NUCLEAR USTRLNENTATION n e nuclear instrumentation system is shown in Figure 7-2A. .hphasis in I the design is placed upon accuracy, stability, and reliability. Instru-ments are redundant at every level. The des61 n criteria stated in 7 1.1.2 have been applied to the design of this instrumentation. ;
7 3 1.1 Design
~ l Be nuclear instrumentation has eight channels of neutron infcm ation l divided into three rac6es of sensitivity: source rs=6e, intermediate l range, and pcVer range. The three rsn6es ccmbine to give a centinuous
=casure=ent cf reacter pcver frcm source level to apprcximately 125 per i cent of full power or ten decades of infom ation. A =inimum of cne de-l cade of overlappin6 infor=ation is provided between successive h16her ranges of instrumentation. The relationship between ine,trument ranges is shown in Figure 7-9 i
The source ran6e instrumentation has two redn e nt count rate channels originating in two high sensitivity proportional counters. These channels areusedoveracountingrangeof1to107ccunts/secasdisplayedonthe operator's control console in terms of leg counting rate. The channels also measure the rate of chan6e of the neutren level as displayed for the operator in terms of startup rate frcm -1 to +10 decades /=in. No protec-O tive r==c*1c= = cci tea vita :s o=== = =8 * = =
instrumentation limitations encountered in this ra:6e.
or 1=s = ==
Ecvever, one inter-Icek is provided, i.e., a centrol red withdraw hold and alam en high startup rate in either channel.
The intemediate range instrumentation has two log N channels or161nating in two identical electrically gamma-ccmpensated ica chambers. Each chan-nel provides seven decades of flux level information in ter=s of los ion chamber current and startup rate. The ion chamber cutput range is frcm 10-11 to 10-4 amperes. The startup rate ran6e is frcm -1 to tlO decades per minute. Prot etive action on hich startup rate is provided by these channels. A high startup rate on eitner channel causes a reacter trip.
Prior to a reactor trip, high startup rate in either channel vd initiate a control rod withdraw hold interlock and alam.
The power ran6e channels have four 11 ear level channels originati=6 in 12 uncompensated icn chambers. m channel cutput is directly proportion-al to reactor power and covers the ran6e from 0 to 125 per cent of full pcver. The system is a precision analcg system which empicys a digital technique to provide highly accurate signals for instrument calibration and reacter trip set point calibration. The gain of each channel is ad-justable, providing a means for calibrating the catput a6ainst a reactor heat balanca. Protective actica en h16h flux level consists of reactor trip initiation by the pcver rs=ge channels at preset flux levels.
Additional features pertinent to the nuclear instrumentation system are as follevs:
v? r
- 0001 !l8
, 7-25
Independent power supplies are included in each channel.
- a. Pri-
=ary power originates from the vital busses described in l
8.2.2 7 Where applicable, isolation transfor=ers are provided to insure a stable, hi6h-quality power supply,
- b. The proportional counters used in the source ran6e are desi ned 6
to be secured when the flux level is Breater than their useful openting range. This is necessary to obtain prolonged operat-l in6 1 08. ,
t I c. The intermedi d .an6e "hannels are supplied with an adjustable source of gamma-co=pensating volta 6e.
t 7 3 1 1.1 Test and calibratica Test and calibration facilities are built into the system. The test fa-cilities vill meet the require =ents outlined in the discussion of protec-tion systems testing.
Facilities for calibration of the various channel smplifiers and measur-l ing equipment vill also be a part of the system.
l 7 3 1.1.2 Power Ran6e Detectors Twelve uncompensated ionization chambers are used in the power range chan-
! nels. Three chambers are associated with each channel, i.e., one near the bottom of the cee, a second at the midplane, and a third toward the top of the core. The cutputs of the three chambers are combined in their re-h spective linear amplifiers. A =eans is provided for readin6 the individ-ual chamber outputs sa a manual calibration and test function during nor-
=al operation.
7 3 1.1 3 Detector locations The physical locetrons of the neutron detectors are shown in Figure 7-10.
The power rangr. detectors are located in four primary positions, 90 de-grees apart around the reactor core.
The two source range proportional counters are located on opposite sides of the core adjacent to two of the power range detectors.
l The two intemediate range compensated ion cha=bers are also located on l opposite sides of the core, but rotated 90 degrees from the source range detectors.
7 3 1.2 Evaluation The nuclear instrumentation vill monitor the reactor over the 10 decade ra 6e from source to 125 per cent of full power. The full power neutron flux level at the power ran6e detectors vill be v.pprox1=ately 109 nv.
The detectors employed vill provide a linear response up to approx 1=ately 4 x 1010 ny before they are saturated.
pi ; u~ 119 0001 7-26
The intemediate ran6e channels overlap the source range and the pcver O range channels in an adequate =anner, providing the continuity of infor-mation needed during startup, ce axial and radial flux distributicn within the reactor core vill be measured by the incere neutron detectors (7 3 3). ne cut-of-core de-tectors are primarily for reactor safety, control, and operation info = a-tion.
7 3 1.2.1 Loss of Power The nuclear instraentatica draws its primarf pcVer frcm redundant battery-backed vital busses described in 8.2.2 7 7 3 1.2.2 Reliability and Ccmponent Failure The requirements established for the reactor protection system apply to the nuclear instraentation. All channel functions are independent of ever/ other channel, and where si6 nals are used for safety and control, electrical isolation is employed to =eet the criteria of 7 1.1.2.
7 3 1.2 3 Protection Requirements The relation of the pcVer ran6e channels to the Reactor Protection System has been described in 71. To maintain the desired accuracy in trip ac-tion, the total error frem drift in the power range channels vill be held O =e =1/2 > = :e== = r=11 >c = ove= 3o 4 7 verica-recalibration vill insure that this degree of deviation is not exceeded.
a===1=e : =
Bistable trip set points of the pcVer range channels vill also be held to an accuracy of t1/2 per cent of full pcVer. The accuracy and stability of the equip::ent will be verified by vender tests.
732 NONNUCLEAR PROCESS INSTRUMENTATION 7 3 2.1 System Design The nonnuclear instraentatica measures temperatures, pressures, flows, and levels in the reacter coolant system, steam system, and aux 111 arf re-actor systems. Precess variables required on a centinuous basis for the startup, operation, and shutdevn of the nuclear unit are indicated, re-corded, and centrolled frca the control recm. The quantity and types of process instraentatien provided vill insure safe and orderly operation of all systems and processes over the full operating ran6e of the Station.
The amounts and types of various instrments and centrollers shcvn are in-tended to be typical examples of those that vill be included in the vari-cus syste=s when final design details have been ec=pleted. The acnnuclear process instrumentation for the reacter ecolant is shewn in Figure 7-11 and en the auxiliarf reactor system drawings in Sections 5, 6, 9, and 11.
Precess variables are monitored as shown on the nennuclear instr.r.:entatien and auxiliar/ reacter system drawings and are as fcllevs:
G Q a. In general, resistance ele =ents are used for te=perature =ea-surements. Fast-response resistance elements =enitor the
,s, 0001 120
, -m >
reactor outlet te=perature. The outputs of these fast-respcase elements supply signals to the protective system.
g
- b. Pressures are =casured in the reactor coolant syste=, the steam system, and the auxiliary reactor systems. Pressure signals for high and low reactor coolant pressures and high reactor building pressure are provided to the protection systems.
- c. Reactor coolant pump =otor operation is monitored as an indica-tien of reacter coolant flow. This info = ation is fed to the reactor controls and reacter protection system. In addition, reactor coolant ficv signals are obtained and indicated by cen-tinuous measurement of the pressure drops across the reacter coolant side of each steam generater.
- d. Flev in the steam system is obtained through the use of cali-brated feedvater flev no::les. Flev infocation is utilized for control and protective functions in the steam system.
Steam generator level =easurements are provided for control and alarm functions.
- e. Pressurizer level is =easured by differential pressure transmit-ters calibrated to operatin6 temperature and pressure. The pressurizer level is a function of the reactor coolant system
=akeup and letdown flow rate. The letdevn flow rate is remote ma maM y centrolled to the required flow. Pressurizer level sigral: are processed in a level controller whose output posi-tiens the metaup centrol valve in the makeup line to maintain a constant level.
- f. Reactor coolant system pressure is maintained by a centrol sys-tem that energizes pressurizer electrical heaters in banks at preset pressure values belev 2,175 psig or actuates spray con-trol valves if the pressure increases to 2,230 psis.
7 3 2.2 system Evaluatica Redundant instraentation has been provided for all inputs to the protec-tien syste=s and vital centrol circuits.
l
! ktere vide process variable ran6es are required and precise control is in-volved, both vide-ran6e and carrev-ran6e instrumentation are provided.
htere possible, all instru=entation ecmpenents are selected frcm standard ec==ercially available products with preven operating reliability.
All electrical and electronic instruuentatica required for safe and reli-able operatica vill be supplied frem redundant vital a-c instr =entation busses.
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733 INCCRE MONITORING SYSTEM 7331 Design Basis The incere monitoring system provides neutron flux and temperature detec-tors to monitor core performance. No protective action or direct control functions are perfomed by this system. All high - nssure system connec-tions are teminated within the reactor building. Incore, self-powered neutron detectors measure the neutron flax in the core, and temperature detectors measure the core temperature differential to provide a history of power distributions and disturbances during power operating modes.
Data obtained will provide =easured power distribution infor=ation and fuel burnup data to assist in fuel management decisions.
7332 System Design 7 3 3 2.1 System Description The incere =onitoring' system. consists of a.ssemblies of self-powered neu-tron detectors, temperature detectors, and calibration tubes located at 52 preselected radial positions within the core. The incere monitoring locations are shown on Figure 7-12. In this arrangement, an incore de-tector asse=bly, consisting of seven local flux detectors, one back-ground detector, two (inlet and outlet) te:2perature detectors, and a cal-ibration tube, is installed in the instrumentation tube of each of 52 fuel assemblies (Figure 3 h8). The local cetectors are positioned at p
V seven different axial elevations to provide the axial flux gradient. The outputs of the local flux detectors are referenced to the background de-tector output so that the differential signti is a true measure of neu-tron flux. 'Ihe te=perature detectors, one located at the top of the fuel asse=bly and the other positioned just below the bottcm of the fuel as-se=bly, =easure the temperature difference t. cross the core to verify hot channel calculations and flow distribution in the core.
As shown in Figure 7-12, seventeen detector assemblies are located to act as sy= metr / monitors. The re=aining 35 dete: tor assemblies, plus five of the 17 sy==etry monitors, provide monitoring of ever/ type of fuel asse=-
bly in the core when quarter core'sy= metry er.ists.
Readout for the incere detectors is perfomed by the Station computer system rather than by individual indicators. This system sounes ala ms if local flux conditions exceed predetemined values.
When the reactor is depressurined, the incere detector assemblies can be inserted or withdrawn through guide tubes which originate at a shielded area in the reactor building as shown in Figure 7-13 These guide tubes, after completing two 90 degree turns, enter the bottem head of the reac-tor vessel where internal guides extend up to the instrumentation tubes of 52 selected fuel assemblies. The instrumentation tube then serves as the guide for the incere detector assembly. The incere detector assemblies are fully withdrawn only for replacement. Dur:.ng refueling operations, the inare detector assemblies are withdrawn approximately 13 feet to al-lov free transfer of the fuel assemblies. After the fuel asse=blier are placed in their nev locations, the incere detector assemblies are returned
. :q w
m . 0001' F22 as
to their fully inserted positions in the cere, and the hi6h pressure seals are secured.
g 7 3 3 2.2 Calibration Techniques The nature of the detectors pe mits the =anufacture of nearly identical detectors which will produce a hi6h relative accuracy between individual detectors. The detector signals must be compensated for burnup of the neutron sensitive material. The data handling system integrates each de-tector output current and generates a burnup correctica factor to be ap-plied to each detector signal before printing out the corrected signal in tems of per cent of full pcver. The data hanc. ling system ecmputes an average power value for tha entire core, comalized to the reactor heat balance. This average pcVer value is ccmpared to each neutron detector si6 nal to provide the core power distribution pattern.
7333 System Evaluation 73331 Operatin6 Experience The AECL has been operating incore, self-povered neutron detectors at l Chalk River since 1962. They have been successfully applied to both the NRX and NRU reactors and have been operated at fluxes beyond those ex-( pected in normal pressurized water reactor service.
73332 B&W Experience Self-pcVered, incore neutran detectors have been assembled and irradiated O -
in 'Ihe Babcock & Wilecx Cem,any Development Program that began in 1964.
Results frem this program have produced confidence that self-pavered de-tectors used in an incore instrument system for pressurized water reac-tors vill perform as well, if not better, than any system of incere in-strumentatica currently in use.
The B&W Development Program includes these tests:
- a. Parametric studies of the self-powered detector.
- b. Detector ability to withstand PWR envirec=ent.
l
- c. Multiple detector asse=bly irradiation tests.
- d. Background effects.
- e. Readout system tests.
j f. Mechanical withdrawal-insertion tests.
- g. Mechanical high pressure seal tests. .
- h. Relationship of flux measurement to power distribution experi- g ments. w
~
Et .* 0001 123 7-30
Preliminary conclusions drawn from the results of the test programs at the B E Lynchburg Pool Reactor, the B&W Test Reactor, and the B15 Rock Os Point Nuclear Power Plant are as follows:
- a. We detector sensitivity, resistivity, and te=perature effects are satisfactory for use,
- b. A multiple detector assembly can provide axial flux data in a single channel and can withstand reactor environment. An as-sembly of six local flux detectors, three background detectors, and two thermocouples has been successfully operating in the 1 Big Rock Point Reactor since May 1966.
- c. Data collection systems are successful as read-out systems for incere monitors.
- d. Background effects viu not prevent satisfactory operation in a PWR environment.
Irradiation of detector assemblies and evaluation of performance data are continuing to provide detailed design infomation for the incere instru-
=entation system.
7.4 OPERATING CONTROL STATICIS Folleving proven power station design philosophy, all control stations, switches, controllers, and indicators necessary to start up, operate, and O
shut down the nuclear unit vill be located in one control recm. Control functions necessary to maintain sefe conditions after a less-of-coolant accident vill be initiated from the centrally located control room. Con-trols for certain auxiliary systems may be located at re=ote control sta-tions when the system controH ed does not involve power generation control or emergency functions.
7.4.1 GENERAL LAYCUT Ee control roem win be designed so that one =an can supervise operation of the Station during nomal steady-state conditions. During other than nc=al operating conditions, other operators vill be available to assist the control operator. De control reos vin be arranged to include an operating console to house frequently used and emergency indicators and controllers at close proximity and visibility to the operator. Vertical panel boards vin house less frequently used controllers and infomatioral displays. De console vin be fomed by short, straight sectic:.s of bench boards arranged to form a semicircular operati::g console that aHovs the operator easy accessibility to each section.
7.4.2 INFCRMATION DISPLAY AND CONTROL FUNCTION R e necessary info mation for routine =enitoring of the nuclear unit and the Station viu 'ce displayed on the control reem console or on visible pv panel boards in the immediate vicinity of the operator. Info::ation dis-play and control equipment frequently employed on a routine basis, or pro-tective equipment quickly needed in case of an emergency, vill be =ounted u'
i 0001 124-7-n
1 e l l
l cn the operating censole. Recorders and radiatic: =cnitoring equipment I vill be =cunted on vertical panels in the control rec =. Infrequently used equipment, such as indicators and centrollers used pri=arily during jg I startup or shutdown, vill be =cunted on adjacent side panel boards. !
l The operating censole vill be horseshee in shapa with the inclined bench-board surface-=cunted with controllers and cc=binatica centroller-indica-ters. Behind each section of benchboard vill be located vertical boards having indicators and recorders associated with the respective section of ;
benchboard; those ite=s necessary for intelligent operation vill be
=ounted high en the vertical board in full view of the operator. These not essential to continual scrutiny vill be =cuated lever en the vertical boards . The two center sections of the operator's console vill house the
= ore i=portant operating centrols.
A Station ec=puter vill be available in the control roc = for ala n =cni-toring, perfor-a-ce =cnitoring, and data legging. On-de=and printout is available to the cperator at his discretien in addition to the cc=puter periodic legging of the, Station variables.
7.k.3
SUMMARY
OF ALARMS l
Visible and audible alars units vill be incorporated into the control roc = to warn the operator if unsafe conditions are approached by any syste=. Audible reacter building evacuation alar =s are to be initiated frc= the radiation =cnitoring syste= and frc= the source range nuclear instru=entation. Audible alar =s vill be sounded in appropriate areas throughout the Station if high radiation conditions are present. llh /
T.h.4 CCMMUNICATION Station telephone and paging syste=s vill be provided with redundant pcVer supplies to provide the control rec = operator with constant ec=-
=unication with all areas of the Station. Accustical phones will be supplied in areas where the background noise level is high. Cc==unica-tion cutside the Station vill be through the full period leased lines of the Bell Telephone Cc=pany of Pennsylvania and the Metropolitan Edison
=cbile radio system.
7.k.5 OCCUPANCY Safe occupancy of the control rec = during abner =al conditions vill be provided for the design of the centrol rec =. Adequate shielding vill be used to =aintain tolerable radiation levels in the control roc = for
=axi=u= hypothetical accident conditions. The centrol rec = ventilation syste= vill be provided with radiation detectors and appropriate alar:s.
Provisions vill be =ade for the centrol rec = air to be recirculated through EEPA and charcoal filters. 2:ergency lighting vill be provided.
The potential =agnitude of a fire in the control rec = vill be li=1ted by the folleving facters:
- a. Materials used in the centrol roc = construction vill be non-flm=nable.
lh e sib 0001 125
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7-32 (Revised 7-21-67). So
. i, n '
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- b. Centrol cacles and switchbcard viring vill be constructed of Mterials that have passed the flame test as described in In-sulated Pcver Cable Ingineers Association Fue.Licatien S-61-h02 and National Z1ectrical Manufacturers Association Publica:ica
- 4C 5-1961.
- c. Furniture used in the con *r:1 rec = will be of =e*al ecnstruc-tien.
- d. Cc=bustible supplies such as legs, records, precedures, =anuals, etc., vill be limited to the a= cunts required. for Station ep- ;
eration.
- e. All areas of the control rec = will be readily accessible for
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- f. Adequate fire extinguishers vill be previded.
- g. The centrol rec = vill be occupied at all ti=es by a qualified person who has been trained in fire extinguishing techniques.
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- a. Paper in the #cr= cf icgs, records, procedures, =anuals, dia-gra=s, etc.
- b. The coaxial cables required for nuclear instrumentation.
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- c. Small a: cunts cf ec=bustible =aterials used in the =anufacture of varicus electrcnic equipment.
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7.k.T SAFITY FIATURIS Pri=ary objectives in the control rocs layout are to provide the nec-essary controls to start, operate, and shut down the nuclear unit with l
i sufficient information display and alars =cnitoring to instre safe and reliable operation under normal and accident conditions. !pecial e=-
phasis will be given to maintaining control integrity durits accident conditions. The layout of the engineered safeguards secticn of the con-l trol board vill be designed to sinimize the time required :'or the oper-ator to evaluate the system performance under accident conditions. Any deviations fres predeter=ined conditions vill be alarmed so tnat cor-rective action =ay be taken by the operator using controls provided en the centrol panel.
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. 0001 127 y4 p. 7-3 (Revn ed 7-n - m m
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O CHAIGEL I CHARNEL CHANNEL CHANNEL i
1 2 3 4 l
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~ _ l HIGH REACTOR OUTLET TEMP.
=
l HIGH REACTOR COOLANT PRESSURE "0R" l "OR" "0R" "0R" LOW REACTOR FOR FOR FOR FOR COOLANT PRESSURE - MIP TRIP l TRIP TRIP HIGH REACTOR START- l UP RATE (below 10 per
[
cent rated power)
LOSS OF REACTOR '
COOLANT PUMPS l I
I INP17TS TYPICAL OF AE FOUR CHANNELS l l
O BISTABIE l BISTABLE BISTABLE BISTABLE I
I A B C D
___________J 2/h 2/4 CODCIDENCE COINCIDE CE 4
ROD DRIVE ROD DRIVE POWER SOURCE NO. 1 POWER SOURCE NO. ;
BREAXERS BREAJGRS O
REACTOR PROTECTION SYSTEM Bloi:K OlJ NM FICURE 7-THREE MILE ISLAND NUCL f T10
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