ML19094A134
| ML19094A134 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/13/1970 |
| From: | Virginia Electric & Power Co (VEPCO) |
| To: | US Atomic Energy Commission (AEC) |
| References | |
| Download: ML19094A134 (686) | |
Text
{{#Wiki_filter:. 1' *~'---',* PART B - VOLUME 5 Vepco* SURRY POWER STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT _.,, VIRGINIA ELECTRIC AND POWER COMPANY
i 2-13-70 TABLE OF CONTENTS OF THE FINAL SAFETY ANALYSIS REPORT 1 INTRODUCTION AND
SUMMARY
1.1 INTRODUCTION
- 1. 2
SUMMARY
- 1. 3 COMPARISON WITH OTHER STATIONS 1.4 COMPLIANCE WITH CRITERIA
- 1. 5 COMMON AND SEPARATE FACILITIES 1.6 RESEARCH AND DEVELOPMENT
- -** 2.1 2
GENERAL DESCRIPTION SITE 2.2 METEOROLOGY AND CLIMATALOGY 2.3 HYDROLOGY 2.4 GEOLOGY 2.5 SEISMOLOGY 3 REACTOR 3.1 GENERAL DESCRIPTION 3.2 DESIGN BASES 3.3 NUCLEAR DESIGN 3.4 THERMAL AND HYDRAULIC DESIGN AND EVALUATION 3.5 MECHANICAL ~ESIGN 3.6 TESTS AND INSPECTIONS
ii 2-13-70 4.1 4 DESIGN BASES REACTOR COOLANT SYSTEM 4.2 SYSTEM DESIGN AND OPERATION 4.3 SYSTEM DESIGN EVALUATION 4.4 TESTS AND INSPECTIONS 5 CONTAINMENT SYSTEM 5.1 GENERAL DESCRIPTION 5.2 CONTAINMENT ISOLATION 5.3 CONTAINMENT SYSTEMS 5.4 DESIGN EVALUATION 5.5 TESTS AND INSPECTIONS 6 ENGINEERED SAFEGUARDS 6.1 GENERAL DESCRIPTION 6.2 SAFETY INJECTION SYSTEM 6.3 CONSEQUENCE-LIMITING SAFEGUARDS 7 INSTRUMENTATION AND CONTROL 7.1 GENERAL DESIGN CRITERIA 7.2 PROTECTIVE SYSTEMS 7.3 CONTROL SYSTEMS 7.4 NUCLEAR INSTRUMENTATION SYSTEMS DESIGN AND EVALUATION 7.5 ENGINEERED SAFEGUARDS INSTRUMENTATION 7.6 INCORE INSTRUMENTATION 7.7 OPERATING CONTROL STATIONS 7.8 AUTOMATIC LOAD CONTROL 7.9 COMPUTER
iii 2-13-70
- 8.1 8 ELECTRICAL SYSTEMS GENERAL DESCRIPTION AND
SUMMARY
8.2 DESIGN BASES 8.3 UTILITY SYSTEM INTERCONNECTIONS 8.4 STATION SERVICE SYSTEMS 8.5 EMERGENCY POWER SYSTEM 8.6 TESTS AND INSPECTIONS 9 AUXILIARY AND EMERGENCY SYSTEMS 9.1 CHEMICAL AND VOLUME CONTROL SYSTEM 9.2 BORON RECOVERY SYSTEM
- - 9.3 9.4 9.5 RESIDUAL HEAT REMOVAL SYSTEM COMPONENT COOLING SYSTEM FUEL PIT COOLING SYSTEM 9.6 SAMPLING SYSTEM 9.7 VENT AND DRAIN SYSTEM 9.8 COMPRESSED AIR SYSTEMS 9.9 SERVICE WATER SYSTEM 9.10 FIRE PROTECTION SYSTEM 9 .11 WATER SUPPLY AND TREATMENT SYSTEMS 9.12 FUEL HANDLING SYSTEM 9.13 AUXILIARY VENTILATION SYSTEM 9.14 DECONTAMINATION FACILITY
iv 2-13-70 10 STEAM AND POWER CONVERSION 10.1 GENERAL DESCRIPTION
- 10. 2 DESIGN BASES 10.3 SYSTEM DESIGN AND OPERATION 11 RADIOACTIVE WASTES AND RADIATION PROTECTION 11.1 GENERAL DESCRIPTION
- 11. 2 RADIOACTIVE WASTE SYSTEMS 11.3 RADIATION PROTECTION 12 CONDUCT OF OPERATIONS 12.1 GENERAL 12.2 12.3 12.4 ORGANIZATION TRAINING SHIFT PERSONNEL 12.5 HEALTH PHYSICS 12.6 OPERATIONS PROCEDURES 12.7 RECORDS 12.8 REVIEW AND AUDIT OF OPERATIONS 12.9 INSERVICE INSPECTION 13 INITIAL TESTS AND OPERATION 13.1 TESTS PRIOR TO INITIAL REACTOR FUELING 13.2 FINAL STATION PREPARATION 13.3 INITIAL TESTING IN THE OPERATING REACTOR 13.4 OPERATING RESTRICTIONS
V 2-13-70
- 14.1 GENERAL 14 SAFETY ANALYSIS 14.2 CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS 14.3 STANDBY SAFEGUARDS ANALYSIS 14.4 GENERAL STATION ACCIDENT ANALYSIS 14.5 LOSS-OF-COOLANT ACCIDENT 15 STRUCTURES AND CONSTRUCTION 15.1 STRUCTURES AND MACHINERY ARRANGEMENT 15.2 STRUCTURAL DESIGN CRITERIA 15.3 MATERIAL 15.4 CONSTRUCTION PROCEDURES AND PRACTICES
- 15.5 15.6 SPECIFIC STRUCTURAL DESIGNS OTHER CLASS I STRUCTURES APPENDIX A - REPORT, SITE ENVIRONMENTAL STUDIES, SURRY POWER STATION APPENDIX B - SEISMIC DESIGN FOR THE NUCLEAR STEAM SUPPLY SYSTEM TECHNICAL SPECIFICATIONS 1.0 DEFINITIONS 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 5.0 DESIGN FEATURES 6.0 ADMINISTRATIVE CONTROLS
15-i 12~1-69 Section 15 TABLE OF CONTENTS Title STRUCTURES AND. CONSTRUCTION 15.1-1 15.1 STRUCTURES AND MACHINERY AAAANGEMENT 15.1-1 15.2 STRUCTURAI,, DESIGN CRI'rERIA 15.2-1 15.2.1 GENERAL 15.2-1 15.2.2 NORMAL WIND LOADING 15.2-13 15.2.3 TORNADO CRITERIA 15.2-13 15.2.4 SEISMIC DESIGN 15._2-16, 15.2.5 HYDROSTA'l?IC LOADINGS 15.2..:20 15.2.6 LATERAL EARTH l?RESSURE 15.2-20 15.3 MATERIAL 15.3-1 15.3.1 CONCRETE 15.3-i
- 15.3.2 15.4 15.4.1 REINFORCING STEEL CONSTRUCTION PROCEDURES AND PRACTICES CODES OF PRACTICE 15.3-3 15.4-1 15.4-1 15.4.2 CONCRETE .15.4-3 15.4.3 REINFORCING STEEL 15.4-6
- 15. 4. 4 . CONSTRUCTION PROCEDURES 15.4-8 15.4.5 CONSTRUCTION PRACTICE 15.4-9 15.4.6 QUALITY ASSURANCE PROGRAM 15.4.6-1 15.4.6.1 General 15.4.6-1 15.4.6.2 Organization 15.4.6-3 15.4.6.3 Virginia Electric-and Power Company 15.4.6-5 Quality Assurance Program 15.4.6.4 Stone and Webster Quality Assurance 15.4.6-10 Program 15.4.6.5 Westinghouse Quality Assurance Program 15.4.6-30
15-ii 12-1-69 Section Title 15.5 SPECIFIC STRUCTURAL DESIGNS 15.5.1-1 15.5.1 CONTAINMENT STRUCTURE 15.5.1-1 15 .5 .1.1 General 15.5.1-1 15.5.1.2 Design Criteria 15.5.1.2-1 15 .5 .1. 3 Bouyant Loads 15.5.1.3-1 15 .5 .1.4 Dynamic Analysis 15.5.1.4-1 15.5.*l.5 Static Analysis 15.5.1.5-1 15 .5 .1. 6 Reinforcing Steel Arrangement 15 ;5 .1. 6-1 15.5.1.7 Penetration Design 15.5.1. 7-1 15.5.1.8 Steel Liner and Penetrations 15 .5 .1. 8-1 15 .5 .1.9 Materials 15 .5 .1. 9-1 15 .5 .1.10 Construction Procedures and Practices 15 .5 .1.10-1 15 .5 .1.11 15 .5 .1.12 15.5.1.13 Missile and Piping Rupture Ground Water Protection and Corrosion Testing and Inservice Surveillance 15.5.1.11-1 15.5.1.12-1 15.5.1.13-1 15.6 OTHER CLASS I STRUCTURES 15.6-1 15.6.1 OTHER STRUCTURES 15.6-2 15.6.2 REACTOR COOLANT SYSTEM SUPPORTS 15.6-3 15.6.3 CONTAINMENT INTERNAL STRUCtURE 15.6-6
15 .1-1 12-1-69 15.1 STRUCTURES AND MACHINERY ARRANGEMENT The site arrangement, plot plan, and the general arrangement of equipment within the principal Class I structures are shown on the figures .listed in the following tabulation: Item Figure Site Plan 15.1-1 Plot Plan 15.1-2 Containment Structure and 15.1-3, -4, -5, -6, -7, -8, Containment Auxiliary -9 and -10 Structures Auxiliary Building 15.1-11, -12 and -13
. Fuel Building 15.1-14 and -15
- Control Area -15.1-16
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. K OCT.15, 1970 LADDER
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NOTES1 SCALE - ~*. l'-0 1' AREA<, ~HO'IIN 'TIIU5' iJ&:l ARit. QEMOV4e!J< 51.Ad:; SET ON Gli.ATING ANO TOBE LEF'TOFF" 0Ulil:IN6 MD1!"1ALcPEii.ATION AUX BLDG TO PE!OVIDE PRESSURE EEL!Ef' AND \IEIITILATION Sllloj:.2, 8 SLABS Ali!E TO 6E 51t>RECQIITHE 18'4.LE:'IEL INntE NClffll*
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WEST ','!UADAAl"T. MAINTAINANCEOR -ua;-. AREA SHOWN 'TMUSlaEi -&: 11EMO#ll!IL£ $LASS FOR.*. ALL 6RA'TIN& AT El..47"4TO BE ltEi"!OVAlilLE, PROVIDE 12.EMOVABl.£ GI.W2D i;:AIL Al!OUIC> eJMTY AND H015TING SPACE:, SUPPClT FOR ~TORA.GE OF
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OCT. 15, 1970
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_(FM-IB) (FM-IA) SCi'I\LE~ J'8 z \'* d' VERTICAL SECTION-SH 3 R E.F'ER ENCE. DWGS & 6EN. NOTE_S -~M*I_A SURRY POWER STATION
FIG.15.1-10~ OCT.15,1970,
\ ~It .COIJTAIIJMEUT , ( UIJIT !JO.I)
I M
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VL VEAT1CAL LAOOEA: . II I METAL HEATIIJG ( VEIJTILATIOIJ EQUIP l?O.l22 COUG~EH: BLOCK WALL (REMOVABLE) ELEVATOR MrTl~G(VEI.ITl\lf TIOIJ EQUIP 51DUJG (~B- 58) REFEREtJC'E: D~AWI-IJ6!>: I A'RQGT AUX.BLDG-SHEET 2 * (FM*5B)
.<< ARRGT AUX BLOG-SHEETS - ( FM*5C)
- COMP C.OOL !!l ARRGT AUX BLDG-SHEET 'I ( OM-,o) 5U~~E TALIK l*CD*E*IA AUXILIARY BUILDING
- A-A B-B ARRANGEMENT SHEET I SURRY POWER STATION
FIG.15.1-11 OCT.15,1970 B<;i-lt'-J.:!-8171711 :~
!1___
N I 1-~-"'""""'" Q U\.rJIT NO.l L K . LE.AKAG,( MONITORING SYSTEM. .
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EVAPORATOI( BOTTOMS TK l*BR*P*5A 4 REMOVABLE 61..ABS FOR , SPEUT J:ILTEI.' TRA~S.i:ER CA':lk (1VP FOR 5)
¢: COIJTAMII.JATED D~A.1 U 5 TAUIG NOTES:
scAt.E.: K :J'-o"
=~\~~~~E :1i~~D7N~E~~~~~g~~7.ES "\FM-~
F e, - e.AR'T'TES ;
- CONCRtit E # t: MIN.IMUM ToRt>JAt>o Ml'lslLE ..tn1c*1,c.1,tESs E VL 'fE.R.T\CAL'. ~ACl>EI<.. .
I El!IEV l ~.PENTHOUSE M/1.C!-4 l SPACE.: ' G H I HCM"ING I< VENTILAJ,NG £QUIP I (i'B-55) I VE~T bU,CT EL .~s'-ro" I
- --~
't; t'L !1'*0' COM? *COOL---'--H-,Jt,11-,---,.C--l SURGE T.t..Nk t*CC:*'TK*I E-L 'l.7'-c.," D-D L.sERv1cE:
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*\*BR-I*1B AUXILIARY BUILDING ARRANGEMENT SHEET 2 SURRY POWER STATION
*~ FIG.15.1-12 OCT. 15,1970
' I H I L
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REFERENCE:
DRAWINGS t GE.N. NOTES -FM-SA I I VOLUME~ l*CV*EJ*I com TANK ~ i "" MINIMUM ,SHIELOlNG 1'1-HCKNESS
*¥ MINIMUM TORNAbo Ml'5S\\..'E. 'THICKNESS VL VE1<11CAL LADDER .
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FIG.15.1-13 . 1 1 FU EL 8LDEl OCT.15 1970 oc;-V"JJ-8vv11 F\l[L BLDG EXH FAN.";.\ I-VS F*7A
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- 1'-0" . . ,
1 l*RL*P*1Al'PU~IF l'A,TIOIJ REFERENCE, DRAWINGS,t GEN- NC)T:ES *,, 1/M~~A. 1-QL*P*I~ .PUMPS *MINIMUM :Sri1£.LDlNG; THl~KN£5~* . 5UMP ~olt'-MININ\UM TO.R~AOO 1M\S$\Lf. TttkK~EiS PU~ E VL VERTICAL LA.DOER 9 9~ 10 H-H 100 1-0'
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FIG.15.1-16 L OCT. 15, 1970 AUX ILIA?.'/ BU! L D iKJ.U EX?, JT, Fl R.=. WALL
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1 VAULT CORRIDOR
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CE.ll\t-C.1-\T.. LO'-cf *:. REFLECTED CEILING PLAI\J CONTROL RM. OFFICE ' 5CA1..E:: ~8-* r'*O" TH[S P0?,;"!01~ OF WA.LL1 r-- I 1\1STRUIAEN'7' Rl'.?A!R CLEAN TO BE. C.l:IM?:...E:TCD 1\r-'"*I SHOP WASHROOM T::~ 1:::iu1?, INS"!'.i..:..:....~-L* i,.
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1*~*.* 'a C!ElLINGt.allSP!W.STOt-l SYSIEM TO B~ S1Ml~ 1C TYFl:"R'.'A~ MFRD. ev E-A~i&RN.PR'Ot'UeTS" CORP., BA.LIIMORE,MD.; . BAiTt~Y l?.OOM RO~ SLAB "°L~\CIN CONC. BLOC!£. , .. G-NCLOf>U2.E-'S-£L.~AT~..:.:-* * .. *,
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.SURRY POWER STATION
Fl G. 15. 1-15 OCT.15,1970 '** 86-V\J.:J-8vv11 Uu u, sa8 NE.\N FUEL CRf>..NE PAYOUT Tr/AV I
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l*FM-2. UNOERHUNG BE'AM (IN TRANSFER PO~ITJON) 6 1 1:MERWATE.R DECONTAM BLOG (FM-IOB) D F-F i MN .Q R e l.-f------------"FU~EL~8~L:..'0~Gi_==::::j:::::::*~u"=Ec'C(Jl\'T)(1'\~*~ "BLDG t FUEL TAANSFER CANAL
. , UNIT '2.
lOTON ..<<llST R. S. DOOR L.. PLAN EL 27'- r;;* --1 TOP'OF CRANE RAIL EL81'-,4! C IJ E
,:1 V.L...
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~
T.O.S.EJ:.44'-B" AUX &DO TOP OF- FUEL UPENOEA (FM*!!,A) EL 2.'2 1*8" / I I(becoNT'AM (Ft,HOC) BLDG n lJ
- c-c
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EL1'*1' "-'---'-'-- I FUEL BUILDING E - E ARRANGEMENT.SHEET 2 A-A SURRY POWER STATION
FIG. 15. 1-14 OCT. 15, 1970 V6-V'J ..:l- 8vv 11 !=UEL BLDG 5UMPPUMP!, ARRGT*NORTH VAAD (S"*IOA) E
,o'lfl II~ Q ~~~~ ,~*c:r~J!o~* N
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/ PAVOUTTRAV (ELE.C.TAICAL)
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--1 E
(FM-96)
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- 9'-c'".
EL '21'-1;._ 1------ - -SLOTS FOR GA.TE
- 3. \,lOR\.A/\L 'NA.IE.'R \.'i:.\/EL IN SPENT FUEL PIT TO BC MA.lNTA1NED AT LL 45'-4.".
'4. UPPER SE.C.TIOk Of Gilt.TE ~.... i EXTEMOED JF (FM*98) 5PeNT F-Ufl A55 1Y STEM TO BE IU.t<<WAM.E. TO Al.I.OW PAs.e.A.C:iE OF MOTOR DR\\IEN PL.ill.TFORM,
.5. V. L.-11JDJCATES VERTICAL LADDER.
-+-------*-
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~ ~MIT OF HOOK AUXILIARY BLOC,
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FUEL BUILDING ARRANGEMEN~SHEETI SURRY POWER STATION
15.2-1 12-1-69
- 15.2 STRUCTURAL DESIGN CRITERIA 15.2.1 GENERAL The structures, systems, and components of the Surry Power Station, Units No. 1 and No. 2 are classified into groupings requi~ing seismic, tornado or conventional design. J Class I design encompasses those structures, systems or components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to the mitigation of their consequences *
- 'Structures, systems and components are designed fabricat~d and constructed to*
performance standards that will enable the facility to**:withstan.d, without loss of capability to protect the public, the additional forces ~li~t-,'might be imposed by:
- 1. Operational Basis Earthquake and the Design Basis Earthquake.
- 2. Tornadoes and other local site effects including flooding conditions, winds and ice. Radiation levels which constitute a hazard to the public are defined in 10CFRlOO.
The Class I structure is designed for r,esistance to seismic loadings in accordance with Section 15.2.4 and for tornado, where applicable, in accordance with Section 15.2.3. There are some structures, ~ystems, or components whose loss or failure by earthquake will not affect the public health or safety and will permit
.:)..5 .2-2 12-1-69 safe station shutdown, although their loss could interrupt power generation.
These structures, systems, or components are not designed tor specific'seismic or tornado loadings. Structures not designed for seismic or tornado loadings are designed according to "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," Aisc..:..1963 "Building Code Requirements for Reinforced Concrete" AC! 318-63, Part IVA - Working Stress Design. These structures are designed for dead, live and normal wind loads using allowable stress levels given in the above codes. Some structures, systems and components of the station are necessary for a safe and orderly shutdown during a tornado. These structures are designed for tornado loadings, and systems and components are protected by tornado resistant structures *. ~ A list of the structures, systems, and components designed to satisfy seismic and/ or tornado criteria is given in Table 15.2.1-1.
- e 15.2-3 2-13-70 TABLE 15.2.1-1 STRUC11JRES, SYSTEMS AND COMPONENTS WHICH ARE DESIGNED FOR SEISMIC AND TORNADO CRITERIA LEGEND W - Westinghouse Electric Corporaticr:.
SW - Stone & Webster Engineering Corporation I - Refers to Class I seismic criteria. All Class I components and structures are designed to resist the operational basis earthquake within allowable working stresses. A check has been made to determine that failure to function will not occur with a design basis earthquake. T - Refers to structur,es, systems and components which will not fail during. the design tornado. P - Refers to systems and components which will not fail during the design tornado since they are assumed to be protected by tornado resistant structures NA-' Not applicable
15.2-4 12-1-69 TABLE 15.2.1-1 (Continued) Earthquake Tornado* Criterion Criterion Sponsor Notes Item STRUCTURES SW Reactor Containment I p Reinforced Concrete Substructure Reinforced Concrete Superstructure I T Reinforced Concrete Interior Shields and Walls I NA p p for containment integrity. Steel Plate Liner I T for shield wall and critical Piping, Duct, and Electrical I T system penetrations only Penetrations and Shield Wall p for containment integrity I p Personnel Access Hatch p for containment integrity I p Equipment Access Hatch Cable Vault and Cable Tunnel I T SW Pipe Tunnel to Containment from Auxiliary Building I T SW Auxiliary Steam Generator Feed Pump Cubicle I T SW Cubicle for Main Steam and Feedwater Isolation Valves I T SW Recirculation Spray and Low Head Safety Injection Pump Cubicle and.Pipe Tunnel I p SW Safeguards Ventilation Room I NA SW Auxiliary Building SW Reinforced Concrete Structure I T Steel Superstructure I NA Vacuum Equipment Area I NA Fuel Building SW Reinforced Concrete Structure I T T for horizontal missile only Steel Superstructure I T T for tornado winds only I p p for horizontal missile only Spent Fuel Storage Rack Fuel Handling Trolley Support Structure I p T for tornado winds only Control Room . I T SW Emergency Switchgear and Relay Room I T SW Battery Rooms I T SW Air Conditioning Equipment Room I T SW For control room and relay room only Reactor Trip Breaker Cubicle I T SW
.I 15,2-5 2-13-70 TABLE 15,2.1-1 (Continued)
Earthquake Tornado Item Criterion Criterion Sponsor Notes STRUCTURES (CONT'D) Auxiliary Diesel Generator Cubicles SW Reinforced Concrete Floor I T Walls, EAcluding Louvers I T Structural ~,e~l Supported Roof and Roof Slab I T Louvers* T Protected by miss.ile rack Turbine Building NA NA SW By design, building collapse will not damage any Class I structures and components during earthquake or tornado resistant structures and components during tornado Circulating Water Pump *intake Structure I T SW T for emergency service water pump cubicle only High Level Intake -Structures I T SW T, no missile protection required Seal Pits* I T SW T, no missile protection required High Level Intake Canal I NA SW Fire Pump House* I T SW Engine driven pump only
~*uel Oil Transfer Pump Vault I T SW B~r;;:.\ Recovery Tank-Dikes I T SW
15.2-6 12-1-69 TABLE 15,2,1-1 (Continued) Earthquake Tornado Criterion Sponsor Notes Item Criterion SYSTEMS Reactor Coolant System I p w Steam Generators p SW Steam Generator Supports I I p w Reactor Coolant Pumps* p SW Reactor Coolant Pump Supports I Pressurizer and Pressurizer Heaters I p w I p SW Pressurizer Support p w Pressurizer Relief Tank I Reactor Vessel p w Reactor Core Support Structure I Reactor Control Rod Guide Structure I p w I p w Fuel Assemblies p w Control Rod and Drive Shaft Assemblies I I p w In-Core Instrumentation Thimbles Reactor Vessel Supports and Neutron p SW Shield Tank I Control Rod Drive Mechanisms I p w Reactor Coolant Piping, Valves and p w Supports** I Reactor Coolant Bypass Piping, Valves p w and Supports I I p w Pressurizer Surge Line Pressurizer Spray Lines, Valves and I p SW Supports p w Pressurizer Safety and Relief Valves I Notes: *All references to "Pumps" include drivers
**All references to "Piping and Valves" include root valves connecting_ to non-Class I systems, and valve operators
- e 15.2-7 12-1-69
- TABLE 15~2.l-l (Continued)
Earthquake Tornado item Criterion Criterion. Sponsor Notes SYSTEMS (CONT'D)
** .safety Injection* System
*Accumulators and* Supports I NA w Low Head Safety Injection Pumps and
'Piping .* I p w P for containment integrity Boric Acid Injection Tanks and Piping I p w*
All Other Piping, Valves and Supports I NA SW Except drain/sample lines Containment Spray System Refueling Water Storage Tank I NA SW Assume full of 'water Containment*Spray Pumps I NA SW
*All Piping,* Valves and Supports I NA SW Except recirculation lines Refueling .Water Chemical Addition Tank I NA SW Assume ful,l .Pf, ..~at,er ,
Recirculation- Spray Systems Recirculation*. Spray: Pumps *and Piping I p SW P for containment integrity Recirculation Spray Heat Exchangers Reactor Containment Sump and Screens I I. NA NA SW SW All Other'iiping; Valves and Supports I NA SW Containment*Vacuum System Vacuum.Pumps I NA SW Process Vent I NA SW Vacuum Punij, Piping, Valves and Supports I NA SW
. Chemical and Volume Control *system Boric Acid Tanks I NA w
*Boric Acid Transfer Pumps I p w Boric Acid Blender: I p w Charging/Safety Injection Pumps I p w
.Regenerative Heat Exchanger I p w Nonregenerative Heat Exchanger I p w Mixed Bed *Demineralizers *I p w Reactor Coolant Filter I p w Volume Control Tank I p w Seal Water Heat Exchanger I p w Seal Water Filter I .P w Excess Letdown Heat Exchanger I p w Piping, Valves and Supports Boric Acid Piping I p SW Feed and Bleed Piping I p SW Hydrogen, Nitrogen and Vent Piping for Volume Control Tank I p SW
15.2-8 12-1-69 TABLE 15.2.1-1 (Continued) Earthquake Tornado Item Criterion Criterion Sponsor Nol:es SYSTEMS (CONT'D) Residual Heat .. Remova1*system Residual Heat Removal Pumps I p w Residual Heat Exchanger~ I p w All Piping*, Valves .and Supports.** I p SW Boron Recov~ry System Gas Stripper I p SW Gas Stripper Overhead.Condenser I p SW Primary.Drain Tank I p SW Component Cooling.System Component Cooling Pumps I p SW Component Cooling Water* Heat Exchangers I p SW Component Cooling Surge Tank I p SW Piping, -Valves and Supports ' For Residual Heat Exchangers I p SW For Fuel Pit Coolers I p SW P for horizontal mis.sill! only Fuel Pit Cooling System Fuel Pit Pumps I p SW p for horizontal missile only Fuel Pit Coolers**.* I p SW p for horizontal missile only Piping, Valves and Supports Connecting *'. Above Equipment to 'Spent Fuel -Pit I p SW p for horizontal missile only Compressed Air System Instrument.Air Compressors I p SW Instrument Air Receivers and Driers I p SW Containment Instrument *Air Compressors I p SW Containment I_nstrument Air Receivers and Driers. I p SW Piping, Valves and Supports to Critical Instruments and Controls l* p SW
-- 15.2-9 12-1-69 TABLE 15.2.1-1 (Continued}
Ec!rthquake Tornado Criterion Criterion Sponsor Notes SYSTEMS, (CONT'D) . Servi'ce Water System Engine Driven Emergency Service Water Pumps T SW Charging Pump Oil Coolers Service Water
*Pumps , I p SW All Service Water Piping, Valves and
,Pipe Supports :for: . .
Recirculation-spray Heat Exchangers I NA SW Component Cooling Heat Exchangers 'I p SW
, Engin,i* Driven E11ergency Service Water Pump . . , I T SW Auxiliary Generator* Cooling* I p SW Control RQom Air-Conditioning Equipment Condensers I p SW Charging Pwiip.Lube Oil Coolers I p SW Diesel*. Oii Tank for ,Emergency Service Wa.ter Puap I p SW Fire-Protection System Engine Driven' Fire Pump I p SW Diesel Oil Tank (300 gal} I p SW Emergency Water Supply Piping, Valves and Supports fr011 High*Level *rntake Canal I T SW Make provisions for future piping ooly Yard.Hy~ragt Pipi~~ Sy~te~ I P, SW Fuel Handlin'? Syst_~
- Manipulator Crmie *in Containment I w Crane will be parked and secured so no damage to reactor control rod drive mechanisms can occur during earthquake Movable.Platform*with Hoist in Fuel Building I NA SW Platform will be parked and secured so no damage to fuel can occur during earthquake* or tornado Fuel Handling Trolley in Fuel Building I NA SW Trolley will be parked during tornado warning periods and secured during torn,Jdo w:irning periods so no damage to spent fuel can occur Fuel Transfer Tube with Blind Flange I p SW P for containment isolation Fuel. Elevator in Fuel Building I NA SW
l 15.l--lO 12- l--,,'.) Earthquake Tornado Item Criterion Criterion Sponsor liotes SYSTEMS (CONT'D) Ventilation System* I ** Ventilation Equipment for Safeguards Ventilation Room I NA SW Duct Work fr0111 Safeguards Ventilation
-* Room to Ventilation Vent I NA SW Duct Work for Containment Purge System Penetrating* Containment Between and Including Isolation Butterfly Valves I p SW P for containment isolation Air-Conditioning Equipment for Main Control Room and Relay Room. I p SW Emergency Main Control and Relay Room Ventilation
- I p SW Ventilation Vent I NA SW Secure to earthquake resistant section of turbine building Control Rod Drive Ventilaton Fans NA T SW Main Steam *system Steam Piping.fr011 Main Steam *Lines to Auxiliary* Steam.Generator Feed Pump Turbine
- I T SW Main Stelim Piping from Steam Generators to and Including-Main Steam *Trip Valve I p SW Main. Steam Piping, Valves and Supports from Trip Valve to and Including Turbine Stop Valves** **
- NA NA SW) Check will be made that design Turbine Steam Bypass Piping, Valves and ) basis earthquake would not cause Supports to Condenser NA NA SW) failure to function Circulating Water System
.Condenser NA NA SW Check will be made that condenser water boxes will not fail during earthquake Circulating Water Piping, Valves and Supports from High Level Intake Canal
*to Circulating Water Discharge Tunnel to and Including Condenser Intake Butterfly Valve; Condenser Discharge Butterfly Valve I T SW P, underground Circulating Water Discharge Tunnel I p SW Circulating Water Pump Vacuum Breaker I NA SW
*e-
15.2-11 12-1-69 TABLE 15,2.1-1 {Continued) Earthquake Tornado Item Criterion Criterion Sponsor Notes SYSTEMS {CONT'D) Condensate and *Feed-Water System 100,000 Gal Condensate Storage Tank I p SW Assume full.of.water Auxiliary Steam Generator Feed Pumps I p SW Piping, Valves and.Supports From 100,000 Gal Condensate Storage Tank to Auxiliary.Steam Generator Feed Pump I T SW
- From Auxiliary Ste.am Generator Feed Pumps to Steam Generator Feed Lines I p SW Steam Generator Feed Lines Inside Contain-ment to and Including First Isolation Check Valve Outside_ Containment I p SW Primary.Vent and Drain System Primary Drain Cooler I p w Piping, Valves and Supports I p SW Primary Drain Transfer Tank I p SW Secondary Vent and Drain System Steam'Generator Blowdown Piping, Valves
*and Supports Inside Containment to and
- Including First Isolat_ion Trip Valve I p SW Gase_ous Waste 'Disposal *system Waste Gas Decay Tanks I p SW Waste *Gas Recombiner System I NA SW Waste'Gas Compressors I NA SW Waste Gas Filter I NA SW Process Vent Blowers I NA* *SW Waste Gas Piping, Valves and Supports from Stripper to Process Vent I NA SW Process Radiation Monitoring System Process Vent Particulate Monitor I NA SW Process Vent Gas Monitor I NA SW Spray Recirculation Heat Exchanger Service Water Monitors I NA SW Area Radiation Monitoring System Main Control Room Monitor I p SW
15.2-12 12-1-69 TABLE 15.2.1-1 (Continued) Earthquake Tornado Criterion Criterion Sponsor Notes SYSTEMS . (CONT'D)
- Instrumentation and Control All Instrumentation and Control to
-Operate-and Monitor Operation of Critical System Component Shown Above During MCA or Controlled Shutdown *
- Systems* Include:
Reactor Protection .(in Par.t) I p w Safeguards Initiation I NA W/SW Containment Isolation I p W/SW Reactor Control (in part) I p w Includes Trip Breakers Steam Generator Water* Level Control System I p w Reactor Make-up Control.. I p w Nuclear Instrumentation* (in part) I p w Non-nuclear Process* Instrumentation I p W/SW (in part) Electrical Systems* Auxiliary Diesel Generators I p SW p Fuel.Oil Day Tanks I p SW p Fuel Oil Transfer *Pumps
- I T SW Underground Fuel Oil Storage Tanks I p SW Assume 1/2 full of oil Fuel Oil Piping, Valves *and Supports SW T for piping to protected_
Auxiliary Generators* I T generators Station Service Batteries and Charges I p SW Vital Bus and-Inverters I p SW Emergency Station Service Transformers I p SW Emergency S.tation Service Switchgear I p SW Control Panel Boards I* p SW Pressurizer Heater*Control Group Only I. p w All Cable to Critical Components, Instru-ments, and Controls as Shown Above I p SW Cable passing through unprotected* are.as will be in rigid conduit Miscellaneous Reactor Containment Crane I p SW
- Piping I p SW
15.2-13 12-1-69
- 15 .2.2 NORMAL WIND LOADING All structures were designed to withstand the following wind loads applied*to the projected area of all surfaces:
El. 26 ft-6 in. to El. 56 ft-6 in., 30 lb per sq ft El. 57 ft-6 in. to El. 75 ft-6 in., 35 lb per sq ft El. 75 ft-6 in. to El. 130 ft-0 in., 45 lb per sq ft El. 131 ft-0 in. and above, 55 lb per sq ft Roofs were designed for uplift using 1.25 times the wind load taken at the corre-sponding elevation of the roof *
- Members subject to stresses produced by this wind load combined with live and dead loads were proportioned for stresses 33 1/3 percent.greater than conventional working stresses provided that the section thus required is not less than that required for the c~ination of dead and live loads computed without the one-third increase.
15.2.3 TORNADO CRITERIA Section 2.2 outlines the probability of a tornado occurring at the site. Although no structural damage is known to have resulted to a reinforced concrete building in a tornado2, the structures and systems so indicated in Table 15.2.1-1 are designed to assure safe shutdown of the reactQr when subjected to tornado load-
- ings.
15.2-14 12-1-69 The tornado model used for design has the following characteristics: Rotational velocity 300 mph Translation velocity 60 mph Pressure drop 3 psi in 3 sec Overall diameter 1,000 ft Radius of maximum winds 200 ft Applicable structures are designed to resist a maximum wind velocity associated with a tornado of 360 mph which is obtained by adding the rotational and trans-lational velocities. Structures and systems are checked for tornado pressure loading, vacuum loading, and combination of these two. The tornado wind velocity is converted to an equivalent pressure which is applied to.the structures uniformly using the formula: 2 P = 0.00256 v where P = equivalent pressure, lb per sq ft V = wind velocity, mph This pressure is multiplied by applicable shape factors and drag coefficients as given in ASCE Paper No. 3269 by Thomas W. Singell 3 , and applied to the silhouette* of the structure.
15.2-15 12-1-69
- A reduction of the full negative pressure differential is made when venting of the structures is provided. The amount of the reduct-ion is a function of the venting area provided.
Tomado wind loads are combined with other loads as described in Section 15.5*.l.2. Tomado and earthquake loads are not considered to act simultaneously. A uniform wind velocity and a nonuniform atmospheric pressure gradient is incorpor~ted in the design of .the containment structure. Structural design criteria for _tomado loading for the containment structure is given_in Table 15.5.1.2-1 and Section 15.5.1.5 *
- .It is assumed that a tomado could generate either of the following potential missiles:
- a. Missile equivalent to a wooden utility pole 40 ft long, 12 in. diam, weighing 50 lb per cu ft 3 and traveling in a vertical or horizontal direction at 150 mph.
- b. Missile equivalent to a one ton automobile traveling at 150 mph.
The amount of penetration of the missile is determined from the Ballistic Research 4 Laboratorie~ formula
- The design assumes maximum wind forc~s and partial.vacuum to occur simultaneously
- with the impact of either of the missiles singly. Allowable stresses do not exceed 90 percent of the certified minimum yield strength of the steel, the
15.2~16 12-1-69 capacity reduction factor given in Section 15.5.1.2 times the certified minimum yield strength of the reinforcing steel, and 75 percent of the ultimate strength of the concrete. 15.2.4 SEISMIC DESIGN Class I structures, systems~ .and components designed to resist seismic forces are listed in Table 15.2.1-1. The design is based on two separate seismic criteria: the Operational Basis Earthquake (OBE) and the Design Basis Earthquake (DBE), as desc~ibed in Section 2.5 and Appendix A. Acceleration response spectra for each earthquake are given on Fig. 2.5-1 and 2.5-2. Seismic loading includes the horizontal or vertical responses acceleration or
- combinations of both where the effects, as measured by the separate acceleration components, of horizontal and vertical accelerations are combined to produce maximum stress intensities, taking into account any potential adverse effect due to phase of the separate accelerations.
I I Damping factors for the structures, systems, and components are given in Table 15.2.4-1. The design of the containment structures is based on ultimate strength design and loading factors as described in Section .15.5.1.2. Maximum allowable stress levels for both the Operational Basis Earthquake and the Design Basis Earthquake is based on proportions of the minimum yield strength.
15 .2-17 12-1-69
- For other Class I structures, the Operational Basis Earthquake loading is combined with dead, live, and other static loads. Normal wind or tornado loadings are not assumed to occur simultaneously with the earthquake loading. Members are pro-portioned for stresses 33 1/3 percent greater than conventional working stresses provided that the section thus required is not less than that required for the combination of dead and live loads computed without the one-third increase.
Allowable soil bearing values are increased one-third. For Class I structures other than the containment structure, the Design Basis Earthquake is combined with static loads using loading combinations given.in , Table 15.5.1.2-1. For these structures under the Design Basis Earthquake loading, the allowable stresses do not exceed 90 percent of the certified minimum ie iY~eld strength for structural steel, the capacity reduction factor, given in Section 15.5.1.2, times the certified minimum yield strength for reinforcing steel, and the capacity reduction factor times the specified strength for concrete. Allowable soil bearing values are increased by one-half. To allow for unimpeded relative motions between structures, a minimum 2 in.
"rattlespace" is provided between the:
- 1. Containment structures and the auxiliary building
- 2. Containment structures and the fuel building
- 3. Containment structure and the containment auxiliary structures around
*.. the periphery of each containmen:t
15.2-18 12-1-69 4. 5. Fuel building and auxiliary building Auxiliary building and control area In general, the periphery of the "rattlespaces" between buildings is arranged to prevent material entering the space with the inner areas left as a void. Maximum relative notions between adjoining structures are included in the stress analyses of all piping which extends from one building to another. Sand type "A" as described in Section 2.4.3.3, was removed from under the fuel building, auxiliary building and control area and replaced by a dense graded granular fill material as described in Section 2.4.5.1.
15.2-19 2-13-70 TABLE 15.2.4-1 ~ ..)-: I l-DAMPING FACTORS FOR CLASS I STRUCTURES o /3 ; v AL~.J Percent of Component Critical Damping
- 1. Reactor vessel internals and control rod assembly drives
- a. Welded assemblies 1.0
- b. Bolted.assemblies 2.0
- 2. Reinforced concrete reactor support struc*ture, including the reactor vessel 5.0
- 3. Vita1 piping systems
- a. Carbon steel 0.5
- b. Stainless steel 0.5
- 4. Containment structure and foundation 5.0
- 5. Steel framed structures,. including $Upporting structures and *foundations
- a. Bolted 2.5
- b. Welded 1.0
- 6. Concrete structures aboveground
- a. Shear-wall type 5.0
- b.
- Rigid-frame type 5.0
- 7. Mechanical equipment*, including pumps, fans, and similar { terns 2.0
15 .2-20 12-1-69 15.2.5 HYDROSTATIC LOADINGS Finish ground grade at the station is El. 26 ft-6 in. Natural ground water level is at approximately El. 4 ft-0 in. The exterior wall of the containment structure extends approximately 66 ft below the finished ground level water resistant. Membrane protection for this structure is defined in Sections 15.5.1.9 and 15.5.1.10. External pumps for reducing the hydrostatic head on the containment structure are described in Section 15.5.1.3. This latter section also discusses the effect of the buoyant pressure on the containment structure. Exterior surfaces of walls of other Class I structures with floor levels below ~I El. 26 ft-6 in. are covered with a mopped on bitumastic coating below I El. 26 ft-6 in. to establish a water resistant membrane. 15.2.6 LATERAL EARTH PRESSURE Where structures extend below the surface of the finish ground grade, their external walls are designed for the active and passive earth pressures inherent in their location and to meet site design criteria.
15.2-21 12-1-69
- 1.
REFERENCES Task Committee on Wind Forces, "Wind Forces on Structures." Transactions of the American Society of Civil Engineers, Vol. 126, Part II, Page 1, 124.
- 2. Gilbertson, v.c. and Mageanu, E. E., "Tornadoes," AIA Teclmical Reference Guide, TRG 13-2 U.S. Weather Bureau.
- 3. Singell, T. W. , "Wind Forces on Strue tures: Forces on Enclosed Structures," Journal of the Structural Division of the ASCE, July, 1958.
- 4. Russell, C.R., "Reactor Safeguards," published by MacMillan, 1962.
- 5. Newmark, N. M., "Design Criteria for Nuclear Reactor Subjected to Earthquake Hazards," Urbana, Illinois, May 25, 1967
- 15.3-1 12-1-69
- 15.3 MATERIAL 15.3.1 CONCRETE Cement All cement was an approved American brand conforming to the specification for Portland Cement, ASTM Designation Cl50, Type II, low alkali. It is suitable for Class I structures because of its lower heat of hydration and improved resistance to sulphate attack. A low content alkali was specified to minimize the possi-bility of reaction with aggregates. Certified copies of mill tests, showing that the cement meets or exceeds the ASTM requirements for Portland cement, were fur-
- nished by the manufacturer. An independent testing laboratory performed tests on the cement for compliance with the specifications.
Admixtures An air-entraining agent was used in the concrete in an amount sufficient to en~ tr~in from 3 to 5 percent air by volume of the concrete. This agent conformed to the requirements of Standard Specification for Air-entraining Admixtures for Concrete, ASTM C260, when tested in accordance with Standard Method of Testing Air-entraining Admixtures for Concrete, ASTM C233. The air-entraining agent was added separately to the batch in solution in a portion of the mixing water. The solution was batched by means of a mechanical dispenser
- capable of accurate measurement and in a manner which ensured uniform distribution of the agent throughout the batch during the specified mixing period.
15.3-2 12-1-69 Water reducing agents were used when their use was approved in writing by the Engineers. Water reducing agents were Master Builders NB-100, type R or N, manu-factured by Master Builders of Cleveland, .Ohio. Type N is normal NB-100 and is used when a normal rate of hardening is required~ Type R contains a retarder and is used in warm weather to reduce the rate of hardening and to avoid cold joints. Calcium chloride was not used uµder any circumstances. Water Mixing water was obtained from a deep well and was kept clean and free from inju-rious amounts of oils, acids, alkalies, salts, organic materials or other sub-stances deleterious to concrete or steel. equivalent of that suitable for drinking. The quality of the water was the The water was continuously checked and tested for compliance with the above requirements by an independent testing labo-ratory. Aggregates Fine and coarse aggregates conformed to the requirements of the Standard Specifica-tions for Concrete Aggregates ASTM C33. Aggregates were evaluated for potential chemical alkali reactivity. Aggregate~i were free from any materials that could
; \
have been deleteriously reactive in any amount sufficient to have caused excessive expansion of mortar or concrete. All aggregates were tested for compliance with the above requirements by an independent testing laboratory~
15.3-3 12-1-69 Proportioning Proportioning of structural concrete conformed to ACI 301, Chapter 3. Working stress type* ~oncrete and ultimate strength type concrete conformed to the require-ments of ACI 301, Paragraph 302. Ultimate strength type concrete was used in the construction of the foundation mat, exterior wall, and dome of the reactor con-tainment. In general, working stress type concrete was used for other areas. Concrete mixes had a 28-day specified strength of 3,000 psi, except as otherwise noted on the Engineers' drawings. Proportions of ingredients were determined and tests .conducted by an independent laboratory in accordance with the method detailed in*ACI 301, Paragraph 308, for.
- combinations of materials established *by trial mixes.
The maximum slump of mass concrete, as defined in* ACI 301,
- Chapter- 14, ,in general did not exceed 3 in. Slump of other concrete conformed ~o ACI 301, Paragraph 305
- The samples for the slump tests were taken at the end of the last conveyor,* chute or pipeline before the concrete were placed in the forms.
15.3.2 , REINFORCING STEEL Except for the No. 14 and No. 18 reinforcing bars for the foundation mat, exterior wall, and dome of the containment structure, a).l reini;orcingconformed to Grade 40 of the Standard Specification for Deformed Billet-.Steel .Bars for Concrete Rein-forcement ASTM A615
- 15.3-4 12-1-69 For No. 14 and No. 18 reinforcing bars and splices for the foundation mat, exte-rior wall, and dome of the containment structure, see Section 15.5.1.9.
Mill Test Reports showing chemical and physical properties were obtained and evaluated for each heat of steel used in making all reinforcing steel furnished.
15 .4-1
.12-1-69
- 15.4 CONSTRUCTION PROCEDURES AND PRACTICES
- 15. 4.1 CODES OF PRACTICE Materials and workmanship conformed to the following codes and specifications:
td)' ACI 301.:,66 "Structural Concrete for Buildings" and all specifications of the American Society for Testing and Materials referred to in Section 105 and declared to be a part of AC! 301-66 as is fully set forth therein. ACI 304 Reconnnended Practice for Measuring, Mixing, and Placing
- ACI 305 Concrete.
*Recommended Practice for Hot Weather Concreting.
ACI 306. Recommended Practice for *cold Weather Concreting. ACI 318-63 Building Code Requirements for Reinforced Concrete ACI 347 Recommended Practice far Concrete Formwork. Section lII of the ASME Boiler and .Pressure Vessel Code for Nuclear Vessels was used as a guide in the selection of materials, design stresses, and fabrication of the steel containment liner *
- 15. 4-2 12-1-69 ACI 301-66, "Specifications for Structural Concrete for Buildings," together with.
ACI 347-63, "Recommended Practice for Concrete Formwork," and ACI 318-63, Build-ing Code Requirements for Reinforced Concrete," formed the basis for the concrete specifications. ACI 301-66 was supplemented as necessary with mandatory requirements relating to types and strengths of concrete, including minimum concrete densities, propor-tioning of ingredients, reinforcing steel requirements, joint treatments, and testing agency requirements. Admixtures, types of cement, bonding of joints, embedded items, concrete curing, additional test specimens, additional testing services, cement and reinforcing steel mill test report requirements and additional concrete test requirements were specified in detail. Concrete protection for reinforcement, preparation, and cleaning of construction joints, concrete mixing, delivering, placing and curing, with the following exceptions, equaled or exceed~d the requirements of AC! 301: Section 1404 (a) - Maximum slump was generally restricted to 3 in. to permit placing concrete in the heavily reinforced containment stru~tures. The slump was increased to 4 to 5 in. ;i.n the areas of the containment wall adjacent to the equipment and personnel hatches where the large steel inserts and additional reinforcing steel required a n;tore plastic mix for adequate concrete placement .* All concrete mixes WE)re designed and tested before use. All concrete mixes used in the work were fully documented. Section 1404 (b) - Maximum placing temperature of the concrete when deposited conformed to the requirements of ACI 305-59, "Recommended Practice for Hot Weather Concreting." .
15 .4-3
..12-1-69
- Section 1404 (c) - Minimum placing temperature of the concrete when deposited conformed to the requirements of ACI 306-66, "Recommended Practice for Cold Weather Concreting. i, 15.4.2 CONCRETE Concrete ingredients were batched in a batch plant and transferred to transit mix trucks for mixing, agitating and delivering to the point of placement.
Water was added to the mix with the other ingredients b.efore the truck left the batch plant area. Batching and mixing otherwise conformed to ACI 301, Chapter 7. Placing of concrete was by bottom dump buckets, concrete pumps, or by conveyor
- i ~elt*. Bottom dump buckets did n,ot. exceed 4 cu yd in size. The discharge of concrete was controlled so that concrete could be effectively compacted around embedded items and near the forms.
For placing of concrete for the wall and dome of the containment structure, see Section 15.5.1.10. I Vertical drops greater than 6 ft for any concrete were not permitted, except where suitable equipment was provided to prevent segregation. All concrete placing ~uipment and methods were subjected to the approval of the Structural
-Engineer.
The surfaces of all construction joints were thoroug~ly treated to remove all
- laitance and to expose clean, sound aggregate. Surfaces of fresh concrete were
- 15. 4-4 12-1-69 rouihened by cutting with an air-water jet after the initial concrete set had ~
occtlrred, but before the concrete had reached its final set~ After cutting, the surface was washed and rinsed. Where, in the opi~ion of the Engineers, the use of an air-water jet was not advisable in any specific instance, then that surface was roughened by hacking with hand tools or other satisfactory means to produce the .requisite clean surface. Before placing subsequent concrete lifts, the surfaces of all construct!~ joints were thoroughly cleaned and*wetted, all excess water that was not absorbed by the concrete was removed. Horizontal construction joints were then covered by a 1/2 in. thick layer of sand/cement grout of the same sand/cement
- ratio as the concrete, .and new concrete was then placed immediately against the fresh grout.
Curing and protection of freshly deposited concrete conformed to AC! 301, Chapter 12, using curing compounds conforming to ASTM C309 as approved by the Engineers. For curing of the top surface of the containment foundation mat, see Section 15.5.1.10. Concrete strength tests were performed in accord:ance with AC! 301, Chapter 16, . Section 1602 (a), Paragraph 4~ supplemented as follows: Not less than two sets of compres:sion test specimens for. each mix design. of concrete placed were made during the first two days of placing concrete or at
- 15. 4-5
.l.2-1-69
- least one set of test specimens for each 250 cu yd placed. Thereafter, one set of test specimens was made for each 250 cu yd, or fraction thereof, for each mix design of concrete placed in any one day. In addition, one set of specimens was made whenever, for any reason, the materials, methods of con-creting, or proportionirtg was changed *
. The test specimens for compressive strength were 6 in. diam and 12 in. long cylinders. Each set consisted of five specimens, at least one of which was tested at 7 days and three at 28 days age. The remaining cylinder was retained at the laboratory for further tests at 60 days* age if the result of the previous tests made such a test desirable *
- .Concrete strength tests were evaluated by the Engineers in accordance with ACI 214-65, "Recommended Practice for Evaluation of Compression Test Results
- of Field Concrete" and ACI 301-66, Chapter 17.
Strengths of working stress type concrete were considered satisfactory if the average of any five consecutive strength tests of the laboratory cured speci-' mens at 28 days age was equal to or greater than the specified compressive strength, f'c, of the concrete. Strengths of ultimate strength type concrete were considered satisfactory if the average of any three consecutive strength tests of the laboratory cured specimens at 28 days age was equal to or greater than the specified compressive strength, f'c, of the concrete *
- 15. 4-6 12-1-69 If any tests for individual cylinders or group of cylinders failed to reach the specified compressive strength, f'c, of the concrete, the Engineers were immediately notified to determine if further action would be required.
,*I.'.
The field tests for slump of Portland cement concrete were in accordance with ASTM Cl43~ Any batch not meeting specified requirements was rejected. Slump tests were made frequently during concrete placement ,and each time concrete test specimens were made. Statistical quality control of the concrete was maintained by a computer program. This program analyzed compress*ion test* results reported by the testing laboratory in accordance with methods recommended by ACI 214, "Recommended Practice for Evaluation of Compression Test Results of Concrete." 15.4.3 REINFORCING STEEL Placing of reinforcing steel conformed t~ the requirements of Chapter 5 of AC! ~01, "Structural Concrete for Buildings," and Chapter 8 of AC! 318, "Building Code Requirem~nts *for Reinforced Concrete."
. I All Cadweld splices were made in accordance*with the instructions issued by the manufacturer, Erico Products, Inc., Cleveland, Ohio.
In order to qualify operators for making Cadweld process joints, each operator was required to dernonstrate,to the Senior Quality Control Engineer,his ability
15.4-7
'l.2-1-69 to make an acceptable fixed joint using the Cadweld process. Cadwelders were requalified after every 200 Cadwelds. Testing was by tensile testing a Cadweld made under simulated field conditions.
The ends of the reinforcing steel bars to be joined by the Cadweld process were square cut by the fabricator. The ends of the bars were thoroughly cleaned of all rust, scale, grease, oil, water, or other foreign matter before the joints were made. Welding was performed using the '.'Metallic Arc Welding Process" with coated
.. J electrodes, or the "Metallic Inert Gas Shielding Welding Process" (MIG) using e bare wire. The filler metal for the Metallic Arc Welding Process conformed to ASTM A316 Coated Arc Welding Electrodes, Classification E-10016-D2 or E-10018-D2.
The filler metal for the "MIG" welding process was a*spooled bare wire 0.30 in. or 0.35 in. diam, Linde or Arcos Type 515. The shielding gas used for the "MIG" welding process was Linde C-25, a mixture of 75 percent argon and 25 percent carbon dioxide. The ends of the bars to be joined by butt welding were prepared by sawing or flame cutting and dressing by grinding, where necessary, to form a single vee butt joint. Mil~ test reports of the heats of steel used for making the rebars were .e obtained by the Senior Quality Control Engineer to confirm the grade of steel welded. Where preheating was required, temperatures were checked with Tempilstiks.
15.4-8 4-15-70 In order to qualify welders for work on the reinforcing steel bars, each welder made a reinforcing bar test weld in the horizontal fixed position, weld-ing vertically up. Each test weld was sectioned through the center of the weld by power sawing and machining.* The cross-sectioned surface was etched with a 10 percent solution of nitric acid and water. The etched surface was examined by the field w~lding supervisor, who determined the qualification of the welding operator. Tack welding of rebar was not permitted. Special criteria for placing reinforcement for the containment structure is in Section 15.5.1.6. 15.4.4 CONSTRUCTION PROCEDURES The portion of the site to be covered by structures was cleared and general excavation performed to the underside of the foundations for the various build-ings. In general, this excavation was from El. +34 to +10 with some build-ing foundations slightly higher or* lower. The major Class I structures (except for ~he Fuel Building and main steam valve enclosure structures) are supported on mat foundations; the Fuel Building and the main steam valve enclosure structures are supported on pile foundation. For additional construction procedures for other Class I structures, see Section 15.6.1.
15.4-9 12-1-69 15.4.5 CXJNSTRUCTION PRACTICE The Applicant maintained on the site at all times Quality Control personnel to serve as qualified inspectors in all phases of work to assure and document that all construction operations met the rigid requirements of the specifications as outlined in the quality assurance report. The qualification of all welding procedures and welders was performed in accordance with Part A of Section IX of the ASME Boiler and Pressure Vessel Code. Concrete was sampled and tested during construction in accordance with AC! 318 to assure compliance with the specifications. A competent independent testing laboratory was retained to design the concrete mixes, take samples, perform all tests of aggregates and concrete cylinders and report to the Applicant for approval. Special practices to be followed for the containment liner are in Section 15.5.1.8. See also 15.4.6, "Quality Assurance Program."
15 .4.6-1 12-1-69 15.4.6 QUALITY ASSURANCE PROGRAM 15.4.6.1 General Since the issuance of the construction permit, the Surry Quality .Assurance Program has been continually upgraded. The significant changes are as follows: I I
- 1. Reorganization of the applicant's Architect-Engineers, Stone & Webster Engineering Corporation, Quality Assurance structure to provide a separate independent Quality Assurance ef.fort.
- 2. Applicant's assignment of a full time resident engineer to provide owner
- surveillance and audits of the Architect-Engineer's Quality Control effort on site.
- 3. Applicant's establishment of a full time supervisory Quality Assurance Engineer on the system level to provide owner surveillance over all fa~ets of the total Quality Control effort.
The Quality Assurance Program encompasses all phases of this project as follows:
- 1. Detailed engineering and design
- 2. Equipment, installation, and testing specification
15.4.6-2 12-1-69
- 3. Vendor selection
- 4. Surveillance of Vendor's shop inspection 5* Shop tests
- 6. Field inspection and Quality Control duri~g erection
- 7. Check-out of mechanical, fluid, and electrical systems
- 8. Startup testing 9.
10. Periodic in-service performance tests Assembly and maintenance of Quality Control documentation for shop and field All structures, systems, and components whether safety related or not, were subject to engineerirlg review, and shop and field inspection. Structures, systems, and components which are safety related were subjected to rigid and closely prescribed procedures including any special controls, processes, test equipment, tools, and skills necessary. The degree*of attention was unrelated to cost, but only to the contribution of safety and reliable operation.
15.4.6.:.:3 12-1-69 - Table 15.2.1-1 summarizes such safety related structures, systems, and components; in addition, each drawing and/or specification was considered individually to determine the applicable category. 15.4.6.2 Organization Ultimate responsibility for safety aspects of the station, including design and conformance of equipment, and equipment installation in accordance with the design, rests with the Applicant. The Applicant delegated the establishment and execu-tion' of the Quality Assurance Program to the Applicant's Eng,ineers, Stone & Webster Engineering Corporation, and to the supplier for the Nuclear Steam Supply System equipment, Westinghouse Electric Corporation, but the Applicant retained
- responsibility therefor.
The Stone & Webster engineering design, equipment specifications, qualification of potential equipment Vendors, and Quality Assurance programs were submitted for the Applicant's approval. The Westinghouse engineering design, equipment speci-fications, including quality standards, and qualifications of potential equipment I Vendors were submitted for the Applicant's approval.* Conformance to approved requirements and programs was assured by close liaison among the project Engineers of the three companies and by such surveillance and audits as the Applicant deemed necessary
- 15.4.6-4 12-1-69 Surveillance included audits of design, specifications, potential Vendors, and the Quality Assurance Programs. The Applicant's Resident Engineer and supporting staff at the site confirm that components and systems related to selected major systems and equipment are of approved designs and materials.
Provisions for audits were clearly established in contracts. The Applicant regularly reviewed the status and adequacy of the Quality Assurance Programs and assured that the originating Corporations regularly reviewed their own programs. Fig. 15.4-1, Quality Assurance Project Organization, indicates how the Quality Assurance Program relates to the overall Surry Project Organization. Fig. 15.4-1 shows the accountability lines between all Vendors and the Stone & Webster Quality Control Organization and depicts the independence of the Quality* Assurance Organization from the individual or group directly responsible for performing the specific activity.* Fig. 15.4-2 depicts the organization of Westinghouse, the Nuclear Steam Supply System supplier. The Applicant's Quality Assurance Organization is described in Section 15.4.6.3.
- The Stone & Webster Quality Assurance Program is discussed in Section 15.4.6.4.
The Westinghouse PWR Systems Division Quality Assurance Plan is discussed in Section 15.4.6.5.
15.4.6-5 2-15-71
- 15.4.6.3 Virginia Electric and Power Company Quality Assurance Program A Quality Assurance Organization was established within Vepco to ensure that all systems and structures affecting the safety of the station were specified, fab-ricated, shipped, stored, installed, inspected, and tested in accordance with sound engineering principles including all applicable codes, specifications, regulations and procedures. This organization is shown as part of Fig. 15.4-1.
The organization was headed by a full-time experienced Quality Assurance Engineer with sufficient personnel supporting him to implement the program. All items pertaining to quality assurance were under his surveillance. The overall responsibility for the Applicant's Quality Assurance Program rested with the Vice President-Power. During design and construction, the responsibility was assumed by the Director-Power Station Design and Construction. Quality Assur-ance during the testing and operational phases was the responsibility of Director Production Operations and Maintenance. All were assisted in these responsibilities by the Director-Nuclear Services. The Quality Assurance Engineer audited both the Stone & Webster and Westinghouse phases of the program. This auditing function included review, along with Vepco Staff Engineers, of all specifications, evaluation of key Vendor's Quality Assur-ance Programs, witness of key shop tests, and review of the Field Quality Assur-ance.Programs and their implementation by the various construction Contractors .
~ '15.4.6-6 2-15-71 I I I The duties of the Quality Assurance Engineer are as follows:
- l. Review equipment design and specl.ficati.ons for compliance with codes, Atomic Energy Commission criteria, and proper design for inspection and tests.
- 2. Review Stone & Webster and Westinghouse shop and field quality control manuals.
- 3. Review quality assurance and control procedures.
- 4. Investigate the proper execution of quality assurance and control procedures .
- 5. Coordinate the Quality Control review by Vepco Engineers of Stone & Webster general design concepts, equipment specifications, and correspondence and drawings exchanged between Stone & Webster and equipment suppliers.
- 6. Spot inspections of Vendor's shops and witness of equipment testing.
- 7. Make frequent visits to site.
The Vepco personnel supporting the Quality Assurance effort consisted of the Director-Nuclear Services, Director-Power Station Design, Director Production
15.4.6-7 2-15-71 Operations and Maintenance, and their staffs. The Director-Power Station Design assisted by a staff of competent engineers performed the following functions in relation to the Quality Assurance Program: 1, Review and approve Stone & Webster general design concepts.
- 2. Review and approve Stone & Webster calculations in special cases.
- 3. Review and approve Stone & Webster designs for adequate materials, ease of operation, maintenance, and reliability.
- 4. Review and approve equipment specifications, bidders lists, bid
- evaluations, and purchase requisitions for safe adequate equipment and quality control requirements.
- 5. Review all correspondence and drawings exchanged between Stone & Webster and equipment suppliers, initiating appropriate action if required.
- 6. Inspection of Vendor's shops with Stone & Webster and Westinghouse Engineers.
During the design and procurement phases of the project all Stone & Webster design studies, drawings, specifications, bidding documents, Vendor's corre-spondence, Vendor's prints, etc. were submitted to Vepco for review and approval. Each item was reviewed by engineers within whose jurisdiction the item falls for
- technical adequacy, reliability, and compatability with the station in general
15.4.6-8 12-1-69 and the Vepco system. '\'::' The Quality Assurance Engineer was responsible for reviewing all items and Vepco comments for reliability and safety. Vepco engineers were in frequent telephone contact.with their counterparts in S*tone & Webster and Westinghouse to discuss technical design details. An engi- ~eering meeting and a general review meeting were held as needed between Stone & Webster, Westinghouse~ and Vepco to review basic concepts, design details, required actions, responsibilities, and job progress. Selected Vendor's shops were inspected for quality control by Vepco engineers along with engineers from Stone & Webster and Westinghouse. Key shop tests on major components, such as hydrostatic tests, pump capacity tests, control systems checkouts, core and coil inspections, final assembly inspections, etc. were wit-nessed.
.::.*"I The Vepco Resident Engineer was assigned full time to monitor station construction.
In addition to reviewing quality control and safety, the Resident Engineer provided liaison between the construction forces and Vepco Operations personnel for equip-ment clearances and final equipment acceptances. In addition, engineers from the Vepco engineering.departments made frequent inspection trips to the job site. The Vepco audits did not relieve the Stone & Webster personnel of responsibility f~r day to day job progress and quality but advised Stone & Webster of unsatis-factory practices, if any, and requested Stone & Webster supervision to have such piactices stopped or revised.
15.4.6-9 12-1-69
- The Surry Station Operating Staff were onsite during the last year df construction of Unit 1 to become familiar with. the station and perform component and system preoperational testing with the assistance of Stone & Webster and Westinghouse.
The Vepco Resident Engineer and Operating Staff also checked the readiness of equipment for initial operation, and checked the cleanliness of piping and equip-ment for initial operation of controls and protective devices. All operations of equipment (such as valves, motors, pumps, controls, etc.) were accomplished by Vepco operators. As each operating system was completed, a set of equipment acceptance and starting sheets were filled out by Stone & Webster supervision, Vepco Resident Engineers, and Vepco Operating Supervisors signifying that the equipment was complete and checked operationally and was provisionally accepted
- by Vepco for operation. These procedures were audited as part of the Project Quality Assurance Program.
From test specifications prepared by Stone & Webster and Westinghouse, Vepco operating personnel prepared test procedures for the nuclear and safety related power plant equipment, operated said equipment for the prescribed tests, recorded the necessary data, and prepared summaries of test results. These test data and results were reviewed, evaluated, and approved by Stone & Webster, Westinghouse, and Vepco engineers as applicable
- 15.4.6-10 12-1-69 15.4.6.4 Stone & Webster Quality Assurance Program General The Stone & Webster portion of the Applicant's Quality Assurance Program was designed to provide a broad service of quality assurance surveillance. This service included adequate technical specifications by the Engineers which were reviewed by the Engineering Assurance Review Connnittee to assure a quality end product as shown in Fig. 15.4-1, preparation of field and shop Quality Control procedures, field and shop Quality Control manpower and technical assistance, assembly and maintenance of records necessary to assure and document the use of correct material and workmanship specified by engineering specifications and drawings, and the accumulation of sufficient documentary evidence to indicate the conformance and corrective action taken. The program included as a function of quality assurance, the ~ontrol and surveillance as shown below:
- l. Surveys and audits conducted by' the Engineering Assurance Review Committee for compliance to procedures prepared by Quality Assurance Engineers.
- 2. Preparation of technical specifications and drawings.
- 3. Review of specifications and drawings.
- 4. Source inspection covering the inspection and surveillance of supplier Quality Control in the mills, fabricating shops, and equipment manu-facturer's shops.
15.4.6-11
. ;':.{- _( 12-1-69
- 5. Inspection upon receipt at the construction site.
*....6...,S,u:r'l1'e.il.lance Qf proper st:qt;'age; ha,~dling,*.:;an,d ,*preQperaUonal:,~intenace
- at . t~~ co.nstruc;tion .site *..
J *:-
. -* ,* '... ;_*; :: ~_'; *-: _r:: *.
- 7. *. ,Qu.~li;y :Control inspect:i,on ,an,d, .s.1.p;y.~illance, at:-: the ,.~Qnst-r:uqtf.,on, -s.it~ *
., . ":,_13 .:.: ,,__: :.~
- 8. Material control at the construction site.
- 9. Quality Control of general materia],5..,.,suc,h as):con,ci;.ete .and s;t:eel.
- I
.1
- 10. Detection of nonconformances during constructiqn .,or,* l:!_hqp _inspe.ct:i,,o~
and recommendations of corrective action.
- 11. Authority to cause stoppage of work at the construction site for nonconformances.
- 12. Establish and monitor source Quality Control ratings *.
- 13. Assemble and maintain Quality Control records and documentation package.
The Manager-Quality Assurance Depal:'tment in the S.tone & Webster. Headquarters cool:'dinated the quality assurance prog.ram in, all effortsin the three major. areas. of activity: Engineering, ~pop, .and field_, as shown, in Fig. 15.,4-1.
15.4.6-12 2-13.,-70 Design Control Written procedures were prepared to provide .for systematic a11d Tedundant assess-ment and r~view of adequacy of design. The procedures required the review and signature of at least one other engineer, in addition to the individual primarily responsible for preparation of the engineering document. This ~h~ineering review included the following original documents and their revisions:_
- 1. Applicable criteria, list of quantitative and qualitative design requirements, and conceptual-design reports 2.
3. Engineering *calculations Engineering specifications
- 4. Engineering drawings
- 5. Memoranda of engineering changes Design Criteria Documents Design cri.teria documents were reviewed and approved by the principal and project engineers on the project before submission to the Applicant for fina~ review and approval; such documents were the PSAR, FSAR, and design memoranda and reports.
PSAR and FSAR chapters defining unit safeguards were reviewed and approved by
15.4.6-13 12-1-69
- the supervising engineer of the Safety Analysis Group, as well as by the Project Engineer, Senior Project Engineer, and the Lead Nuclear Engineer on the project.
Design criteria documents included evaluation of:
- 1. Accident analyses
- 2. Compatibility of materials
- 3. Compatibility of design interfaces
- 4. Accessibility for inservice inspection, maintenance, and repair
- 5. Accep.tance criteria for inspections and tests
- 6. Resolution of regulatory and Vepco criteria Engineering Calculations Calculations for the Surry project fell into two categories: Hand calculations, and machine calculations. Hand calculations prepared on' this project were reviewed, approved, and initialed by another competent engineer. Hand calculations prepared in a technical support group were reviewed, approved, and initialed by the supervising engineer of the technical support group. Machine calculations were pre-pared with a summary sheet, including a;1 pertinent design criteria and other design input and results, including curves, tables, etc, and spot hand calculations were
15.4.6-14 12-1-69 made to verify the results. The supervising engineer of the technical support group reviewed, approved, and signed these documents. The data print out sheets were cataloged and filed in the technical support office. Procurement Document Control, Instructions, Procedures and Drawings Internal Stone & Webster reviews were designed to assure that the drawings and specifications were adequately supported by *the engineering and design groups. The review assured not only completeness and correctness of detail but also assured basic concepts and policies were correctly interpreted and expressed in the detail. The concept and philosophy of safety related systems, structures, and components were reviewed by the Responsible Lead Engineer of the appropriate discipline involved to assure conformance with AEC criteria and to obtain the combined best judgment and experience of Sto~e & Webster Engineers. i All calculations relating to engineered safeguards and other critical station aspects were confirmed by an eng~neer other than the originator, and signed by him, as well as the originator. Calculations and references were recorded in standard format in sufficient detail so that they could be easily audited. Other calculations were also subject to internal review, although less formal procedures were used depending upon the significance of the equipment or structure involved.
15.4.6-15;. 12-1-69
- After drawings were produced, they were thoroughly reviewed by another designer, as were all subsequent revisions, for adherence to standards, supporting calcula-tions, engineering instructions, compatibility with other drawings, and correct-ness of dimensions. The drawings were then examined and initialed by the Design Supervisor, the R~sponsible Engineer, and the Project Engineer prior to issue.
Specifications incorporated applicable standards developed from the Engineers. experience and special requirements applicable to systems, structures, and com-ponents, and related to engineered safeguards and station reliability. Speci-fications were prepared in standard format developed by a committee of specialists and were appended with special requirements for the Surry. project. SpeciJications were sufficiently complete to provide adequate information for
- Quality Control Personnel. For Class 1 materials and components the specifica-tions included instructions to the Stone & Webster Inspection Division.
These specifications were reviewed and signed by the Project Engineer, by a Specialist for each type of equipment, and by the Stone & Webster Engineering Assurance Review Committee before submittal to the Applicant for approval prior to bids being solicited and an order placed. Bids were solicited only from Vendors or Contractors with whom satisfactory per-formance had been experienced, or from whom satisfactory performance could rea-sonably be expected as a result of a specific review of their performance and. after the Applicant's approval. Dunn and Bradstreet reports and shop inpection reports were required of all Vendors with whom Stone & Webster had no previous or recent experience.
15.4.6-16 12-1-69 Vendor drawings, including Westinghouse drawings, performance curves, data sheets, and similar information were reviewed by the Responsible Engineer, assisted as necessary by specialists, for adequacy of materials, conformance to specifica-tions, _and*sound engineering practice. Logic diagrams and electrical control and instrumentation drawings related to engineered safeguards were reviewed to assure not only that they were techni-cally workable but that they completely expressed the intent of the safety system philosophy, and to assure complete communication between the disciplines involved. Noncompliances which could not be corrected within the scope of the specification~ by Vendors or site Contractors were referred by Quality Control Personnel to the. Project Engineer for resolution. The Project Engineer, assisted by the Lead
- Engineer responsible for specification of the component, and such specialists as required because of the nature 0£ the noncompliance or*end use of the component, determined the action to be taken. Acceptanc~ of a noncompliance for equipment related to engineered safeguards was determined by technical considerations and not modified by considerations of cost or schedule.
Consultants who specialize in structural analysis, seismology, or other selected disciplines, were engaged, as necessary, to supplement the Stone & Webster staff*, with Stone & Webster retaining responsibility for engineering and design, 1 The Engineering Assurance Review Committee within the Engineering Department was composed of one representative each from the Nuclear, Structural, Mechanical, Electrical, Building Service, and Hydraulic Divisions of, the Engineering Department. The Connnittee Chairman was accountable to the Manager-Quality Assurance Department.
15.4.6-17 12-1-6.9
- The Committee was responsible for the following:
- 1. Establishing levels of Quality Assurance necessary to satisfy Applicant requirements, AEC criteria, and applicable codes.
- 2. Implementing a program of Quality Assurance in engineering that was identified by engineering assurance procedures, published, and widely distributed.
- 3. Conducting surveys and audits of all projects and technical divisions to assure compliance with its procedures.
- 4.
5. Asssessing Quality Assurance
- Preparing standards concerning Quality Control for insertion in specifications.
- 6. Auditing of bidding specifications to assure conformance with Quality Control Standards.
The preparation and issuing of procedures for committee approval, t~e preparation of audit plans and the assessment of audit results for committee action, the follow;up duties relating to the corrective action recommended by the committee, the review of Shop and Field nonconformance reports to evaluate engineering per-formance, the preparation of training programs were all responsibilities of
- Engineering assurances.
15.4.6-18 12-1-69 Document Control Project General Instructions and Quality,Assurance Procedures established instruc-tions to control the issuance of documents, such as instructions, procedures, specifications and drawings, inclµding changes thereto, which prescribed all activities affecting quality. These measures assured that documents including changes were reviewed for adequacy and approved for release by authorized per-sonnel. Distribution to the location, and use at this location, by the designated activity was carefully controlled by means of document logs and signed document receipts. The Document Control Program was audited by the Chief Quality Control Engineer.and Senior Quality Control Engineer for proper functioning of the pro-cedures established. Control of Purchased Material, Equipment, and Services The surveillance of quality assurance in the Vendor's shops was supervised by the Chief Quality Control Shop Inspector. He wa~ in direct charge of seven district offices throughout the country and was accountable to the Manager-Quality Assur-ance Department who was accountable to the Applicant. Each of these offices was under the supervision of District Chief Inspector. Direct contact between the district offices and field projects was maintained. The inspectors assured Vendor compliance with the applicable contract and product conformance with the purchase order and specifications, which included detailed inspection instructions. The extent of shop surveillance depended upon the nature of the item, the purchase order and specification requirements, and the quality history of the Vendor.
15.4.6-19 12-1-69 Surveillance.of inspection and testing in the Vendor's shops was usually planned so as to assure a minimum amount of interference with the shop's normal work performance. The Vendor's shop Quality Control program assured that all material, equipment, and systems were in accordance with the applicable purchase orders, specifications, and codes. Such material or equipment and systems were then ready for source inspection by Stone & Webster shop inspectors at such check points as.were established after review of the purchase orders and specifications. The Stone & Webster inspector visited the Vendor'~ plant at sufficient intervals during fabrication to allow him to review results of all tests required by pur-chase orders, specifications, and codes. The tests and/or inspections that were performed were radiographic, dye penetrarit, ultrasonic, Magnaflux, hydrostatic,
- leak, performance, alignment, final dimension, code data reports, and preparations for shipment. The Stone & Webster inspector kept a checklist of the inspections and inspection reports required by the specification. Copies of this checklist were.submitted to the Applicant with test data.
Shop Quality Control maintained a system and procedu~e for continuously rating the Vendors for the purpose of: \
- 1. Evaluation of the adequacy of the Vendor's Quality Control System
- 2. De.termination that the Vendor followed his approved Quality Control System
15.4.6-20 12-1-69
- 3. Determination that the Vendor's Quality Control System assured required product quality levels
- 4. Advice to Vendor company personnel of the Quality Control status of its product The field Quality Control personnel supplemented the shop Quality Control personnel with information on the quality of material, equipment, and workman-ship of specific Contractors and/or Vendors.
From the above Vendor rating system, information was made available to the Stone & Webster Purchasing Department for guidance in the selection and evalua-
'tion of Vendors and to shop Quality Control personnel for use in s~hed~ling and*
- setting the frequency of visits and inspections.
Field Quality Control Organization .The Quality Control OFganization in the field, including the Senior Quality Control Engineer, was supervised by the Chief Field Quality Control Engineer, who was directly accountable to the Manager-Quality Assurance Department. He, in turn, was accountable to the Applicant. The Senior Quality Control Engineer in the field was also accountable to the Applicants' Resident Engineer. Direct contact with the Contruction Superintendent and the Project Engineer and their groups was established to assure correct interpretation of project specifications
- 15.4.6-21 12-1-69 .
- and drawings and to assure compliance with those specifications and drawings.
The Senior Quality Control Engineer was assisted by competent structural, mechan-ical, electrical; and welding engineers to provide the necessary inspection and surveillance services. In addition, Quality Control testing laboratories were under the supervision of the Senior Quality Control Engineer. n* Direct inspection authority of the Senior*Quality Control Engineer was derived from the Headquarters Quality Assurance Department, thus divorcing him from con-struction pressures. The Senior Quaiity Control Engineer had at least five years' experience in his
***f respective field with special training in the field of inspection *
- Procedure Quality Control personnel had full authority to instruct construction supervisors to prohibit to start of work when conditions prevented the attainment of required quality, to halt work that deviated from specifications or drawings, to institute correction.of faulty workmanship, to effect replacement of material not in accord-ance with the applicable specifications, and to prohibit the use of manufactured equipment .that failed to meet specifications.
A set of Quality Assurance reports, including mill test reports, laboratory reports, manufacturer's reports, field inspection and test reports, were main-
15.4.6-22 12-1-69 tained in an official fireproof job file, properly compiled, and protected. This file served as documentary evidence of compliance to all applicable speci-fications and codes. This file is now part of the Applicant's permanent records. Identification and Control of Materials, Parts, and Components Quality Control procedures required to provide proper identification of material were issued by the Quality Assurance Department. All material and equipment used was reviewed and programmed prior to arrival at the job site from the shop or other source as referenced in receiving inspection procedures. A Quality Control receipt inspection procedure was implemented, and appropriate material which was inspected upon receipt and found improperly identified was correctly identified and provided with proper markings that were legible. Thus, such material recor.ds
- were available for traceability from manufacturer t;o its fina.l installation and documentary evidence can be provided to indicate conformance to applicable speci-fications and. codes.
Quality Control reviewed delivery schedules and shop inspection reports, prior to receipt of material at the job site, for the purpose of determining additional inspections required. Through the combined efforts of the shop and field inspect-ions, Quality Control assured that procured and received material and equipment met the requirements of applicable purchase orders and specifications.
15.4.6-23 12-1-69
- The construction receiving group provided support services, as follows:
- 1. Received and checked shipping containers for damage, unpacked, verified quantities against shipping papers and/or purchase order, ahd maintained such shipping papers/lists and purchase orders available for Quality Control inspection.
- 2. Provided assistance to construction crafts supervision in connection with material and equipment problems.
The extent of additional inspections depended upon the intended functions and the critical nature of the material. Such material'inspected_was reported in
.i ,detail on receiving a Quality Control report which gave special notice to proper identifi~ation, correct type, condition, cleanliness, and preventive maintenance, etc. The receiving report also noted any nonconformances. The cognizant respon-sible engineer was contacted to provide technical support as necessary for Quality Control inspection in the areas of correct interpretation of specifica-tion, codes, and drawings. He was also available for support in making correct dispositions of received material that was declared rejected.
Control of Special Processes Field inspection included areas such as:
- 1. Documenting of welding, heat treating, nondestructive testing, and all
- special processes to verify compliance with codes, specifications, standards, and other special requirements.
15.4.6-24 12-1-69
- 2. A commercial concrete testing laboratory was employed to perform routine testing of concrete materials and to propose design mixes for approval by the Structural Engineer, in the Headquarters Office, Boston. The commercial laboratory provided qualified inspectors to maintain continuous control by checking and recording of concrete batching for all structural concrete for compliance with specification.
- 3. Welding was subjected to Quality Control inspection for conformance to the engineering specifications. This work included, to the extent specified, inspection of joint fit-ups prior to welding, assurance that proper welding procedures were used, welding electrode controis were planned, dye penetrant and/or radiographic inspection of the root pass for inert gas welds was made, radiographing of partial and completed welds was performed, and checking and dye penetrant inspection of the final surface grinding were made. The inspec-tion was performed by a Quality Control Engineer Inspector, who was qualified to perform nondestructive. testing. Qualification of personnel performing nondestructive testing was by written examination.
- 4. Equipment was available on site for performance of both .tensile an,d compressi_on tests. These tests we~e performed extensively on all materials of construction.
The results of these tests were recorded, kept and made available'to all quality assurance personnel.*
15.4.6-25 12-1-69
- Storage, Handling, and Maintenance Material and equipment was stored, h~dled, and maintained in a protected manner by approved written procedures, substantiated and documented hr surveys and audits. Material and equipment rejected upon receipt were segregated from accepted items prior to being properly approved and documented for disposition.
Special ~nvironments; that is, inert gas atmosphere, and specific moisture content and temperature levels were verified and documented. Inspection Equipment, electrical circuits, and fluid systems were inspected during installa-
- ' t,ion and before operation for conformance to specifications, drawings, and func-tional requirements, with particular attention to systems-related to public safety.
In addition to specific assigned duties, Quality Control personnel were constantly alert in the areas of work in progress and work completed for quality of workman-ship, cleanliness, alignments, material protection and preservatives, clearances, and safety regulations. Improper.workmanship that deviated from drawings and specifications, and unsafe or unsatisfactory conditions were promptly reported to responsible supervision for corrective action
- 15.4.6-26 12-1-69 Inspection, Test, and Operating Status Materials and equipment that deviated from approved specifications, codes, dra,Hngs, or other applicable documents were considered as nonconforming supplies.
They were tagged to indicate their acceptable or unacceptable status, reported to cognizant supervision, and segregated. Nonconforming Material, Parts, or Components Nonconformances discovered during shop inspection, receiving, construction, installation, and testing were reported by the cognizant Quality Control Engineer and their disposition determined in accordance with detailed procedures. Examples of nonconforinances that were reported are:
- 1. Nonconformances that were rejected for workmanship
- 2. Nonconformances that were rejected for noncomplying material Fi.eld Quality Control Inspectors and/or Engineers attached reject tags to non-conforming material or equipment and segregated them. Such identification was removed only by field Quality Control personnel when disposition of material or equipment was completed and documented. Shop Quality Control Inspectors did not physically tag nonconformances of material and equipment, but did document and advise on such nonconformances. Rework was inspected by Quality Control as 1
necessary to clear records and to support final acceptance.
15.4.6-27 12-1-69
- Corrrective Action When a nonconformance could be corrected to meet the requirements within the scope of the specification and/or drawings, the Senior Quality Control Engineer indicated the corrective action required, Any nonconformance which required technical evaluation and/or specification changes to justify were referred to the Project Engineer for a statement of disposition.
When the resolution of nonconformance was established, the Senior Quality Control Engineer forwarded the resolution details to the Superintendent of Construction who initiated corrective action *
* .D~ta on nonconformances were collected and evaluated by the Quality Assurance*
Department for the purpose of quality improvement. Field Quality Control personnel instituted corrective action and, as indicated by analysis of these data, assisted to preventing recurrence. At the discretion of the Senior Quality Control Engineer, who ascertained the seriousness of the deficient or rejectable condition, applicable construction was *stopped. Work stoppages were reported immediately to the Chief Quality Control Engineer and the Construction Superintendent or his authorized repre-sentative. The Quality Control Engineers Inspectors were authorized to:
- 1. Prohibit the start of various phases of work until adequate inspection had been provided; for example, placing concrete before reinforcing and
- forms had been inspected or, pipe welding prior to fit-up inspection.
L
15.4.6-28 12-1-69
- 2. Forbid the use of materials, equipment, or workmanship which did not conform to applicable specifications and codes
- 3. Institute removal or correction of faulty construction The shop Quality Control Inspector did not have any authority to stop work in the Vendor's plant. However, if while inspecting material and equipment he recognized nonconformances or deficient conditions to said materials, he immedi-ately notified the Vendor's Quality Control Supervision.
Test Control I A program was established to assure documentation of all required testing, including proof testing, acceptance testing, and operational testing. This program identified each test and assured that the tests were performed in accord-ance with written procedures including the performance requirements and operating limits as specified in applicable desigri documents. These test procedures included all prerequisites and test instrumentation required. Calibration of Measurement and Test Equipment The Quality Control procedures provided instructions to assure that tools, gages,.; instruments, and other measuring and testing devices used by the inspectors were, calibrated to certified standards and properly adjusted, and the results docu-mented at specified periods to maintain accuracy within necessary limits. Veri-fications of records were made by shop inspectors to assure that Vendor's were -~ also complying with these requirements.
15.4.6-29 12-1-69
- Quality Assurance Records The Project General Instructions and the Quality Assurance Procedures provided instructions covering maintenance of Quality Assurance Records and ultimate . ,
delivery of these records to the Applicant except those required by codes and contractual documents to be retained by Vendor. These records included such, documents as:
- 1. Construction weekly reports and operating-logs
- 2. Results of reviews, inspections, test, and audits
- 3.
4. Records of work performance monitoring Records of materials analyses Inspection and test records identify the inspector or reviewer, type of observation, results, acceptability, and action taken for deficiencies. Audits The Quality Assurance Procedures provided instructions and check lists for audits performed in all aspects of Quality Assurance at designated intervals. The audits were performed by qualified personnel not having direct responsibility in the area being audited. Audit results were documented, reviewed by Stone & Webster and Vepco management, and followed by corrective action. Audits* were also performed by the A.E.C. Division*of compliance.
15.4.6-30 12-1-69 15.4.6.5 WESTINGHOUSE PWR SYSTEMS DIVISION QUALITY ASSURANCE PLANT
- Quality Assurance Planning Introduction The Quality Assurance Plan of Westinghouse PWR Systems Division for the Nuclear Steam Supply System is set forth in this document. Its purpose is to describe the procedures and actions used by Westinghouse to assure that the design, materials and workmanship employed in the fabrication and construction of systems, components and installations within the Westinghouse scope of responsibility in a nuclear power plant are controlled and meet all applicable requirements of safety, reliability, operation and maintenance. It is recognized that the ultimate responsi-bility for Quality Assurance rests with the applicant; for this reason provision has been made in the plan for the transmittal of necessary records and information to the applicant.
This plan is a requirement for, but _is_not necessarily limited to,.those components and systems of the plant having a vital role in the prevention or mitigation of the conseq~ences of accidents which can cause undue risk to the health and safety of the public. These are Class I items as described in the Safety Analysis Report. e
15.4.6-31 12-1-69 .. Procedural Documents and Work Instructions Written administrative and technical policies, procedures and instructions are in use in Westinghouse to implement the Quality Assurance Plan. They are in formats appropriate to their applications, such as: Management Responsibility Statements Position Descriptions of Management and Professional Personnel Engineering Instructions Quality Assurance and Reliability Procedures Quality Control Notices Quality Control Plans Projects Procedures Purchasing Manual Procedures Construction Site Procedures Technical and contractual information to assure effective implementation of these policies and procedures is developed, documented and controlled through a standard Westinghouse system which consists in part of: System Design Paramet*ers Equipment Specifications Corporate Process Specifications Corporate Material Test Specifications
15.4.6-32 12-1-69 Corporate Purchasing Department Specifications (including specifications for materials) Drawings Purchase Orders Procedures are reviewed and revised on a continuing basis by the issuing authorities so that the procedures meet the needs for which they are intended. Management reviews performance in accordance with these procedures to assure compliance. Independent audits, as described later, provide objective assurance of both the adequacy of the procedures and compliance with them. Organization Organization Chart Figure 15.4-2 shows the functional organization as related to quality assurance of the Westinghouse Nuclear Energy Systems Divisions. The authority and responsibility of the manager of each activity on this organization chart is set forth in writing in an approved statement of management responsibility. The Westinghouse organization provides the checks and balances needed to foster an effective overall quality assurance program through the functional organization as well as through three levels of organization controls.
15.4.6-33 12-1-69
- At the first level, 'process audits of Nuclear Energy Systems Div.isions ,
are performed by other Nuclear Energy Systems Divisions to assure that functional areas are adequately covered .. At the second level, a Nuclear Energy Systems Quality Assurance Committee under the direction of the Nuclear Energy S~~tems Executive Vice President is responsible for providing Nuclear Energy Systems management assurance t~at the qu~lity assurance policies and practices of the divisions result in products and services that meet safety and reliability requirements. At the third level, the Headquarters Quality Control Staff, reporting organizat~onally independently from the Westinghouse Power Systems Company, is responsible for providing Westinghouse Corporate management assurance that the quality programs of all divisions in the Corporation including those in the Nuclear Energy Systems Divisions are effective in meeting customer quality requirements and Westinghouse objectives. The PWR Systems Quality Assurance depqrtment consists of four sections: Mechanical Equipment, Pressure Vessels, El~ctrical, and Plant Quality Assurance. The Quality Assurance department has responsibility for supplier surveillance, and quality assurance data feedback and analysis, as described elsewhere in this plan. Other Westinghouse divisions are orgari-ized ,for independence of a quality assurance function, as shown in Figure 15.4-2
- 15.4.6-34 12-1-69 Functional Relationships, PWR Systems
- PWR Systems Division is divided into a number of functional groups having both direct and indirect responsibility for aspects of the design, fabrication and construction phases of the project. Close association and interchange of information at all levels exists among the functional groups.
The table of Figure 15.4-3 illustrates the relationships among these groups. For example, contractual requirements originate in Projects, and are distributed to Licensing and Reliability, system functional requirements groups, system design groups and the equipment design and procurement groups. It can 9e seen thaf all aspects of the project are considered at each stage in the overall program, with the respective lead functional group coordinating the efforts of the associated functional groups. Assurance of Design Adequacy Specification of Technical Requirements Engineering is responsible for designing or specifying equipment that conforms to the requirements of the applicant and of the application for which it is intended. This responsibility includes the specification of quality control requirements that will assure that the equipment will function as required in the system and plant.
15.4.6-35 12-1~69
- Systems Engineering designs the plant to meet functional, safety and regulatory requirements. The components design engineers work closely with Systems Engineering to identify equipment limi.tati~ns and to resolve:
functional requirements with equipment capabilities. The design of. equipment also provides for access to components for in-service inspection and maintenance as required to assure continued integrity throughout the life of the plant. Written parameters are forwarded to component design engineers by Systems Engineering detailing the design requirements for the specific.plant. Equipment Specifications or drawings are prepared by the component design 11 engineers to cover these requirements, The term Equipmeiit Specification"
- as used in this Quality Assurance Plan includes drawings when they are used in conjunction with or instead of Equipment Specifications. Detailed quality control requirements are specified in the Equipment Specification, or its references. Examples of these are nondestructive tests, acceptance standards, functional tests, and recording the measured values of key characteristics. In the few cases when Equipment Specifications or design drawings are not used, the specific quality control requirements, tests and acceptance standards are identified in the purchase order.
Review for Compliance with Technical Requirements Preliminary Equipment Specifications are reviewed within Westinghouse by systems engineers, materials and process engineers, licensing engineers,
15.4.6-36 12-1-69 Quality Assurance, Projects, and, including the applicant, others as required.* These_ind~pendent reviews assure that Equipment Specifications meet systems requirements, conform to established engineering standards, are adequate from a metallurgical and welding point of view, meet all code requirements, satisfy all safety requirements including those specified in safety analysis reports, contain necessary quality control requirements, and conform with the applicant's contractual provisions. Written Engineering Instructions describe the requirements of the review. Aspects of the equipment design that have an effect on that part of the plant design performed by the applicant or architect-engineer are forwarded to them for their review. Applicant or architect-engineer drawings which have an effect on the Westinghouse scope of supply are likewise sent to Westinghouse engineers for their review. Technical requirements are provided in the bid package to qualified suppliers of components within the Westinghouse scope of responsibility. Suppliers' proposals responding to these bids are sent to engineering for review. The component design engineer evaluates the supplier's proposal for technical adequacy. He insists on sufficient functional design data to make an independent review of the supplier's design to assure that the equipment will meet all requirements. Consultants from the Westinghouse Research and Development Laboratory and outside experts are also used
15.4.6-37 12-1-69
- to review specific design features, as required. The component design engineer reviews how the supplier intends to meet the specified quality requirements. He reviews the proposed equipment for its capability to perform its function for the design life of the plant.
Westinghouse does not permit exceptions in tpe proposal specifications that adversely affect the safety or reliability of the equipment. Purchase requisitions prepared by the component engineer are the basis for purchase orders issued by Purchasing to suppliers. Purchase requisitions are reviewed by Component Engineering, System Engineering, Projects and
.other functions, as necessary, to assure that technical requirements
- ,have been transmitted correctly to suppliers of the components. The purchase order is the official contract document that covers the technical requirements in the form of the Equipment Specificat'ion.
Purchase orders require suppliers to submit drawings, and manufacturing, inspection and test procedures as the work under the p~rchase order pro&resses. This phase of the design is reviewed independently by Westinghouse component engin~ers. The written instructions for this phase are contained in an Administrative Specification and the Equipment Specification, which form part of the purchase order
- 15.4.6-38 12-1-69 Formal Design Reviews
- In addition to the routine reviews of technical requirements discussed above, formal design reviews are conducted by the Reliability section on critical systems, subsystems and components to improve their reliability and to reduce fabrication, installation and maintenance costs. The design reviews are comprehensive, systematic studies by personnel representing a variety of disciplines who are not directly associated with the development of the product.
Specialists from other Westinghouse divisions and outside consultants are used in the reviews as necessary. Information developed by the reviews is recorded for evaluation and action by the cognizant design engineer. The design review program is projected over a substantial period of time because of the comprehensive nature of each review. Both the scheduling of the review and the selection of specific equipment for review are i ' based upon many considerations including whether the equipment is of a new design, its import.ance to public health and safety, its importance to plant availability and performance, and previous experience with the equipment. In this priority scheme, the equipment of proven design which has little effect on public health and safety, and plant availability is reviewed last.
15.4.6-39 12-1-69 - Supplier Quality Assurance Preaward Evaluation of Prospective Suppliers Prior to considering a new supplier for placement of a purchase order, a supplier evaluation is conducted. This is*done in accordance with a written check list. The results are documented in .a report issued to management personnel of Purchasing, Engineering, Quality Assurance, and Projects. The evaluation is conducted by a team consisting of personnel from Purchasing, Engineering and Quality Assurance. Other personnel, such as material and process engineers and manufacturing engineers, participate as required. The applicant is kept inf6rmed of changes in the status of bidders. Considerations of the evaluation include: Previous experience with the supplier Physical plant facilities Quality control program and system Number and experience of design personnel Material control and raw material inspection In-process inspection Assembly and test capability Tool and gage control
15.4.6-40 12-1-69 Special processes required Nondestructive testing Inspection and test equipment Records function Deficiencies in the supplier's organization or systems are resolved with the supplier's management prior to placing a purchase order. If an existing supplier does not maintain the required quality level on Westinghouse orders, a similar team will review the supplier's problems and make recommendations to his management to correct the situation immediately. When problems arise, Westinghouse specialists aid the supplier in specific areas such as welding, manufacturing and nondestructive testing to resolve the problem. In this manner, Westinghouse assures the continued high level of supplier performance necessary to obtain the quality level required by the contract. Supplier Quality Control Requirements Quality requirements that apply specifically to a component are contained in the Equipment Specification. Requirements of a quality systems nature, not peculiar to a component, are contained in two standard documents. The first is entitled, "Administrative Specification for the Procurement of Nuclear Steam Supply System Components." This document is applied e
15.4.6-41 12-1-69
- in all component purchase orders. The Administrative Specification requires the supplier not only to manufacture equipment that conforms to purchase order requirements, but to assure himself and Westinghouse by means of appropriate inspections and tests that the equipment conforms to these requirements. The quality control section of this specification contains specific requirements in areas such as:
Calibration of measurement and test equipment Control of drawings, specifications, procedures and other documents used in design or manufacture, and revisions to these documents
- Control and identification of material Maintenance of quality control records, Test control through written test procedures and test records Nonconforming supplies, including identification and control to preclude further use The second document that specifies quality requirements is QCS-1, "Manu-facturer's Quality Control Systems Requirements." This document is
15.4.6-42 12-1-69 applied to orders for more critical equipment such as Class 1 components. This document requires the supplier to maintain an adequate quality control system. This specification meets the intent of Appendix IX of Section III of the ASME Boiler and Pressure Vessel Code in the area of quality control system requirements. QCS-1 requires the following, among other things: Establishment and maintenance of a system for the control of quality that assures that all supplies and services meet all specifications, drawing, and contract requirements. Application of the system to subcontracted items. Written procedures that implement the system. Qualification of a personnel. Qualification and control of processes including welding, heat treating, nondestructive testing, quality audits and inspection techniques. Operation under a controlled manufacturing system such as process sheets, travelers, etc. e
15.4.6-43 12-1-69
- Written inspection plans for in-process and final inspection.
Submittal of Inspection Check Lists for approval by Westinghouse; these check lists show inspection and test status. Recording of results of each inspection operation. Repair procedures, with provision for Westinghouse approval of all procedures utilizing operations not performed in the normal manufacturing sequence. Written work and inspection instructions for handling,
- storage, shipping, preservation and packaging.
As required, inspection hold points are specified by Westinghouse in the Equipment Specification or elsewhere in the purchase order. These are points of witness or inspection by Westinghouse beyond which work may not proceed without approval by Westinghouse. Hold points are reviewed by the applicant and major points of inspection are witnessed by the applicant or his assigned agent. Planning of Supplier Surveillance Westinghouse PWR surveillance of suppliers during fabrication, inspection, testing and shipment of components is planned in advance and performed
15.4.6-44 12-1-69 in accordance with written Quality Control Plans. These plans are prepared by Quality Assurance engineers and are based on the technical requirements of the purchase order. The plans are reviewed and approved by engineering. The purpose of a Quality Control Plan is to provide planned guidance to the Quality Assurance field representative by (1) focusing attention on those items which contribute most to quality and reliability, and (2) providing specific instructions for the witnessing, documentation, and acceptance of the equipment, and for auditing to assure the supplier's compliance with all quality control requirements. The plan identifies the points during manufacturing and test that Quality Assurance intends to witness. The plan covers (1) the auditing of the supplier's quality control system and operation procedures; (2) surveillance of key operations such as welding, nondestructive testing, production and nonoperating electrical testing; and (3) inspection verification (for example, sampling review of radiographs, material test reports, key dimensions, and operating electric tests). Special emphasis is placed on the aspects of manufacture and inspection that most directly affect performance of the equipment. Lead units of a new design get particular attention in the supplier's shop by botr,.. Quality Assurance and E;ngineer.ing representc).tives. When surveillance is indicated, Quality Assurance develops a visit schedule depending on the supplier's performance. Visits. are more frequent during
- 15. 4 .6-:45 12-1-69
- the initial stages of manufacture, particularly to a new supplier, with frequency diminishing as the supplier demonstrates his capability.
Surveillance of Suppliers The purpose of Westinghouse surveillance of suppliers is to provide Westi~ghouse management and applicant first-hand objective assurance of compliance with specified requirements. The principle followed is that the supplier is responsible for inspecting and testing his product. The Westinghouse field representative assures that the supplier has done this, rather than attempting to perform the supplier's inspection for him or duplicate the work he has d:me *
- The frequency and scope of Westinghouse surveillance varies with degree of importance of equipment, supplier perfoi;mance, complexity of the component, and other factors. This determination is made by Quality Assurance in conjunction with engineering. Quality Assurance Residents are established as necessary *.
Surveillance is accomplished in accordance with Quality Control Plans. In addition, the field representative confirms on a continuing basis that the supplier's system is adequate to ensure that a quality product will be built. He sees that written instructions and procedures are kept current, that application of drawings and specifications is controlled, that corrective action is imp1emented, and that other necessary controls are effective.
lS.4.6-46 12-1-69. The Quality Assurance representative informs the supplier directly of problems he discovers and obtains commitments to correct them.* He brings these problems to the attention of the supplier's management as required to obtain resolution. Release of Equipment for.Shipment The Administrative Specification requires the supplier to write a formal shipping release when he_ is satisfied that purchase order requirements have been met. When the Westinghouse Quality Assurance representative is satisfied.that the equipment can be released for shipment, and after receipt of the supplier's release, he prepares a Quality Control Release form, and distributes copies to the supplier and cognizant personnel within Westinghouse. The equipment can then be released through normal engineering-purchasing channeis for shipment. The supplier forwards the Westinghouse Quality Control Release with the equipment to the plant site. Construction Site Quality Assurance Control of Site Work Work on nuclear steam supply equipment as performed by the construction contractor and subcontractors, is monitored by W representatives assigned to the construction site. The necessary procedures and actions are
15 .4. 6-'47 12-1-69
- coordinated with the construction site representatives of the app*licant and the construction contractor. Special processes such as welding, cleaning and nondestructive testing are observed by qualified Westinghouse personnel to'assure the work is performed in accordance with written procedures.
I. During component installation, Westinghouse Nuclear Power Service monitors work on nuclear steam supply and engineered safeguards equipment. Qualified personnel provide technical advice on various disciplines
*of construction such as welding, mechanical and electrical systems, instrumentation and control equipment*, and start-up .
- , The construction contractor is *responsible for overseeing that the Westinghouse nuclear s.te*am supply equipment is in good condition when received and that it is stored, handled and installed properly according to applicable specificatiO'ns, procedures, and manufacturers' instructions.
A written p.rocedure describes the system for identifying, reporting and obtaining disposition of nonconforming material or ec,_uipment discovered at the site. Nuclear Power Service personnel fill out a Field Deficiency Report to provide the cognizant engineering group with the information necessary for making.proper and timely disposition of each problem. After the cognizant personnel make a disposition, it is noted on the Field Deficiency Report and returned to the field for action. Files of these reports are maintained tp record all field deficiencies and
- to provide for long-term corrective action.'
15.4.6-48 12-1-69 Qualification of Westinghouse Personnel Nuclear Power Service welding engineers are qualified to Level II per SNT-TC-lA, NDT Personnel Qualification Recommended Practices, in the same manner as required by the ASME Boiler and Pressure Vessel Code, Section III, Appendix IX. All Nuclear Power Service personnel are experienced in their disciplin~s and are familiar with the codes, specifications and process procedures pertaining to their respective scope of work *. Quality Control Records The Administrative Specification described above requires suppliers to maintain records for each test (nondestructive, electrical, performance) specified in the purchase order. The record must show the test procedure, equipment and materials used, the acceptance standards applied, and the test results obtained. The p?rt.or assembly tested, date of test, and test operator identity is shown. The administrative specification and equipment specification also require maintenance of other records as required, such as material test reports, welder qualifications, inspe~tion records, etc. Records such as trip reports, deviation notices, and other quality-related documents form a part of the Quality Assurance records maintained by Westinghouse, and are subject to applicant's reviews and audits.
15
- 4 .,5_,:49 12-1...-69
- Suppliers are required to maintain these records for specified periods, after which they notify Westinghouse in order that a record file for the life of the plant can be arranged. Suppliers are also required to transmit records to Westinghouse as work is completed for added assurance of record availability.
After completion of construction and erection work, those records required by the applicant to meet the continuing regulatory responsibilities for operating the .station are turned over to the applicant. Records generated at the construction site are filed and maintained there, and are available for applicant's review during the course of
- the project.
Nonconforming Material, Trend Analysis and Corrective Action Deficiencies at Suppliers' Plants The Administrative Specification and QCS-1, described above, contain
.specific contractual requirements for controlling nonconforming material or workmanship.
The supplier must physically identify all material that does not conform to purchase order requirements and take necessary actions to preclude its further use. All deviations are documented in writing and reviewed
15.4.6-50 12-1..-69 by engineering, quality assurance and other appropriate groups. First, consideration is given to restoring the material to its specified condition or scrapping it. If this is not applicable, the deviation is considered from both an engineering and a quality control. point of view. If acceptable, the deviation is formally approved in writing by the cognizant engineer. A permanent file of these records is maintained. QCS-1 requires tQat the supplier's quality system provides for the identifi-cation and evaluation of significant or recurring discrepancies and for alerting the supplier's cognizant management to the need for corrective action. The supplier must review corrective action for effectiveness and the need for further action. Deficiencies at the Construction Site A written procedure provides for documented reporting of deficiencies found during plant construction. These reports are submitted by site engineering personnel to the cognizant engineering department. Like
. I reports.from suppliers' plants, these reports are reviewed for necessary action, formally approved by the cognizant engineer and permanently filed.
Trend Analysis and Corrective Action Plant Quality Assurance analyzes all significant deficiency data on Westinghouse-supplied equipment received from suppliers and from
- 15. 4. 6-*51 12-1-69
- construction sites to determine patterns of occurrence by supplier, by component, or by process. With this as a guide, Quality Assurance and cognizant engineers determine corrective actions that are needed to_prevent recurrence. This action is in addition to assuring that the supplier or site personnel take corrective action of the individual deficiencies reported.
Audits
§uppliers' Plants The Westinghouse audit function of suppliers is described in the section, '"Supplier Quality Assuran,ce", above.
_Construction Site Plant Quality Assurance is responsible for conducting independent audits of Nuclear Steam Supply System work at the construction site to assure that proper procedures and instructions are available and in use, and that adequate controls exist and are effective. Reports of audits are sent to top management OF the PWR Systems Division. Westinghouse Nuclear Energy System (WNES) Divisions Audits The WNES Divisions which are supplying products or equipment to other WNES divisions are audited by those divisions. Electromechanical, Tampa
15.4.6-52 12-1-69 and Pensacola Divisions,are audited by PWR Systems Division, and Specialty Metals Divi"sion by Tampa and Nuclear Fuel Dh*isions ~*
- The process audits are focused on basic functional areas such as welding, nondestructive testing, *personnel qualifications, procedures, and so forth. The purpose of the audits is to assure by an independent evaluation that the detail implementation of the divisions' quality assurance programs is fully effective.
Division Quality Assurance personnel conduct these audits. The frequency of the audits varies currently from monthly to bi-monthly depending on the need. Westinghouse Nuclear Energy Systems Quality Assurance Committee Audits The Westinghouse Nuclear Energy Systems (WNES) Quality Assurance Committee is composed of quality assurance managers from each division. The committee members are appointed by the WNES Executive Vice President. The chairman of the committee is presently the PWR Systems Division Quality Assurance Manager. The WNES Quality Assurance Committee has established an audit program which applies to all the Westinghouse divisions engaged in nuclear supply system design or manufacture of PWR equipment. The purpose of the audits is to provide in-depth evaluation of the quality assurance policies
15.4.6-53 12-1-69
- and processes of the various NES divisions in order to verify that they result in products and services which meet safety and reliability requirements.
I Particular emphasis during the audit of the quality assurance programs is placed on compliance with safety requirements. In addition to carrying out audits, the committee serves as a forum to communicate quality and reliability activities, and to establish improved and consistent division policies of quality assurance in light of nuclear industry requirements. Annual q_uality assurance system audits are conducted of each WNES division by an audit team composed of representatives from the Committee. Typical
- * *team membership i$ three men. Each audit normally takes three days.
The Corporate Headquarters audits, described below, of a WNES division substitutes for the annual WNES audit the year it is held. At the conclusion of each audit an oral presentation is made by the auqit team to the division General Manager and Quality Assurance Manager of the division which has been audited. Following t~e audit, a written report containing the findings of the audit and recommendations for improvement in the quality assurance program and its implementation is sent to the responsible division personnel, to the committee members, and to the WNES Executive Vice President. This procedure assures high level management attention to actions needed to carry out recommendations of the audit.
15.4.6-54 12-1-69 Westinghouse Corporate Audits The Westinghouse Headquarters Quality Control Staff has a formal audit program which applies to all divisions in Westinghouse including divisions furnishing equipment or services to the nuclear industry. The purpose of the audits is to provide an independent verification that the quality assurance programs of the Westinghouse divisions are effectively assuring that the product quality complies with the requirements of their customers and that the programs include the most effective approaches to prevent the manufacture of defective product. Audits are performed of each division's quality assurance effort 9y a two man team, consisting of a member of the Headquarters Quality Control staff and the Quality Control Manager of another division in the same product group as the division audited. The audit normally takes five days. The Corporate Headquarters audit of each Westinghouse division is held on the average of once each three years. The quality assurance systems and procedures that have been established by the division are reviewed to determine if these systems and procedures are sufficient to provide an effective program. Observations are then made to assure that the established systems and procedures are being correctly followed.
15.4.6-55 12-1-69
- An oral presentation of the findings and conclusions of the audit is made to the Division General Manager, Quality Assurance Manager, and other personnel affected by the audit findings. The items recommended £or improvement in the quality assurance program are presented as well as recommendations of approaches for accomplishing these improvements.
Following the audit, a written report containing the findings and recommendations reviewed in the oral report is prepared and sent to the responsible division personnel. In addition, a copy of the report is sent to the Executive Vice President to whom the division reports and to the corporate Vice President of Manufacturing .
FIG. 15.4-1 FEB. 15, 1971 CONTRACTORS STONE .6. WEBSTER ENGINEERING CORPORATION VIRGINIA ELECTRIC AND POWER COMPANY l l I I _J I
----------------- I I r - --------1----------_J PRESIDENT I I I
I SENIOR VICE PRES !DENT I MANAGER POWER I POWER PRODUCTION INOUSTRY I w u I lL I 0 lL I w I ::E 0 I VICE PRESIDENT ENGINEERING DIRECTOR DIRECTOR VICE PRESIDENT MANAGER DIRECTOR
& SENIOR ASSURANCE POWER STATION PRODUCTION
- 6. ENGINEERING QUALITY NUCLEAR CONSTRUCTION REVIEW DESIGN & OPERATIONS e; MANAGER ASSURANCE SERVICES COMMITTEE CONSTRUCTION MAINTENANCE WESTINGHOUSE PROJECT MANAGER QUALITY ASSURANCE COORDINATOR QUALITY CONTROL NUCLEAR CHIEF CHIEF CHIEF
& POWER CONSTRUCTION PROJECT QUALITY FIELD QUALITY QUALITY CONTROL SUPERVISOR STAFF e.
RELIABILITY SERVICE POWER STATION ASSOCIATE MANAGER ENGINEER ASSURANCE CONTROL SHOP CONSTRUCTION ENGINEERS MANAGER MANAGER ENGINEER ENGINEER INSPECTOR r-1
--1-~
___J r - DISTRICT- "l RESIDENT ENGINEER
---+
I GENERAL SUPERINTENDENT SENIOR QUALITY CONTROL I QUALITY CONTROL I I I I I I CONSTRUCTION ENGINEER L ENGINEERS e. ___::!SPECTORS_ J 0 __J I ,------------~---- ------- w lL I
,-_l __
WESTINGHOUSE l rsTONE& WEBSTER'\ QUALITY CONTROL ENGINEERS 0 z
<f I EQUIPMENT I LEAD I EQUIPMENT CONSTRUCTION MECHANICAL w 1-I SUB I ENGINEER I SUB I SUPERVISORS ELECTRICAL U)
L ~NTRACTOR~ _J L ~NTRACTORS _J & CIVIL r - - -, I I FIELD LOCATION OTHER THAN SITE L----' LINE OF RESPONSIBILITY SERVICE QUALITY CONTROL COMMERCIAL ENGINEERS LINE OF ACCOUNTABILITY FOR QUALITY ENGINEERS INSPECTORS LABS LINE OF COMMUNICATION PRODUCTION PRODUCTION CRAFTS CRAFTS (FOREMEN I (FOREMEN) QUALITY ASSURANCE PROJECT ORGANIZATION SURRY POWER STATION
.WESTINGHOUSE POWER SYSTEMS PRESIDENT WESTINGHOUSE NUCLEJJl ENERGY SYSTEMS
~_gd_rn:_iicii_)imm:.
PWR NUCLE;\R FUEL LARGE COMPONENT SYSTEMS DIVISION -_ DIVISION DIVISION GENERAL MA.~AGER GOORA!c MANM;EiZ GENERAL MA.~AGER ELECTROMECHANICAL ENGINEERING PROJECTS MANUFACTURING DIViSION -- .. .. GENE MA.~AGER CORE ENGINEERING PURCHASING QUALITY MA.~UFACTUlUIIG MANUFACTURING ENGINEERING ASSURA.'ICE CONTROL ~D OPERATIO~S PROJECT ELECTRICAL RELIABILITY .t~-.iNitlG A..itO. MANAGER SYSTEMS :RELIABILITY PLANT CONSTRUCTION FUEL QUALITY
"°APPARATUS AND SERVICES RELIABILITY ASSu'RA.~CE MATERili:.S AND -NUCLEAR POWER TAMPA DIVISION>
PROCESSEii SERVICE GENERAL . MA..~AGER SYSTEMS. E"iiGINEERING MA..'IUFACTURING LICENSING. AND RELIABILifr. QUALITY-- RELIAiffLffl* RELIABILITY ASSURANCE UCENSlN!f j-J PRESS_IDIE VESSELS NC:
"I ~Q!1_ALITY CONTROL
....I .,,
I :,o PLANT~ALITY a, ... ASS C!* ""... I N MEC!!AfUCAl. -
.EOUil'MEBT
* .' :*\ *,
*-::** nGUB.E 15.4-3 NUCLEAR ENERGY SYSTEMS FUNCTIONAL GROUPS QUALITY ASSURANCE FLOW CHART Originating Flow Functional Path No. Flow Path Definition Group A B D E F G H I K L 0 Design and Construction Follow and Approval A A B C D E F G H I K L (to and from all gToups) 1 Contractual Requirements A B C D E 2 Safety Requirements B D E L 3 NSSS Functional Requirements C D L 4 System Design D B C F H L S(a) Concurrence on System Design BLH D (Safety Requirements)
S(b) Concurrence on System Design J) (Functional Requirements) 6 Design Review (Selected Critical Areas) F C D E G 7' Equipment Functional Requirements D E t 8 System Functional Requirements D H K L 9 Quality Control Plan F E L 10 Preliminary Equipment Specification E B D F G H K L (including Quality Control Requirements) 11 Concurrence on Preliminary Equipmen~ Specifications BDFGKLH E 12 Fidal Equipment Specificat~ons E D F H I K L 13* Design, Manufacturing and Inspection (a) I E
*Procedure (b) E F G L
- This flow path is in two steps. In the first step (a) information goes to a functional gToup which becomes the originating gTOup in the second step (b).
FIGURE 15 .4-l (Continue~ NUCLEAR ENERGY SYSTEMS FUNCTIONAL GROUPS QUALITY ASSURANCE FLOW CHART . ~ t.:> 0 zH t Cl.I ~ Cl.I
~ iii
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i Originating 11< i:,::i 11< i:,::i Cl.I OH Flow Functional Path No. Flow Path Definition Group A B C D E F G H I K L 14* Concurrence of Design, Manufacturing and (a) FGL E Inspection Procedures (b) E I 15(a) Surveillance, Process Audits and Quality F I Control Plan Actions 15(b) Corrective Action, Follow through on Deficiencies F" E K L (Fabrication and Field) 16* Materials Evaluation (a)I E (b)E G L 17* Materials Concurrence (a)GL E (b)E I 18 System Layout H B D E K L 19 Concurrence on ~ystem Layout BDEKL B 20 Fina1*system Layout and Field Follow B L 21 Inspection and Erection Procedures H E F G K L 22 Concurrence on Inspection and Erection Procedures EFKGL H 23 Final Equipment !nspection and Erection Procedures H K 24 Equipment for Receipt Inspection I L .... "ii NH I c., 25 Erection Follow
- K H L i~
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I 27 Pr~liminary Testing Requirements and Processes D B C E K L I,,)
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CJ ,-:i oo 0 I tll Originating ~ ~ iil ~ iil >a po: tll C, l'.:'.l ~ &c., l'.:'.l tll ~~ ~ Flow Functional Path No. Flow Path Definition. Group A B C D E F G H I K L 28 Concurrence of Testing Requirements and Processes BCEl\'.L D 29 Final Testing Requirements and ~rocesses D K L 30 Operating Test General Procedures and K D L Technical Direction 31* System Performance Test Results (a} L K (b} K B C D 32 Concurrence on System Performance BCDK L 33 .Handing over of Completed Plant K L 34 Preliminary Operating and D B C E K L Emergency Instructions 35 Concurrence on Preliminary Operating BCELK D and Emergency Instructions 36 Final Operating and Emergency Instructions D K L 37 Preliminary Technical Specifications B C D E G L 38 Concu:rren'ce on Preliminary Technical CDEGL B
..... "I Specifications l'JH I c;')
1--' c::: I ::It! 39 Final Technical Specifications B L a- t>l
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~o Plant Operation Information L K .
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- See note on f:l,:i;~t 11heet of t h:lij t<lb le
- 15.5.1 -1 4-15-70
- 15.5 SPECIFIC CONTAINMENT STRUCTURAL DESIGNS 15.5.l CONTAINMENT STRUCTURE .
. 15.5.1.1 General For arrangement of the containment structure, see Figs. 15.1-3, -4, -5, -6, -7, -8 and -10.
Each of the reactor containment structures are similar in design and construc-tion to that of the Connecticut Yankee Atomic Power Plant at Haddam, Connecticut. Each is a steel lined, heavily reinforced concrete structure with verti,~al cylindrical wall and hemispherical dome supported on a flat base mat. Below grade, the containment structures are constructed inside a cofferdam of steel sheet piling. The structures are soil supported. The base of the foundation mats are located approximately 66 ft below finished ground grade. Each containment structure has an inside diameter of 126 ft-0 in. The spring line of the dome is 122 ft-1 in. above the top of the foundation mat. The inside radius of the dome is 63 ft-0 in. The interior vertical height is 185 ft-1 inch, and the base mat is 10 ft-0 in. thick. The steel liner for the wall is 3/8 in. thick except over the base mat where 1/4 in. and 3/4 in. plate is used. The steel liner for the dome is 1/2 in. thick. A waterproof membrane, as shown in Fig. 15.5.1.1-1, is placed below the containment structural mat and carried up the containment wall to ground level. Attached to and entirely enveloping the part of the structure below grade, the membrance protects the
15.5.1-2 12-1-69 structure from the effects of ground water and the steel liner from external hydrostatic pressure. Ground water immediately adjacent to the containment e st-ructure is kept below the top surface .of the foundation mat by pumping as required. Access to the containment structure is provided by a 7 *ft-0 in. ID personnel hatch penetration and a 14 ft-6 in. ID equipment hatch penetration. Other smaller containment structure penetrations include hot and cold pipes, main steam and feedwater pipes, fuel transfer tube, and electrical conductors. The reinforced concrete structure has been designed to withstand all loadings and stresses anticipated during the operation and.lifeof the unit. The steel lining is attached to and supported by the concrete. The liner functions primarily as a gastight membrane and transmits incident loads to the concrete. The containment structure does not require the participation of the liner as a structural component. No credit-has been taken for the presence of the steel liner in designing the containment structure to resist seismic force or other design loads. The steel wall and dome liners* are protected from potential interior missiles by interior concrete shield walls. The base mat liner is protected by a 1 1/2 to 2 ft thick concrete cover, except where a 3/4 in. thick liner plate was used beneath the reactor vessel incore instrumentation and at a drainage trench where floor grating provides additional protection.
- 15. 5. i-3 12-1-69
- As an added precaution against water seepage which might penetrate the wate:r-proofing membrane in small quantities, pipe sumps are provided in each of the instrument observation pits located outside the cylindrical wall of the con-tainment but within the waterproofing membrane. The sumps penetrate the base
-----~=-*---n,,-...,."""~'J>....... -,r.o_,,,_,:.:~-... ..,.... ~.............,:.,~__.
mat and terminate in the porous concrete immedi-ately below the mat.
**-*--*-----.... - - . =~-.,._______-~"-~ . *.~ __..,.,. ,-~~---.. ,-m=-ce~
Pumps are provided to remove ground water outside the wa*terproofing membrane as described in Section 15. 5 .1. 3
- 15 .5.1. 2-1 12-1-69
- 15. 5 .1. 2 Design Criteria The design of the .containment structures is based on:
- 1. Biological shielding requirements
- 2. The temperature and pressure generated by the Design Basis Accident (DBA), Section 14.5.2
- 3. The Operational and Design Basis Earthquake discussed in Section 2.5
- 4. Severe weather phenomena
- 5. The maximum calculated power level of 2,546 MWt The DBA was selected as the design basis for the containment structure since all other bases would result in lower temperatures and pressures. The contain- **
ment structure is also designed for the normal subatmospheric operating condi-tions. Fu~ther, the containment structure is designed for a leakage rate not to exceed 0.1 percent of the con.tain~d volume per day at 45 psig. The normai operating pressure for the containment is approximately 9.0 psia partial air pressure with about 1.0 psia additional partial water vapor pressure. The resulting total containment pressure is approximately 10.0 +/-0.5 psia. The temperature of the containment air fluctuates between a maximum. temperature of 105 0 F and a minimum of 60 0 F during normal operation and 60 0 F
- during shutdown, depending upon the ambient temperature of available service water. The normal
'operating pressure allows accessibility for inspection and minor maintenance
15.5.1.2-2 12-1-69
- during ope.ration without requiring containment pressurization or the use of supplementary breathing equipment for personnel. The air partial pressure corresponds to that at an atmospheric elevation of approximately 10,000 ft.
The containment structure is designed by ultimate strength methods conforming to ACI 318-6'3, Part IV-B. Design load criteria based on AC! requirements and others given below conform to current containment design. The ultimate load capacity of the containment structure as modified by the safety provisions of AC! 318-63, Section 1504, is not less than that required to meet the containment structural loading criteria. Loads imposed on the containment shell design include:
- 1. Dead load
- 2. DBA pressur~
- 3. Temperature rise in liner associated with DBA
- 4. Normal operating temperature gradients
- s. Earthquake
- 6. Wind loads, including tomado winds Loads imposed on. the containment mat design include:
- 1. Mat and interior structures during construction
- 2. Dead load for complete structure and contents
15.5.1.2-3 12-1-69 3. 4. 5. Dead load and DBA pressure and liner loading Dead load, DBA pressure, liner loading, and earthquake Dead load and earthquake The ultimate load capacity of the containment structure as modified by the safety provisions of AC! 318-63, Section 1504, is not less than that required to satisfy the following structural loading criteria tabulated in Table 15.5.1.2-1.
- 15. 5.1.2-4 4-15-70 TABLE 15.5.1.2-1 CONTAINMENT STRUCTURAL LOADING CRITERIA Loading Required Load Capacity Case Combination of Structure
- 1. Operating plus DBA = (1.0 +/- O.OS)D +I.SP+ 1.0 (T + TL)
- 2. Operating plus DBA plus operational basis earth-quake = (LO +/- O.OS)D + l.OP + 1.0 (! + TL) + I.SE
- 3. Operating plus DBA plus design basis earthquake = (LO +/- O.OS)D +I.OP+ 1.0 (! + TL) + l.OHE
- 4. Operating plus 1.25 DBA and 1.25 operational basis earthquake (LO +/- O.OS)D + (1. 25P) + (T' +TL')+ l.25E
- 5. Operating plus tornado loading = (LO +/- O.OS)D + l.OT' + l.OC C - Load due to negative pressure and horizontal wind velocity resulting from tornado and missiles. For description of tornado, refer to Section 15. 2. 3
- D - De'ad load of structure and contents including effect of earth and hydrostatic pressures, buoyancy, ice and snow loads. To provide for variations in the assl.lliled dead load, the coefficient for the dead load components is adjusted by +/-5 percent as .indicated in the above formulas to provide the maximum stress levels.
P - Pressure load from DBA. Pressure for containment design is 45 psig as described in Section 14.5.5.
15.5.1.2-5 12-1-69 T Load due to maximum temperature gradient through the con-crete shall and mat based on temperature associated with 1.5 DBA pressure. T' Load due to maximum temperature gradient through the con-crete shell and mat based on normal operating temperature. TL ~ Load exerted by th~ exposed liner based upon temperature associated with 1.5 times DBA pressure. TL' - Load exerted by the exposed liner based upon temperatures associated with 1.25 times DBA pressure. T Load due to maximum temperature gradient through the con-crete shell and mat based upon temperature associates with 1.0 times DBA pressure. TL Load exerted by the exposed liner based upon temperature associated with 1.0 times DBA pressure. E Operational basis earthquake loading. Based on a ground acceleration of O.07 g horizontally at ze.ro period and a damping factor of 5 percent. For description of the opera-tional basis earthquake, refer to Section 2.5.
15 .5. 1.2-6 12-1-69 HE Design Basis Earthquakia loadiµ..g. . B_ased on a ground acceleration of O.15., g. horizontally* at. zero period and a dSIJ1ping factor of 10 percent. For description of the
*Design Basis Earthquake, refer to Section 2.5.
Note: Normal wind loadings replace earthquake loads where they* exceed earthquake loadings. Normal wind or tornado loads are not' considered' ~oirtcident with earthquake loads. The seismic design coefficients and critical damping fac,tors used in the design of the reactor containment structure are given in Section 15. 5 .1. 4. The average ,i ac*celeratiori spectra* ~tirves are incfuded in Section 2 .5. The earthquake loads include the horizontal or vertical acceleration or a combination of both where the effects, as measured by the stresses resulting from the separate accelera-tion components, of horizontal.and vertical ground accelerations are combined algebraically.
- The load capacity of the tension members is based on the guaranteed minimum yield strength of the reinforcing steel. Load ~apacitt~s of flexural and compressfon tilethbei:s are determined in. accordance with the Building Code Require-ments for Reinforced Conc~ete: A.CI 318; . The load capacity so determined is decreased by ,a reduction factor multiplier "<J", to compensate for small adverse variations in material and workmanship. The reduction factors are listed in Table 15.5.1.2-2.
15 .5 .1.2-7 2-13-70 TABLE 15. 5 .1. 2-2 CAPACITY REDUCTION FACTOR FOR CONCRETE Member Reduction Factor
- a. Tension and flexure 0.90
- b. Diagonal tension, bond and anchorage 0.85 The load capacity reduction factor for stresses in concrete produced by tornado-carried missiles, in combination with other tornado-produced stresses as given in Loading Criteria 5, is 0.75.
The dominant design load is the 45 psig containment design pressure which creates major tensile membrane stresses in the reinforcing steel, coincident with moments at the junction of the containment wall and mat. The design tornado wind loading and pressure drop criteria are stated in Sections 2.2 and 15.2.3. Since the DBA pressure load is greater than the negative pressure load of tornadoes, the containment structure is able to maintain its integrity and permit an orderly shutdown on the reactor unit should a tornado strike the structure.
15 .5 .1. 3-1 12-1-69 15.5.1. 3 Bouyan,t Loads Yard elevation is at +26 ft-6 in.; the base of the containment mat is at El. -39 ft-7 in. Six seepage drains are provided to drain the area beneath the containment .structure. Four drains extend down to El. -65.0 and two drains extend down to El. -105.0. These drains terminate in a 12 in. thick crushed rock lc!Yer placed immediately below the mat and through which water can travel to the edge of the cofferdam. Seepage from these drains and other seepage into the cofferdam collect inside the cofferdam around the base of the mat. Three permanent deep well multistage vertical submersible dewatering pumps are located in separate pipe casing extending from finished grade to below the containment structure mat. The pump suction inlets are located in an area of
- porous concrete whi<rh extends from El. -42.42 *to El. -21.5.
remove all subsurface seepage wate~. These puJtlps*.: *, The pumps are controlled to maintain the water level in this space between a high .of El. -30.1 and a low of El. -35.5, a range of 5.4 ft, which is equivalent to a fluctuation in buoyant pressure under the structure of +/-170 lb per sq ft from the mean value. Alarms sound if the water exceeds El. -29.5. The dead load of the structure and its contents is 7,200 lb per sq ft. This fluctuation in buoyant pressure amounts to 2.3 percent of the dead load weight. In the unlikely event of multiple pump failure for a sufficient period of time for the ground water to rise to finished ground grade at El. +26. 5, the buoyant pressure would increase to a maximum value of 4,150 lb per sq ft which amounts
- to less than 60 percent of the dead load of the structure.
tion of the *containment is not credible. Therefore, flota-
15.S. L4-l 12-1-69 15.5.1.4 Dynamic Analysi~ Analyses were conducted to determine response stresses in the containment structure due to the application of seismic loading. Earthquake ground motion I values were applied simultaneously in the horizontal and vertical directions~ Vertical ground motions were assigned a magnitude equal to two-thirds of the horizontal motions. The magnitudes of the Operational Basis Earthquake ~{0BE) and the 0 Design Basis Earthquake (DBE) are derived and assigned as described in Section 2.5. Design loading conditions combined with seismic loading and allowable stress levels are stated in Section 15.5.1.2. The earthquake loading was analyzed using a Stone & Webster program, "Container Vessel Seismic Analysis," based upon the dynamic analysis of a containment structure by Messrs. Hansen, Holley and Biggs of M.I.T. The general analytical model of the containment structure responding to hori-zontal earthquake forces is a coupled two mass system in which the wall and dome comprise one mass and the base slab and internals comprise the second mass. This model responds to three degress of freedom; flexure in the wall and dome, translation, and rocking of the structure as a unit. The model includes the first three modes of vibration. The stiffness of the wall and dome was obtained through formulas recommended by Professor R. V. Whitman of M.I.T., based on work by G.N. Bycroft.
15.5.1.4-2 12-1-69
- The output of the computer program was spot checked by manual analysis which confirmed the program basis.
Another independent manual analysis which considered the internals as a third coupled mass resulted in loading values which were not greater than those obtained from the analysis of the two mass system. A preliminary analysis of response to vertical earthquake forces using a singie mass system showed that these forces are not controlling factors in the design. When computing the response of the reinforced concrete containment structure to earthquake forces, the value of 5 percent of critical damping was used with the
- design earthquake acceleration of 0.07 g. This is an over-all value which includ!!!S the damping 1-n both the reinforced concrete structure and the soil.
The magnitudes of earthquake forces applied to the structure were obtained from the response spectrum for 0.07 g at zero period and 5 percent critical damping at the. calculated frequency of the structure and then distributed over the structure in accordance with the relative motions of the structure as determined by dynamic analysis. The forces derived by use of this damping* factor were used for the entire reinforced concrete containment. The value of 5 percent of critical damping, together wi.th the. damping: factors for other systems, structures, and equipment*,* is listed in Table 15.2.4-1.
15.5.1.4-3 12-1-69 The value of 10 percent of critical damping was used with the Design Basis Earthquake of 0.15 g on the basis.of increased cracking in the concrete and increased movement in the concrete and soil. To verify the damping used for design, an analysis *of the soil structure inter-action damping was made in accordance with the procedures suggested in "Analysis of Foundation Vibrations" by Robert V. Whitman, Proceedings of a Symposium
~
- organized by the British National Section of the International Association for Earthquake Emergency.
Damping factors for soil were calculated for the rigid body translation and rocking. Flexure damping was assessed as suggested by Newmark. For each of the four modes of vibration, energy losses in structural flexure, sliding, and rocking were calculated and proportioned to determine the total system energy loss, thereby defining the damping to be used in spectrum response. This analysis demonstrated that the damping factors used for design and the resulting seismic response characteristics are conservative. Earthquake load criteria are included in the loading criteria described in Section 15.5.1.2. Operating and Design Basis Earthquake factors are each combined with other loads, including the Design Basis Accident pressure. Resulting shears are computed by the computer program.
15.5.1.4-4 2-13-70 Lateral earth pressure' under f:i°eismic loadings on .the containment mat was determined by computing the lateral resistance developed in the soil as the structure responds in flexure, translation and*rocking. Iri this analysis, the translational restraining force has two components, a shear across the base of the-structure and lateral soil pressures on the side wall of the containment structure developed by its displacement relative to its static position. The "spring constant," that is force per unit of lateral displc:;1cement by shear, for a circular rigid base on an elastic half space is given by Bycroft 2 as k = 3~(1-ti) Gr 0 X 7-:Su where:. G shear modulus r radius of base 0 u = Poisson's Ratio Note: Values in consistent units For usual values. of u this reduces approximately to: 5Gr . 0 The hprizontal pressure on the side wall of the containment.structure.can be* evaluated from the theories of horizontal subgrade reaction. From Terzaghi 3 the relation between horizontal*deflection and pressure at.a.ny.point is given b:y: p
~1 Yh
- 15. 5.1. 4-5 12-1-69 where: p = horizontal pressure at soil*structure interface y = horizontal deflection of soil at interface
= coefficient of horizontal subgrade reaction
~z further ~ =-B-
.where: **~ =
coefficient dependent upon physical propertie~ of' the soil z = depth below free surface of soil B = width of loaded area, which may be taken as diameter
*of containment structure
. 3 For purposes of this analysis a value of n = 40 .tons per ft was selected from tables presented by Terzaghi. This value is appropriate to dense sand above the ground water table. It is a conservative value .since the higher the coefficient the stiffer the soil and the greater. the loads imposed upon the side walls of the structure.
The rotational, translational, and flexural deflections of the'structure were determined from response analysis and added so as to obtain maximum deflections. The lateral soil pressures on the side wall of the structure were then computed for these total.deflections using the theory of horizontal subgrade reaction. In* determining these pressures, the side wall of the structure was assumed to be rigid radially, since radial deflection of. the side wall woul4 reduce relative soil-structure deflections and thus the soil forces acting upon the structure.
15.5.1.4-6 12-1-69 The analysis was performed for both the Operationa.l Basis Earthquake of 0.97 g and the Design Basis Earthquake of 0.15 g. The analysis for a 0.15 g*earthquake indicates a lateral force of 300 lb per sq ft at El. -8.5 ft which defines approximately the magnitude of this component. It should be noted that these forces, if included in the seismic loadings on the
,?I structure, would reduce the base shear and vertical bending stresses in the shell.
Accordingly, they are not included when computing such stresses in the shell and thereby contribute to the conservatism of the design. Rocking motion of the containment structure was considered in the determination of the natural frequency, the distribution of inertia forces, and in the allllt amplitudes of motions. The containment wales supporting the cofferdam structure do not affect considera-tions of horizontal pressure under seismic loading on the containment wall. Four circular concrete wales originally supported the sheet steel cofferdam in which the containment structure is founded. The top wale, Wale A, has been partially removed at several points to permit completion of adjacent structures; in this condition it does not impose any restraint on the containment* structure. The bottom wale, Wale D, is in the lower plane of the containment mat and below the plane of the wall and offers no restraint. Wale C extends from a height of 4 ft to 8 ft above the base of the wall. Wale B extends -from .. a height of 17 ft-6 in. to 21 ft-6 in. above the base of the wall. These two wales are approximately 3 ft-9 in. from the containment wall, and the space
15.5.1.4-7
.. 9-15-71 between the wales and wall is backfilled with pervio~:a- fill. Under seismic loading, the distribution of the lateral earth pressure through the cofferdam wales would not have any different effect than if these pressures were applied directly to the structure.
References "Design Cirteria for Nuclear Reactors Subjected to Earthquake Hazards," Newmark, Nathan M., May 25, 1967. Bycroft, G.N., "Forced Vibrations of a Rigid Circular Plate on a Semi-infinite Elastic Space and on an Elastic Stratum, 11 Philosophical Transactions, Royal Society, London Series* A, Vol. 248, pp. 327-368. Terzaghi, Karl, "Evaluation of Coefficients of Subgrade Reaction," Geotechnique, Vol. 5, 1955, pp. 297-326.
15 .5 .1.5-1 12-1-69 15.5.1.5 Sta tic Analysis The contaimnent structure was analyzed and designed for all loading conditions combined with load factors as outlined in Section 15.5.1.2. The forces, shears, and mom~nts in the structural shell were obtained from a computer program based on nNwnerical Analysis of Unsymmetrical Bending of Shells of Revolut.ion, 11 by B~ Budiansky and P.P. Radkowski, published in the American Institute of Aeronautics and Astronautics Journal, dated August, 1963. Forces, moments, and shears in the base slab were obtained from a Stone & 11 Webster computer program, Flat Circular Mat Foundations for Nuclear Secondary Containment Structures. 11 The program analyzes ,a flat circular plate supported
- on an elastic foundation and computes the discontinuity stresses at the junction of the mat and cylinder ~nd ~he soil bearings pressure. :,,:
Discontinuity stresses, shears and moments at the junction of the cylinder and mat were determined using an analogy to the Hardy Cross method for distributing fixed-end moments* in continuous frames
- The theoretical fixed-end moments obtained from the.shell and mat canputer analysis were balanced in proportion to the relative s*tiffness of the *mat' and cylinder.
An indepen~ent,
- manual computation, based on "Theory of Plates and Shells," by S. Timoshenko, at a few selected points produces forces, shears, and moments substantially the same as those produced by the computer programs for the shell and the mat *
----~-_J
15.5.1.5-2 12-1-69 The containment shell program used to derive stresses in the shell, assumes an isotropic material. The program does not include considerations of temperature gradients due to the thermal loadings across the containment wall. To compute maximum stresses due to the thermal load, six general strain equa-tions were derived, one equation for each of the four principal areas of reinforcing steel and one for each major axis of the steel liner. These equa-tions relate strain to position, temperature and incident stress for each item considered. To solve these general strain equations six additional equations were used. Four equations for strain compatibility, which equ~te radial and longitudinal strains and two equations for load equilibrium. The solution of these equations for incident conditions gives the stress in each of the prin-cipal areas of reinforcing steel and the stress on the steel liner. These equations permit the thermal stresses to be considered separately without modification of the major shell program. The thermal operating load in the containment.concrete wall combined with incident condition loadings produces a stress difference of approximately 6,000 psi between the reinforcing steel adjacent to the inside face of the wall and the reinforcing steel adjacent to the outs:lde face of the wall. This difference exists in both the longitudinal steel and the hoop reinforcing steel. To permit the addition of these stresses to those obtained from the containment shell program, without exceeding the maximum, the containment shell program stresses are limited to 3,000 psi below the maximum allowable design stress.
15.5.1.5-3 12-1-69
- This approach is considered extremely conservative since it limits the design stress in the interior layers of .reinforcing steel to approximately 6,000 psi, less than the maximmn allowable design stress permitted on the exterior layers of reinforcing steel.
Structural failure cannot occur, however, until the interior reinforcing steel exceeds yield. Up to that point plastic yielding of the outside reinforcing would be controlled by the elastic behavior of the interior steel. In the solution of the general strain equations, the effect of the concrete has* been ignored, since it is assumed to be cracked and incapable of carrying any of the tensile -loads considered. The dead load of the concrete is also ignored,
- as this was found to have little effect on the hoop stresses.
also provides a more conservative result. This assumption The loads.exerted.on the concrete shell by the thermal effects of the exposed-steel liner were obtained from the calculations discussed above. The equivalent pressure, p, equals the hoop stress, f, in the steel liner multiplied by the liner thickness, t, and divided by the radius of the liner r. The computed equivalent pressure associated-with 1.5 times incident pressure equal 5.45 psi. Stiffness factors were used to distribute computed fixed-end moments derived from an an~lysis of the containment cylindrical wall, considered as a shell with a fixed-end moment, and £rom an, analysis of the containment mat, considered as a flat circular plate with uniform fixed-edge moment
- 15 .5.1. 5-4 12-1-69 Stiffness factors for the cylinder were computed fran formulas given in Raymond J. Roark's book "Formulas for Stress and Strain" for long, thin-walled cylinders. Stiffener factors for the mat were computed from* formulas for circular flat plates with uniform edge moment from the same source.
Variation of the modulus of elasticity of the concrete to differentiate between uncracked and. cracked concrete was not considered in determining the stiffness factors chosen. Use of such a variable would modify the distribution of the moments and shear forces to some degree, but we do not believe that this would significantly affect the accuracy of the results. The safety factors. inherent in the present* design would ac.counnodate such small variations
- The actual distribution of the manents and forces at the junction of the wall and mat are a function of the relative stiffness of each member. This is determined by the; design approach used. Provided the total forces are distrib-uted between the two areas under consideration, differences of distribution due to theoretical variations of the theoretical value of Youngs Modulus for con-crete are not considered likely to improve the results beyond the accuracy obtained with the assumptions already used.
The methods for computing soil pressures under the mat were based upon an analogy to E. P. Popov's "Method of Successive Approximations for Beams on an Elastic Foundation" published in the Proceedings of the ASCE, Separate No. 18, dated May, 1950. The program computes the deflection at the center of the mat
- 15. 5.1.5-5 12-1-69
- relative to a point on the mat at the intersection of the center line of the containment shell walls. The elastic curve of the mat deflection is assumed to be parabolic between these two points. Multiplying the deflection by the subgrade spring constant, the program then provides a parabolic soil pressure curve which is combined with the rectangular soil pressure curves to provide final soil pressures under the mat. The subgrade spring constant is derived from Professor R. V. Whitman's formula:
k = __!q_
'11'(1-µ)R Where: k = Spring constant G = Shear modulus of subgrade material R = Radius of mat
- µ = Poissons ratio of subgrade material While the subgrade reaction varies with depth, a single typical value for the reaction was used which is representative of the zone at the level being con-sidered. The shear modulus was computed using a formula developed by Hardin 1
and Black from observed soil samples and substantiated by dynamic triaxial tests of the soil. The stiffness of the soil was also based on work by G.N. Bycroft which is referred. to in the answer to Question S12.2.4 (2). A variable soil pressure conforming to the deformation of the mat was used in determining the stresses in the structure. Maximum wind velocity associated with a tornado is given as 360 mph. This velocity was converted to an equivalent pressure using the formula P = .00256V. 2
15 .5 .1.5-6 12-1-69 Where P = equivalent pressure, lb per sq ft and V = wind velocity, mph~ Wind pressure was distributed over the containment dome in accordance with the methods given in "Wind
,, Stresses in Domes 11 by P. Gondikas and M.G. Salvadori, published in ASCE proceedings No. 2616, dated October, 1960.
Wind pressure was distributed. over the contai.nment cylindrical shell in accordance with the methods given in Wind Forces on Structures" by T. W. Singell, published in ASCE proceedings No. 1710, dated July, 1958. Tornado wind loads were combined with other loads as described in Section
- 15. 5 .1. 2.
An analysis of the containment structure indicated that resulting membrane stresses due to tornado wind loading in the dome reinforcing are less *than 5,000 psi and that discontinuity stresses at the junction of the dome and cylin-der are somewhat less. The wind loading on the cylindrical shell creates bending, direct and shear stresses. The bending and direct stresses in the horizontal reinforcing equal 16,000 psi. An investigation of overturning due to wind shows that the resultant (DL + wind) falls within the Kern point radius of the cylinder indicating that the vertical reinforcing will not be subject to tensile forces from this load.
15 .5 .1.5-7 12-1-69
- Containment torsional loadings from wind were considered negligible in view of the ideal shape of the containment when considered as a torsion resistant shell supplemented by the diagonal reinforcing throughout the walls provided to resist earthquake loads.
The Stone & Webster computer programs for the reactor containment base slab, cylindrical wall and dome use a constant Youngs Modulus of Elasticity and poissons Ratio. No attempt was made to assign varying n\Dllerical values to these factors to differentiate between the relative amount of cracking in different parts of the structure. The output of the mat program furnishes the following information~
- 1. Radial and tangential bending moments and vertical shear at 5 ft intervals along horizontal radii from the center of the mat, spaced at 30 deg intervals
- 2. Discontinuity stresses at the junction of the mat and cylinder
- 3. Soil pressure The output of the shell program furnished forces, shears and momen~s at 1 ft intervals in the* height of the cylindrical wall and at 1 deg intervals in the height of the dome. Similar information is furnished at each of 16 equi-distant points on the circumference of the vessel at each level considered
- 15 .5 .1. 5-8 12-1-69 Scaled load plots obtained from the computer programs for moment, shear, deflec-tion, longitudinal force and hoop tension are shown in Figures 15.5.1.5-1, 2, and 3 for each of three design load conditions. The fourth design load condi-tion did not govern design and is .not r~presented.
The following assumptions were made:
- a. The dead and live structural loads are included in all three of the design load cases.
- b. Pressure load, factored and unfactored, is the dominant load condition.
- c. Wind loading is not considered to be applied coincident with earthquake, and wind loading replaces earthquake loads where wind loads exceed d.
earthquake loads. Tornado loads are included under the general category of wind loads discussed above.
- e. Buoyant water loads as discussed in Section 15.5.1.3 are substantially less than dead loads.
- f. E~rthquake loads, both for the Operational Basis Earthquake and the Design Basis Earthquake, are included in the analysis.
- g. Thermal load from the liner is converted into an equivalent pressure and added to the incident pressure load wh~n computing moments, shear, and tension associated with the Design Basis Accident.
- h. Thermal load from the concrete is discussed in Section 15.5.1.5.
Stresses resulting from this load are combined wi,th incident pressure load stresses.
15 .5 .1. 5-9 12-1-69
- 1.
REFERENCES B.O. Hardin and W.L. Black, "Vibration Modulus of Normally Consolidated Clay," paper prepared for the Symposium on Wave Propagation and Dynamic Properties of Soils at the University of New Mexico in August, 1967
- 2. Surry Preliminary Safety Analysis Report, Supplement
15.5.1.6-1 12-1-69 15.5.1.6 Reinforcing Steel Arrangement The foundation mat of the containment structure is reinforced with both top and bottom layers of reinforcing. Bottom mat reinforcing is placed ~n a rectangular grid pattern with layers at 90 deg to each other. Reinforcing for the top of the mat consists of concentric circU:lar bars combined with radial bars. The reinforcement pattern for the top of the mat is arranged to permit maintaining a uniform spacing of the vertical wall rebars which extend into the mat. Splices in adjacent parallel rebar in the mat*are in general not less than 4 ft apart. Hoop tension in the cylinder wall is resisted by horizontal bars located near both the outer and inner surfaces of the wall. All horizontal circumferential bars including those in the dome have their joints staggered at a minimum of 3 ft apart.
- Longitudinal tension in the cylinder wall is resisted by two rows of vertical bars, one near the interior face and the other near the exterior face of the wall. Vertical bars are placed in groups of 20 bars of equal length. These are arranged so that no adjacent group in the same or opposite face of the wall has splices closer than 6 ft vertically.
The dome reinforcing consists of layers of rebar placed radially extending from the vertical reinforcing of the cylindrical wall and horizontal layers of circumferential hoop bars. Layers are located near both the inner and outer faces of the concrete. The radial pattern of the reinforcing steel terminating
15.5.1.6-2 12-1-69
- in the containment dome results in a high degree of redundancy of reinforcing steel in the dome. Bars are terminated beyond a point where there ia more than twice the amount of steel required for design purposes. The rate of convergence of these bars and low stress requirements dictated by the arrange-ment produces a low bond stress. In a limited number of cases where bars are terminated close to the center of the dome, anchorage stresses are more .,- ,
critical, and bars are hooked to provide the required anchorage. Near the crown, the rebars are welded to a concentric ring cast in th~ concrete. Radial shear loads to resist internal pressure resulting from the DBA are resisted by rebars inclined*at 45 deg with the horizontal and extending between the surfaces of both the vertical reinforcing closest to the interior and
- exterior faces of cylinder wall. This radial shear will vary from a maximum at the base of the wall where the foundation mat restrains the independent movement of the wall to zero at some level .above the mat. Anchorage bond stresses in these shear bars is kept below allowable stress levels to minimize potential cracking of the concrete. In addition., sufficient longitudinal *and circumferential reinforcing is carried to the base of the wall to carry all potential loads without assistance from the radial shear reinforcing.
The tangential shears resulting from the earthquake loading are resisted by
~~ t
~~
rebars inclined at approximately 45 deg in each direction,\in the plane of the
\
wall parallel to the main reinforcing steel. Minimum concrete cover for all principal reinforcing -steel of the containment
- structure exceeds the requirements of ACI 318, Paragrpah 808 (d), which states,
-15.5.1.6-3 12-1-69 "Concrete protection for reinforcement shall in all cases be at least equal to the diameter of the bars." The largest and "principal reinforcing bar is No. 18 which requires a minimum cover of only 2 3/8 in. by the code.
15.5.1. 7-1 12-1-.*69
- 15. 5 .1. 7 Penetration Design Penetrations through the containment structure are divided into one of the following three categories:
- 1. Pipe penetrations 9 in. diam or less No special structural reinforcing is provided for penetrations 9 in. or less in diameter. Penetrations in this category are located to avoid interference with the reinforcing steel.
- 2. Pipe penetrations greater than 9 in. and up to 3 ft-6 in. diam
- For penetrations greater than 9 in. and up to and including 3 ft-6 in. diam, supplementary reinforcement is provided in amount and distribution such that area requirements for reinforcem~nt are adequately satisfied.
At all these size penetrations, reinforcing steel interrupted by the openings is terminated at each side of the opening*. Supplementary reinforcing was placed parallel to the interrupted bars to provide bar continuity. Horizontal, diagonal, and yertical bars were used to effectively frame the opening. The total area of reinforcement provided in any plane is not less than twice the area of steel interrupted or cut by the opening with half of this placed on each side of the opening. Additional reinforcing, around these opening.a is not less than 20 ft in length and of sufficient length to develop the full ultimate strength of the bar in
15.5.1.7-2 12-1-69 ultimate bond stress to conform to the requirements of ACI 318, Section 1801 (c 2).
- Horizontal bars are considered as top bars for this purpose.
- 3. Openings larger than 3 ft-6 in. diam The two openings in this category are the 7. ft-0 in. diam personnel access hatch and the 14 ft-6 in. diam equipment access hatch.* Details of the additional reinforcement provided around the equipment access hatch and personnel access hatch are shown in Figs. 15.5.1.7-1 through 4, iu~lusive.
These penetrations are analyzed by means of a computer program. 4 This program analyzes a ring beam based on the method of virtual work. The program assumes the ring beam to be isolated from the containment shell and loaded in two planes. The analysis includes the effect of the stiffened ring and the moments intro-duced by transferring external loads from the shell at the perimeter of the ring to the center line of the beam. The ring beams are designed to resist biaxial bending moments, axial tension, torsion and biaxial shear resulting from loading criteria listed in Section 15.5.1.2. The biaxial bending moments and axial tension are assumed to be resisted by the reinforcing bars only, the concrete being neglected. The torsional and biaxial shear stresses are assumed to be resisted entirely by binders placed radially around the penetrations. Torsion is computed by the formulas f9r torsion in a rectangular beam. The principal circumferential arid meridional reinforcing is extended to the inner face of the ring beam and. bent at right angles, hereby providing additional shear resistance, the availability of which is considered in the design.
15.5.1.7-3 12-1-69
- The normal pattern of membrance stress in the cylinder wall is interrputed in the area adjacent to the stiffened openings. This redistribution of stress wa:s investigated by means of a computer .program, based upon a paper by B. Budiansky and P. Radkowski 5
- For this investigation, a flat circular plate with a radius equal to three times the distance from the center of the opening to the outside face of the stiffening ring beam was used to establish the stress pattern.
The movement of both the stiffened ring and the adjacent shell was compared to determine if .significant discontinuity stresses were present. Extra reinforce-ment was added -to regions of marked deviation from the normal pattern to keep the discontinuity effects to the level at which they can be considered negligible. The gross concrete area of the ring section was -used to determine the section stiffness and rigidity
- j
15.5.1. 8-1 12-1-69
- 15. 5 .1. 8 -Steel Liner and Penetrations The containment structure has an inside diameter of 126 ft-0 in. and an interior vertical height of 185 ft-1 in. me.asured from the top _of the foundation mat to the center of the dome. The cylindrical steel wall liner is 3/8 in. thick, the hemispherical dorii.e liner plate is 1/2 in. thick, and the flat base liner is 1/4 in. and 3/4 in. thick. The steel lining is a,ttached to and supported by the* concrete; the liner functions primarily as _ a gastight membrane. The steel wall and dome liner are protected .from potential interior missiles by interior concrete shield walls. The base liner is protected by a 1 1/2 to ,2 ft thick concrete mat except in two areas where 3/4 in. thick liner plate is used benea_th the reactor vessel incore instrumentation, and at a drainage trench where floor gra.ting provides additional protection.
The steel liner is designed to withstand the effects of all temperature, earthquake and pressure loads, including the effect of the subatmospheric operating pressure. The liner stress limits and their associated strains are limited to the stress criteria given in Paragraph N-1314 of Section III of the ASME Boiler and Pres-sure Vessel Code for nuclear vessels and to basic primary stress levels taken from Table_N-421 of that Code. The liner material SA-442-GR60 has a specified NDT that is at least 80° F below the minimum liner operating temperature and 0 considerably more than 120 F below the Design Basis Accident (DBA) temperature. Under either of these conditions, the liner material is able to accommodate at least 60 percent plastic strain without cracking. A strain of this magnitude
15.5.1.8-2 ., 12-1-69 is at least 80 times greater than the maximum strain that will be imposed on the liner
- 11
. Reference to a generalized fracture analysis diagram shows that the Crack
. . 0 Arrest Temperature" (CAT) curve crosses the NDT +80 F line at approximately 60 percent of the span between the "Fracture Temperature Elastic" (FTE) and the "Fracture Temperature Plastic" (FTP) ordinates, indicating that the steel can be strained to 60 percent of the required strain to fracture without cracking, even in the presence of large flaws.
To demonstrate that this plate material can accommodate plastic strains of this magnitude when biaxially stressed, tests were conducted on samples of 3/8 in. thick plate of identical specification to the steel to be used in 0 this containment liner at a temperature of 90 F above the NDT of the steel. In these tests, the plate samples were each laid across a 23 in. diam ring, and a 3 5/8 in. diam mandrel was forced into the plate at the center line of the ring. In all cases, the mandrel deformed the plate by an amount in excess of 4 in. before shearing through. The Design Basis Earthquake can be expected to produce tremors to the extent of not more *than 8 to 10 cycles and the Operational Basis Earthquake not mqre than 4 to 5 cycles. Operating pressure v*ariations from 9.5 psia to 14. 7 psia can be expected to occur not more than 100 times during the lifetime of the unit since personnel access is permitted under subatmospheric conditions
- 15.5.1.8-3 12-1-69 Temperature variations from 70° F to 105° F resulting from seasonal swings and shutdowns of the unit can be'expected to occur not more than 400 times during the lifetime of the unit.
The containment liner is designed for l,000 cycles of operating pressure varia-tions, 4,000 cycles of temperature variation, and 20 cycles of Design Basis Earthquake, all simultaneously applied. The containment liner is also designed*for one cycle of DBA pressure, one cycle of DBA temperature, and ten cycles of Design Basis Earthquake, all considered simultaneously applied. The containment liner is designed, within allowable working stresses, to with-stand a vacuum increase of not less than 1.5 psi. '!'he shell and dome plate liner is capable of withstanding an internal pressure of 3 psia and the bottom mat liner is capable of withstanding an internal pressure of 8 psia, with reference to standard atmospheric conditions outside the containment. The change in barometric pressure due to tornadoes is no.t expected to exceed 3 psig and the change due to maximum hurricane will be approximately 1.1 psig. These pressure changes will result in a decrease in the atmospheric pressure, which willdecrease the differential between atmospheric pressure and the containment structure ambient pressure, thereby decreasing the potential for stresses in the containment.
15.5.1.8-4 12-1-69
- The accumulated effects of the above are evaluated in accordance with Paragraph N~415.2(d) of the ASME Boiler and Pressure Vessel Code, Section III, Nuclear Vessels.
The steel containment liner is securely anchored to the concrete wall and dome with Nelson stud type concrete anchors. Failure could occur by stud fa~lure in shear or tension, by studs pulling out from the concrete or by studs tearing off from the liner plate. Tests conducted by Northeastern University, Boston, Massachusetts, using 1/2 in. diam studs and 3/8 in. thick plate show that shear failure occurs in the stud adjacent to the weld connecting the stud to the plate; in no instance was the plate damaged. Tests conducted for the stud manufacturer under the direction of Dr. I. M. Viest indicate that, with the manufacturer's ~ecommended depth of embedment of the stud in concrete, the ultimate strength of the stud material can be developed in direct tension. The principal design load imposed on the studs is due to the subatmospheric pressure operating condition, with the anchor lattice spacing based on con-siderations of plate buckling. A safety factor greater than 10 is provided against stud failure in ~ension. Shear due to DBA conditions and earthquake will result in stresses less than the allowable working stresses. In addition to the concrete stud anchors, the wall and base mat sections are anchored and joined at the intersection of the vertical wall and the.base mat
- with a continuous steel skirt embedded and anchored in the concrete
- 15.5.1.8-5 12-1-69 All anchors are designed so that failure occurs in the anchor, thereby assuring ~
that the leaktightness of the containment liner will be maintained during and after anchor failure. Probable mode of failure will be one of random stud failure due to poor work-
' manship during stud attachment. This type of failure will result in separation of the stud from the liner without impairment of the liner ductility of integrity.
Loss of random anchor points wi.11 not trigger a chain reaction since the design load on each stud is low compared with the stud load capability. Design
- spacing of these studs is such that not less than a group of 10 adjacent studs would have to fail to cause a liner plate to reach its yield stress under design operating conditions. Even with this unlikely condition, the. loads on ~
the studs adjacent to this area would remain within their safe load capability. As shown in Fig. 15.5.1.8-1, the liner.was welded to a skirt ring which .in turn is embedded and anchored into the concrete mat. The skirt-to-liner juncture and the skirt-to-mat anchorage was proportioned to develop the full strength of the liner. Under DBA conditions, the liner at the base juncture will be under a state of biaxial compressive strain due primarily to thermal effects. All thermally hot pipes penetrating the reinforced concrete containment wall pass through individual sleeves that are approximately 1 ft in diameter larger titan the pipe and project inward a distance of*approximately 2 ft f~o~ the liner. A typical application is shown in Fig. 15.5.1.8-5. The pipe is welded to a thick
15 .5 .1. 8-6 12-1-69 ~ cap that is an integral part of the end of the penetration sleeve. The sleeve is equipped with two coolers, each in the form of a waterwall. One cooler was fitted into the inside of the sleeve where it is in intimate contact with the sleeve in the containment wall, and the second cooler was fitted onto the out-side of the sleeve where it is in intimate contact with the portion of the sleeve that projects into the containment. The cooling water circulation pipes do not require any secondary penetration of this containment structure. All penetrations are anchored in the reinforced concrete containment wall. The anchor strength is equal to the full yield strength of the pipe with regard to torsion, bending, and shear, and to the maximum possible pipe jet reaction. All stresses induced in the liner by these combinations of lo~dings are only those reflected by the resulting distortions in the reinforced concrete con-tainment wall and are minor in intensity. , Therefore, loads will not be imposed I on the liner, thereby preserving its integrfty. The pipes anchored to the containment penetrations between containment isolation valves constitute an extension of the containment and are designed in accord-ance with the USA Standard Code for Pressure Piping - Power Piping, USAS B31.l.O - 1967 with respect to materials and allowable stress. Analyses of stresses due to thermal expansion and shock loadings from earthquake, pipe jet reaction, and other causes were made using established digital computer calculation techniques. In orde~ to determine the loading combinations that act on a penetration, the
' i pipe line passing through the penetration sleeve was assumed to have failed
- traversely at several locations along its run. The location at which the
- 15. 5 .1. 8-7 12-'l-69 reaction of the ensuing jet of fluid flowing from the broken end first causes the pipe to completely yield, in either bending or torsion, was taken as the design case from which all resultant combinations of penetration loading were determined for that particular pipe line. The maximum stress allowed on any individual element of the penetration is 90 percent of the minimum yield point.
All liner seams were strength welded. Small steel channels welded continuously along the edges of their flanges to the liner plate cover the plate weld seams, in a manner similar to those installed at the Connecticut Yankee Station. These channels are zoned into test areas by dams welded to the ends of the sections of the channels. Fittings are provided'in the channels for periodic testing of the weld seams for leaktightness under pressure. Typical liner details *are shown in Fig. 15.5.1.8-7. Testing of the liner is described in Section 5.5.
- 1 To transfer the stress adequately around penetration openings, the liner is reinforced in accordance with the rules set forth in the ASME Boiler and Pres-sure Vessel Code, 1968, Section I~I, Nuclear Vessels, or to adequately transfer the. pipe yield* load to the concrete within the limits of this material, which-ever is larger.
All major equipment and pipe loads are carried 9n the interior concrete structure or by the neutron shield tank. A 1 1/2 to 2 ft thick concrete slab placed over the bottom mat steel liner provides anchorage and support for other equi~ment located in the base of the .containment structure. The neutron shield tank skirt is attached to the containment mat by 1 1/2 in. diam anchor bolts. The skirt
- 15. 5. 1. 8-8 3-15-71
- support was welded to the liner and the entire weld, including the anchor bolts, covered by test channels. The internal concrete structure is attached to the containment mat by lengths of 3 in. by 6 in. steel bars which, placed hori-zontally, intersect the steel plate liner as shown in Fig. 15.5.1.8.;.2. The main vertical reinforcing steel bars were welded to the top and bottom faces of these bars, thus providing bar continuity without creating multiy:,le pene-trations through the liner.
The 1 1/2 ft thick concrete slab is anchored thorugh the steel liner plate in a similar manner using 7 in. by 1/2 in. bars, as shown in Fig. 15.5.1.8-3.
'these bars, termed bridging bars, form an integral part of the steel liner and conform to the material and workmanship specifications of the steel liner. All
- welded joints are covered by test channels and tested as all other liner plate joints.
*1 I
Access to the containment structure is provided by a 7 ft-0 in. ID personnel ! hatch and a 14 ft-fi in. ID equipment hatch. Other smaller containment structure penetrations include hot and cold pipes, main steam and feedwater pipes, fuel transfer tube, and electrical conductors. Electrical conductors penetrating the containment structure range in size from No. 16 AWG thermocouple leads to 1 in. diam solid copper rods for 4,160 v power circuits. Each penetration group passes through 8 in. diameter steel sleeves. The sleeves were welded into the containment liner with a test channel around the weld for periodic leak testing, as shown in Fig. 15. 5. 1. 8-4
- 15.5.1.8-9 3-15-71 The basic electrical penetration, as shown in Fig. 15.5.1.8-4, consists of an 8 in. steel tube with bolted-on flanges, through which pass the sealed conductors.
The hermetically sealed connectors, as shown in Fig. 15.5.1.8-4 were bench tested for leaktightness. Each penetration is periodically tested using a halogen leak detector or a nitrogen bubble test. Each flange is held tightly in place with eight bolts that draw each flange
- against a high temperature sealing ring ana a backing plate welded to the sleeve.
Each flange is tapped, as shown in Fig. 15.5.1.8-4, for leak testing. An electrical connector may be replaced, if necessary, without welding or cutting of the contain-ment liner or sleeve. All containment structure piping penetrations consist of a basic containment insert, plus additional items, as required for the individual services. Each insert is a 4 ft-9 in. long pipe sleeve provided with a plate flange near one end and a smaller plate flange near the middle with rebars welded 'to it for anchoring the insert in the containment wall. On hot pipe penetrations and on multiple pipe cold penetration, the insert serves as a sleeve, and a thick plate welded in the outside end of the insert acts as a sleeve plug. On single pipe cold penetrations, the insert serves as a section of the process piping. Details of the various penetrations are shown in Fig. 15.5.1.8-5. The main steam and feedwater penetrations are provided with adequate space between the piping and the sleeve for the necessary pipe insulation and for a pipe coil outside the insulation through which component cooling water is circulated. This cooling coil reduces the temperature of the sleeve and prevents
15.5.1.8-10 12-1-69
- excessive heating of the concrete in contact with the sleeve. All welded seams subjected to containment pressure are leaktested by introducing Freon through each test boss. In addition, the sleeve end,is drilled and tapped, as shown in the details, so that any leakage between pipe wall and sleeve end can be detected during periodic containment leakage testing.
The liquid and gas pipe penetration assemblie~, in nearly all instances, ~on-sist of more than one pipe inside the pentration sleeve. The diameter of the sleeve depends on the number and size of the pipes installed in a given penetra-tion. Each of these penetrations was tested with Freon using the same procedure as that used for the steam and feedwater penetrations.
- A 20 in*. OD fuel transfer tube penetration is provided for fuel transfer between the refueling canal in the containment structure and the spent fuel pit in the fuel building. The penetration consists of a 20 in. stainless steel pipe installed inside a 26 in. pipe, as shown in detail in Fig. 15.5~1.8-5. The inner pipe acts as the transfer tube and connects the containment refueling canal with the spent fuel pit. The outer pipe is welded to the containment liner and provision is made, by use of a special seal ring, for Freon ga~ leak testing of all welds essential to the integrity of the penetration. Bellows expansion joints are provided on the outer pipe to compensate for any differential movement between the two pipes.
The equipment hatch is a 14 ft-6 in. single closure penetration. The equipment hatch cover is mounted inside the containment structure and is double gasketed
- with a leakage test tap between the "O" rings. The equipment hatch cover is
- 15. 5. 1. 8-11 12-1-69 provided with a hoist with two point suspension and a sliding rail for storage
- A positive locking device is furnished to prevent circular swing. The equip-ment hatch was designed, fabricated and stamped in accordance with the ASME I
Boiler and Pressure Vessel Code, Section III~ Class B. A removable concrete tornado missile shield protects the equipment hatch and acts as equivalent shielding. The personnel hatch is a 7 ft-0 in. ID double closure penetration as shown in Fig. 15.5.1.8-6. Each closure head is hinged, double gasketed with a leakage 11 test tap between the 0 11 rings. Both doors are interlocked so that in the event one door is open, the other cannot be actuated. Both doors are furnished with a pressure equalizing connection. The equalizing valves are manually operated by persons entering or leaving the personnel hatch. The personnel hatch was designed, fabricated and stamped in accordance with the ASME Boiler and Pres-sure Vessel Code, Section III, Class B. The personnel hatch is externally protected from tornado missiles by concrete shield walls and roof. Material for the liner and penetrations are carbon steel plates conforming to ASTM A442, Grade 60, which has a specified minimum tensile strength of 60,000 psi, a minimum guaranteed yeild strength of 32,000 psi and a guaranteed minimum elongation of 25 percent in a standard 2 in. specimen. The liner has sufficient ductility to tolerate local deformation without rupture. This material has a nil ductility transition temperature ,pf -20° F which is 80° F below the normal minimum shutdown temperature given in Section 5.4.1. Steel items, except backing plates and anchors, gas testing channels, equipment hatch bolts and equipment hatch nuts are made to fine grain practice and
15.5.1.8-12. 12-1-69
- normalized. In addition, steel items other than the above have passed NDT
*tests performed in accordance with the following:
I. Material 5/8 in. and thicker was tested by the Drop Weight Test method in accordance with ASTM E 208. '*}*
- 2. Material less than 5/8 in. thick was tes'tecl by the Drop Weight Tear Test method as develope.d by the U.S. Naval Research Laboratory (NRL Report 6300).
0
- 3. Material 5/8 in. and thicker has an NDT no higher than -20 F.
The liner plates were ordered to conform with standard mill practice with regard to thickness tolerances *. Therefore, the 3/8 in. thick-cylindrical shell liner plate ranges in thickness from 0.365 in. to 0.406 in. The 1/2 in. thick hemispherical dome liner plate ranges in thickness from 0.490 in. to 0.535 in., and the 1/4 in. thick flat base liner plate ranges in thickness from 0.240 in. to 0.285 in. Physical and chemical properties of materials used in the construction of the containment. liner, weldability tests, and liner thickness were checked by the Stone & Webster Field Quality Control Organization on a random sampling basis. All welding procedures and tests required in Section IX of the ASME Boiler and Pressure Vessel Code for Welding ~ualifications were adhered to in the selection of weld rod material, weld rod flux, heat treatment, and qualifying the welding
15.5.1.8-13 1~-1-69 procedures and the performance of welding machines and welding operators engaged in the construction of the containment liner. 180 deg bend tests of weld material. The welding qualification included These procedures ensured that the ductility of welded seams was comparable to the ductility of the containment liner plate material. Section III of the ASME Boiler and Pressure Vessel Code for Nu.clear Vessels was used as a guide in the selection of materials. Erection of the steel liner followed completion of the concrete mat. The 3/8 in. thick steel wall liner was erected to approximately El. +60 ft. The 1/4 in. thick mat liner plate was installed on top of the concrete foundation mat during this period. On completion of the wall liner to El. +60 ft and completion of the mat liner, all welds were checked for compliance with the approved weld inspection and gas test requirements. Work on the liner was then stopped until the containment inerior concrete structure was completed, the *polar crane was erected, and the concrete containment wall was completed to gr'?und grade (El. 26. 5) ~ The 3/8 in. thick steel wall liner was erected from El. +60 ft to El. +92 ft-6 in. and the containment liner completed'with the construction of the 1/2 in. thick steel dome liner. A-1 welds were inspected and gas tested for compliance with the weld requirements. The reinforced concrete wall, above ground grade, was completed following as closely as pr~ctical the construction of the wall liner.
15.5.1.8-14 9-15-71
- The reinforced concrete dome was constructed upon completion of the dome liner, The steel wall liner was braced internally and locally with temporary bracing to prevent distortion during concrete placement. The exterior concrete forms were supported from* the placed concrete and tied to form a tension ring.
Cantilevered steel strongbacks were used in the construction of the concrete dome to support the steel dome liner against deformation due to the weight of reinforcing steel formwork and wet concrete. Strongbacks were cantilevered from the completed concrete of the dome. The containment liner is not a coded pressure vessel, therefore; there was no section of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels directly applicable to its design and construction. However, to insure that
- good engineering. practices were followed, certain portions of Section III of the Code were reviewed for suggested guidance as to design and construction practices that should be incorporated in the liner specifications. Those sections reviewed for information were:
N-511 Certification of Materials by Vessel Manufacturer N-512 Material Identification N-513 Examination During Fabrication N-514 Repair of Material by Welding N-515 Forming Shell Sections and Heads N-518 Attachments N-519 Cutting Plates and Other Products N-521 Welding Processes
- N-522 N-523 N-524 Welding *Qualifications and Weld Records Precautions for Welding Assembly
- 15. 5 .1. 8-15 9-15-71 Finished Longitudinal and Circumferential Joints N-527 Miscellaneous Welding Requirements N-528 Repair of Weld Defects N-531 Preheating N-541 Modification of Section IX - Welding Procedure Qualification Requirements N-611 Inspectio~, General N-612 Qualification of Inspectors, Engineering Specialists, and Inspection Agencies N-613 Access for Inspector N-614 Inspection of Materials N-615 Marking on Plates and Other Material N-616 Final Inspection N-620 Inspection of Welding N-622 Check of Welder and Welding Operator Performance Qualifications N-623 Check of Nondestructive Examination Methods N-625 Ultrasonic Examination of Welded Joints N-626 Magnetic Particle Examination N-627 Liquid Penetrant Examination N-713 Pneumatic Test N-714 Pressure Test Gages The liner attachments are Nelson concrete anchors, welded on a triangular pattern to the wall and dome liner and cast in the containment concrete as the concrete was poured against the liner. The attachment spacing was determined 9
by the procedure set forth for buckling of a cylindrical shell under combined axial and uniform lateral pressure where each attachment constitutes a buckling
- 15. 5. 1. 8-16 12-1-69 wave nodal point and was so spaced that the critical buckling stress will take place in plastic range of the liner material. The liner dome was treated in a similar manner. Maximum variation from the correct stud location, where relocation was necessary to avoid an obstruction, did no~ exceed 1 1/2 in.
The bottom mat liner was covered with 1 1/2 to 2 ft thick reinforced concr~te slab to protect it from both pressure and temperature loadings so that it will remain virtually unstressed. All penetrations are anchored into the concrete containment structure wall with a loading resistance level greater than the plastic strength of the penetration pipe. Openings in the liner plate are reinforced with reinforcing plate, and/or collar, sized to develop the full relief of the liner plate. The stress around each reinforced opening was *analyzed in accordance with the appropriate pro-
. 10 cedure *
- Departure from the original specified out-of-roundness tolerance of the reactor containment liners was necessary due to erection difficulties. Attempts were made to obtain the specified tolerance by means of an adjustable ririg girder and supplementary anchorage to the cofferdam. As work progressed above the cofferdam level, it was found that it was impractical to obtain the specified liner tolerance.
A thorough review was made of the necessity for this close tolerance, and it was found unnecessarily restrictive. The liner shell and dome are studded to the concrete and the plate is essentially
- - plane within an equilateral triangle, 12 in. at the base and bounded by studs at the apexes of the triangle *. The response of each individual trianglar element
15.5.1.8-17 12-1-69 to its own particular loading system establishes the adequacy of the structure as a whole.* Therefore, actual roundness of the shell has no effect on liner performance. The following revised out-of-roundness tolerances were adopted after a thorough review of the problem.
- a. The out-of-roundness tolerance shall not exceed plus or minus 3 in. from the true radius.
- b. The maximum plus or minus deviation from a true circular form shall not deviate more than 1/4 in. from a straight line in any 14 in. space in any plane in any location on the liner.
The revised out-of-roundness tolerances have no adverse effect on the buckling strength of the liner and ensure that plate buckling between studs will not occur in the elastic range. The adjustable ring girder was found to be of limited value during the erec-tion of the liner due to the many liner penetrations and the stiffness of the liner shell. Therefore, the ring girder was used for rounding the shell only in areas where its application was found advantageous by the liner fabricator.
FIG. 15.5.1.1-1 DEC. I, 1969 - 11 8 PIPE TO TOP OF COFFERDAM CONTAINMENT I* WALL *I PUMP WELL WELL COMPACTED _.------tt-- 4" CONCRETE BLOCK 11 6 REQUIRED l CRUSHER RUN STONE
-. . **.. :1,. * * *b ~
0 "~- Ooo 0 CONCRETE t>.**"-'.'4* Oo,o0o'o Oo o LINER
. . *. 00 o0 WALER /,.'" .. "'*. D().i
- /_* ** t> ** 4
. *A . ..
CONCRETE SLAB
--~*:-:~*.
15" HALF ROUl':JD POROUS EL-21'-7" SET CLEAR OF CONCRETE WALER ~-Got 12 " :t, ALT. AROUND PERIPHERY 1 11 6=1_ EL:-29 -7 _.,__~ 1 PUMP ON WHEN 6dtl2' :t WATER REACHES______,/f IUL--1-=:=p=- 3'-o" LG., ALT. THIS LEVEL WATERPROOFING AROUND PERIPHERY MEMBRANE
+/- ~
11 11 at 12 I II ~ 4 - 6 LG., ALT;
...J AROUND PERIPHERY <(
= a::
0 :::> CONCRETE _ I 1-WALER o U a:: POROUS I-11 (/) CO NC RE T E ---J.U---'-"...... 4 CONCRETE BLOCK II 4 POROUS CONCRETE o"'~* c=i. f WATERPROOFING BOTTOM OF PUMP WELL 1 11 MEMBRANE EL-42 -5 ~ 11 12 CRUSH ED STONE 11 2 LEVELING SLAB SHEET PILING 1 11 SCALE: 1/4"= 1 -0 REACTOR CONTAINMENT WATERPROOFING SURRY POWER STATION L
FLG. 15.5.1.5-1 DEC. I, 1969 I:
F l:G. 15. 5. I, 5 - 2 DEC. I, 196 9 f-
FIG. 15.5.1.5-3 DEC. I, 1969 1' - -+
FIG .. 15.5.1.7 - I DEC. I, 1969
- A 1 EQUIPMENT HATCH SYMETRICAL ABOUT CENTER LINE TYPICAL WALL REBAR
#18 S TYPICAL DIAGONAL No. 9 STIRRUPS I
l 1 7 -3 11
.I. a'-o" REINFORCING DETAILS EQUIPMENT ACCESS HATCH OPENING
- SURRY POWER STAT ION
FIG. 15.5.1.7-2 DEC. I, 1969 63'-o" R. 4'-6" I t 1 #9 RADIAL STIRRUPS 10" 10" #9(EVERY
@ e 0 OTHER SET OF RADIAL STIRRUPS) 0 0 HORIZONTAL REBARS 0 0 lJJ.Jl-1'1----==:c-.A DWE L DE D
#9 ( At EVERY OTHER SET OF RADIAL STIRRUPS I IN GROUPS OF-3 0 0 TYPIC*AL 0
_I 0 0 0 © HORIZONTAL a, REBARS
- - - #9,s" or a" o.c e RADIAL STIRRUPS IN GROUPS OF 2 EQUIPMENT HATCH VERTICAL REBARS CADWELDED -CJ)
_I SECTION A-A B-B
~
LEGEND e TYPICAL WALL REINF. I:> EXTRA WALL REINF. 0 CIRCULAR REINF. REINFORCING DETAILS By SECTIONS THROUGH RING BEAM EQUIPMENT ACCESS HATCH SURRY POWER STATION
l FIG. 15.5.1.7-3 DEC. I, 1969
.......,f---1...........---+--+----14-+-........+--~t----';,'--t-r- # 18 5 2 EACH ROW
- 4- #185 EACH ROW To STIRRUPS-No. 9, SPACING AS SHOWN ON SECTION C-C (FIG. rs.5.1.7-4)
REINFORCING DETAILS PERSONNEL HATCH OPENING SURRY POWER STAT! 0 N
FIG. 15.5.1,7-4 DEC, I, 1969
#1as at 9"= 6 1-0 11 9" 63'...0"R 4'-s #" 6 I N GR O UPS OF 3 (At EVERY OTHER SET HORIZONTAL REBARS OF RADIAL STIRRUPS) CADWELDED 0
co ti 0, 3' - 6" TO PERSONNEL HATCH LINER 0 3/8" LINER IO II)
# 9 at 9=-"~--
RADIAL STIRRUPS IN GROUPS OF 2 0-D ci VE.BT REBARS CADWELDED LEGEND SECTION C-C e TYPICAL WALL REINF. 1.9 EXTRA WALL REINF. 0 CIRCULAR RE INF Df ,REINFORCING DETAILS SECTIONS THROUGH RING BEAM PERSONNEL HATCH SURRY POWER STATION
FIG. 15.5,1,8-1 DEC, I, 1969
- 0
- 0 CTOP OF CONCRETE FILL
*_.*.::*t<: :*_~_- :. _/_---.-~-- *._ *_ -~-: .: r--..*_ -:_ ."-*-
0,. 0 LINER REBAR ANCHOR
*N
~
co
- L
=*
2 DIA. HOLES ON 5'- O"CTRS. 1 1/ STEEL MAT LINER TOP OF CONCRETE MAT CONCRETE WALL AND MAT JOINT
- SURRY POWER, STAT I ON
FIG. 15.5.1.8-2 DEC. I, i969 . BRIDGING BAR LINER PLATE TOP OF CONCRETE MAT
*co 1112" 1 ~4" BACKING STRIP TYPICAL SECTION THROUGH BRIDGING BAR USED TO PROVIDE MAIN REINFORCING STEEL CONTINUITY THROUGH MAT LINER SECTION - TYPICAL BRIDGING BAR SURRY POWER STATION
FIG. 15.5.1.8-3 DEC. I, 1969
- * ~ . .
0
*. 0
,,,---TOP OF CONCRETE 0
D
. ~
* . 0 0
SL AB
. 0 0
D
* ,0. . *
#'6@120.c.
STAGGERED x 1/2" PLATE CONTINUOUS 11 7
- 1/4 .. LINER PLATE TOP OF MAT
..r')_ _ _ _ _ :f_ - -,
I I I I I 1 I /2 11 11 I
/4 0
PLATE (SHOP WELD) I 11 l'--#6 @ 12 0. C. I I STAGGERED
~
TYPICAL SECTION BRIDGING BAR USED TO ANCHOR CONCRETE SLAB TO CONTAINMENT MAT THROUGH MAT LINER SECTION -TYPICAL BRIDGING BAR SURRY POWER STATION
FIG. 15.5.1.8-4 3-15-71 BLIND NUTS WITH STD. 150 CARBON STEEL FLANGE HELICOIL INSERT 7" ASTM-A350 GR LF2 CONTAINER LIN ER ---.,-...,-,.-, MATING PLU.G 0 0 0
'"-.._/' -----
CABLE SLACK ALLOWS FLANGE WITHDRAWAL AND MAKEUP 0
*o SEE DETAIL SILVER BRAZED IN THE HELO SILVER BRAZED SEE DETAIL "L" FLANGE "B" IN CABLE VAULT DOUBLE ENDED RECEPTACLE FLANGE "A" INSIDE. C0911TAINMENT
\.,* *125 FINISH ON BEARING SURFACE "o" RING SEE DETAIL "L" 8- !/4" BOLTS ON 115/4
- DIA. BOLT CIRCLE STRADDLING I. - - - - -
- - - - ~ ~ ...........~....-.... :/.*~=**
, , _ . , ~ * ., ... ; j,:.*-*
\ / ) ( ) :~: . 4 1-6" CONC. BIOLOGICAL SHIELD COUNTER BORE USING PARKER Y-34732 PLUG WITH PARKER 4HPS0N-3
_ _j 7/11"- 20 UNF-2B g,j:j~::
~ l_TAP DRILL THRU
, DET. *L* TEST PLUG TYPICAL ELECTRICAL PENETRATION SLEEVE WITH FLANGES SURRY POWER STAT! ON
FIG. 15.5.1.8-5 OCT. 15, 1969 SH. I OF2 REACTOR CONTAINMENT WALL __ ~ ::-: ....... . f LINER
.:i:".*,:'*:~*-:_:_ (*: *:; .*.* *6 .. **....... \
TEST CAP-REMOVE
*.*.~ *:.*.*,:.*. *."o, D . , AFTER TESTING FUEL BUILDING
. " .. h ....
REACTOR CONTAINMENT
~--.
TRANSFER TUBE WITH
._-
- 3 EQUI
- SPACED LUGS
* . ONE LUG ON HORIZONTAL t. END Pl.ATES AND Bl.IND
[ .*:.::.*_;..-:- ..:~:~:~'..-~.'-:*:;:~ ~:*.~ :~: \ s3'-o"R. ... FLANGE
. . ~ . : ... *..
. .* ... : *.'. ~ *. :._ ~ _*-:
.** .* :*.~. . :*~*=* ~ :_:*~: :_:
VALVE TYPICAL COLD PIPE PENETRATIONS REACTOR AND FUEL TRANSFER CANAL
.i::-.. -~ *. *. ~.: :_ ...... ;._~. . . .~---. ..
**. .-:*.: . *:. ~:-~* ..._~ :-:; ._*:*:.:-
** EXPANSION EXP*NSION JOINT JOINT JOINT
~ : - : * , * , . - : *** ~ * , : , : . ~ * * . : ,
.Y .. , ' . '
/ I TRANSFER TUBE
- : : - *4 EOUI-SPACED WGS -_. ~1--f" LINER 2"-t-t--.' STRADDLING t. 2" ......
*) :**::_<,:.*>::-<**.* .":J, :<. t, *: ~ **
INSI.LATION ** ... . t.' ** -~ * : . " : *** TEST CAP-REMOVE AFTER TESTING REACTOR CONTAINMENT GROOVE PLAN- FUEL TRANSFER PENETRATION s3' - o" B- MATERIAL NOTES CARBON STEEL l. TO I f INCL. - ASTM
- A442- GR60 CARBON STEEL I. OVER 1f - ASTM- A201 *GRB*A300 CARBON STEEL FORGINGS - AS TM* A350
- GRLF2 TYPICAL HOT PIPE PENETRATION CARBON STEEL PIPE - ASTM - A333 - GR3 CARBON STEEL PIPE SLEEVES - ASTM - A333 - GR3 STAINLESS STEEL FORGINGS - ASTM- A 182 - F304
- FOR DETAILS OF HEAT EXCHANGERS SEE FIG.15.5.1.8-5 SH.2 STAINLESS STEEL PIPE - ASTM- A312
- TYP304 STAINLESS STEEL TUBING - ASTM- A269- TYP304 BOLTS - ASTM* Al93*87 NUTS - ASTM
- A 194* 2H WELDING ELECTRODES CARBON STEEL TO CARBON STEEL - AS TM* E 70 I 8 TYPICAL PIPING PENETRATIONS STAINLESS STEEL TO STAINLESS STEEi. - ASTM
- E 308 CARBON STEEL TO STAINLESS STEEL - ASTM
- E 309 SURRY POWER STATION
FIG. 15.5.1.8-5 OCT. 15, 1970 SH. 2 OF 2
- REACTOR CONTAINMENT WALL
; . : . . b,, *.. !7. ._.
LINER
. i,. .: I, . :.
<J . . ~ :: . :
HEAT EXCHANGER OUTER UNIT HEAT EXCHANGER INNER UNIT---...... ATTACHMENT PLATE HOT Pl PE PENETRATION CIRCULATING WATER CIRCULATING WATER OUTER UNIT HEAT 'EXCHANGERS TYPICAL PIP.ING PENETRATIONS SURRY POWER STATION
FIG. 15.5.1.8-6 DEC. I, 1969 CONTAINMENT SIDE
.,,. *~*
* .. (;;]. ' .
D *
/). . '!>-
- A,-. p..
6" SIGHT LIGHTING GLASS FIXTURE CLOSE~ ELEQTRICAL PENETRATION CONNECTION
-~:~~~Gl __ i TEST CONN. 18" EMERGENCY PLUGGED OPENING OPERABLE FROM
~ 36" WALKWAY BOTH SIDES 3"CAPPED EMERGENCY AIR CONNECTION e TEST CONNECTION PLUGGED '.JI.. "* p. . 1)' .*
END VIEW SIDE ELEVATION I. THE HATCH AND LOCKING MECHANISM ARE DESIGNED TO PROVIDE A CONTINUOUS SEAL AGAINST ANY PRESSURE WHICH MIGHT BE DEVELOPED DUE TO THE DESIGN BASIS ACCIDENT.
- 2. OPERATING PROCEDURE TO ENTER THE CONTAINMENT STRUCTURE FROM OUTSIDE:
A. TURN OUTSIDE LEVER OF OUTER DOOR TO STOP BY SAFETY PIN. WHEN PRESSURE IS EQUALIZED SAFETY PIN WITHDRAWS. CONTINUE TO TURN TO UNSEAL. B. SWING DOOR OPEN, ENTER HATCH AND SWING DOOR SHUT. C. TURN INSIDE LEVER OF OUTER DOOR TO SEAL. D. TURN INSIDE LEVER OF INNER DOOR TO STOP BY SAFETY PIN. WHEN PRESSURE IS EQUALIZED S,\FETY PIN WITHDRAWS. CONTINUE TO TURN TO UNSEAL. E. SWING DOOR OPEN, ENTER CONTAINMENT AND SWING DOOR SHUT. F. TURN OUTSIDE LEVER OF INNER DOOR TO SEAL.
- 3. FOR THE CONDITIONS OF LEAVING THE CONTAINMENT TO THE OUTSIDE, THE OPERATIONS ARE REVERSED.
PERSONNEL HATCH ASSEMBLY SURRY POWER STATION
FIG. 15.5.1.8-7 DEC. I , 1969
---126'-o" O.D. ..---126'-0" 0. D.
1* ** Z\ BAR ANCHORS r 4 1t ly BACKING PLATE i TYP:CAL TEST CONNECTION f- 6,000LB. SCREWED HALF COUPLING WITH HEXAGON HEAD PLUG TYPICAL WALL JOINT TYPICAL WALL JOINT WITHOUT BACKING PLATE WITH BACKING PLATE 1rx ,*11 r TEST CHANNEL-~
,,.__ _ _ _ DRIL!-,. AND TAP THROUGH SEAM FOR t NPT SOCKET HEAD WALL TO DOME JOINT PIPE PLUG TYPICAL DOME JOINT TYPICAL LINER DETAILS SURRY POWER STATION
.15.5.1.9-1 12-1-69
- 15. 5 .1. 9 Materials Concrete The description of concrete materials is given in Section 15.3.1.
Porous Concrete **- Porous concrete is used under the base mat to provide drainage for the contain-ment structure. The type of concrete is formed by the omission of the fine aggregate from a standard structural concrete mix. The mix was designed to have a 28 day compressive strength greater than 1,000 psi *
- Water porosity tests were performed earlier in ;an independent laboratory for porous concrete, using 6 in. by 12 in. cylinders prepared in the laboratory by compacting the material in three layers with standard tamping rods. A varying number of strok~s, ranging from 10 to 4b for each layer were used for different cylinders. After the concrete test cylinders had been properly cured, the amount of water which would flow through the 12 in. length of specimen during a 3 min period with a constant head of 4 in. of water above the top of each cylinder was determined. Results indicated water porosities of from 28 to
!/
47 gpm per sq ft, depending upon the amount of compaction and resulting density
/
of the cylinders. The porosity determined by the laboratory tests indicated that the 4 in. porous concrete layer under the base mat provides adequate drainage*, as the leakage
15.5.1.9-2 12-1-69 through the membrane waterproofing of the container would be of a minor amount. This layer serves as the collection means for the seepage removal system in the mat, described in Section 15.5.1.3. Reinforcing Steel Special large size reinforcing bars, No. 14 and No. 18, used in the construction of the reactor containment structure are steel of 50,000 psi minimum yield point, . conforming to Grade 40 of the "Standard Specification for Deformed Billet-Steel Bars for Concrete Reinforcement" ASTM A615 as modified to meet the following chemical and physical requirements: Carbon 0.35 percent maximum Manganese Silicon Phosphorus 1.25 percent maximum 0.5 to 0.25 perce,nt
.05 percent maximum Sulphur .05 percent maxim\llII Minimum yield strength 50,000 psi Elongation 16 percent minimum in a 2 in. test samples Tensile strength 70,000-90,000 psi For these special chemistry bars, all ingots were identified and all billets*
were stamped with identifying heat numbers. All bundles of bars were tagged* with the heat number as they come off the rolling mill. A special stamp marking was rolled into all bars conforming to this special chemistry to identify them as possessing the chemical and mechanical qualities specified.
15.5.1.9-3 12-1-69
- The Engineers'_ Quality Assurance Inspectors witnessed, on a random basis, the pouring of the heats and the physical and chemical*tests performed by the fabricator. Bars containing inclusions or failing to conform to the required chemist~ and physical requirements were rejected.
One 12 in. long test sample was furnished to the Engineers from a finished bar from each heat of the special chemistry rebars* to permit independent verification of physical and chemical analysis tests by the Engineers. Test specimens for the special chemistry rebars conformed to Section 10.1.1 of ASTM A615 and were Standar~ 0.505 in. diam specimens with 2 in. gage length. Rate of loadings was such that the tension tested sample was brought to the yield point in not less than two minutes or more
- For containment structures, reinforcing steel* consisting of No. 11 bars and smaller is steel of 40,000 psi minimum yield point, conforming to Grade 40 of the "Standard Specification for Deformed Billet-Steel for Concrete Reinforcement" ASTM A615.
Tpe reinforcing steel for structures other than the containment structures is described in Section 15.4.3. Cadweld Splices Cadweld reinforcing steel splices, Type T" full tension splices, as manufactured
- by Erica Products, Inc., Cleveland, Ohio, were used to splice 50,000 psi minimum
15.5.1.9-4 12-1-69 yield point reinforcing bar sizes No. 14 and No. 18. These splices, including the sleeves, develop tensile strengths not less than 90 percent of the minimum ultimate strength of the reinforcing bar. The average value of two or more successive splices develop not less than the minimum ultimate strength of the rebar. Information for splices other than No. 14 and No. 18 reinforcing bars is given in Section 15.4.3. Waterproofing Membrane The waterproofing membrane is a fle.xible polyvinyl chloride sheet having a minimum thickness of 40 mils. Associated adhesives and tapes consist of the membrane manufacturer's recommended material for the application conditions.
- 15. 5 . 1. 10-1 2-13-70 15.5.1.10 Construction Prpcedures and Practices After performing the general excavation described.in Section 15.4.4, two 149 ft-5 1/4 in. diam cofferdams were constructed, one for each reactor. The cofferdams consist of interlocking steel sheet piles supported by a system of heavily reinforced concrete internal ring wales. The top of the sheet piles is at El. + 11 ft, tip grade is at El. -48 ft. The interior of the cofferdams was excavated to approximately El. -41 ft. Seepage drains were then driven through a 12 in. alyer of crushed stone placed in the bottom of the excavation, as described in Section 15.5.1.3.
A 2 in. thick concrete leveling slab was placed over the crushed stone and 40 mil thick vinyl waterproof membrane placed over this concrete. A 4 in. layer of porous concrete was then placed over the membrane to protect the membrane and to serve as an internal** drainage system as described in Section 15.5.1.12. Porous concrete was also placed around the sides of the cofferdam to fill the space between the cofferdam and the edge of the concrete mat and to provide a form for the mat concrete. The waterproof membrane was extended vertically in this area: and protected by concrete block. The reinforcing steel,* steel bridging bars as cleS"cribed in Section 15.5.18 and other miscellaneous* steel inserts* required in the containment mat were placed and the concrete poured. The mat was constructed in six sections.
15.5.1.10-2 2..:13-70. The 3/8 in. thick steel wall liner was then erected to El. +60 ft on the con-tainment mat. The steel mat liner plates were installed on top of the concrete mat. All welds were checked for compliance with the approved weld. inspection and gas test requirements. The containment interior concrete structure was then built on the mat liner. On completion of the interior concrete structure the polar crane was erected. The exterior containment concrete wall was constructed to approximately El. 24 ft-6 in. during the construction of the interior concrete. On completion
.of the concrete substructure a vinyl waterproof membrane was attached to the exterior concrete surface with adhesives. The membrane completely encloses the containment structure below grade.
The space between the cofferdam and the containment structure was then back-filled with crushed stone compacted in 6 in. layers. A 2 ft thick layer of compacted impervious fill was placed at El. -4.0 ft to seal the area and' to I minimize the amount of ground water seeping into the area. The liner was then completed, finished with the construction of the 1/2 in. thick steel dome with all welds inspected and gas tested. The steel dome liner was supported during erection with open web steel trusses. The reinforced concrete wall above ground grade was completed following as closely as practical the construction of the wall liner.
15.5.1.10-3 12-1-69
- The completed .steel wall liner was braced internally and locally with temporary bracing to prevent distortion during concrete placement.
- The exterior concrete forms were supported from the preceding concrete.
Cantilevered steel strongbacks were used in the construction of the concrete dome to support the steel dome liner, reinforcing ste~l, formwork, and wet concrete against deformation: Strongbacks were cantilevered from the completed concrt~te of the wall or the dome. Careful inspection of the dome was maintained during concrete placing and until the concrete had definitely taken initial set. Concrete buckets used during the first two lifts of the dome were limited to 2 cu yd in size. Bucket sizes were increased after the second lift had set, where placi~g results of these lifts were satisfactory and warranted such a move.* Concrete in the wall and dome of the containment structure was poured in uniform 6 ft lifts around the entire circumference. Each lift w~s constructed in approximately 18 in. layers~
- Concrete forms were used on the exterior of the concrete dome to a line 50 deg above the horizontal. The permanent steel liner served as, the inner form for pouring concrete. For the area where exterior forms were used, the concrete points were in horizontal planes. Above the 50 deg line, the remainder of the dome concrete is poured as one lift.
Particular care was taken to check the special markings of the No. 14 and No. 18, 50,000 psi minimum yield rebars for the containment structure.
15 .5.1.10-4 12-1-69 Welded splices conform to "Recommended Practices for Welding Reinforcing Steel, Metal Inserts, and Connections, in Reinforced Concrete Construction," AWS Dl2.l. Bars spliced by metallic arc welding develop not less than 90 percent of the minimum ultimate strength of tµe reinforcing bar, andI the average of two or more successive splices develop not less than the minimum ultimate strength of the bar. i Structural ductility was maintained by staggering critical splices where possible. Full scale pressure tests conducted in May, 1967, on a recently 6 completed concrete containment structure in which similar Cadweld splices ,,,,** and welded splices were used showed no stress concentrations or lack of structural ductility. Locations of splice groups were not discernible from inspection of the test crack patterns. All Cadweld Process Type "T" joints were visually tnspected. The visual inspec-tion included inspection of the ends of the bars for dryness and cleanliness prior to fitting the sleeve over the ends. It also included inspection of the completed splice for properly filled joints to ensure that filler metal was visible at both ends of the sleeve and at the top hole in the center of the sleeve. Randomly selected splices were removed from the structure and strength tested for compliance with the specification. Joints that did not meet all these inspection criteria were replaced. Randomly selected Cadweld Type "T" splices were removed from the containment structure and tensile tested for compliance with the specifications in accord~ ance with the following schedule:
15.5.1.10-5 12-1-69
- 1 out of first 10 splices 3 out of next 100 splices 1 out of each subsequent unit of 100 splices Welding inspection of reinforcing bars was by Quality Control Inspectors.
Radiographic inspection, dye-penetrant inspection, magnetic-particle inspection, or other nondestructive inspection methods for welded joints was performed on a random basis under the direction of the Senior Quality Control Engineer. All welds were visually inspected. Any cracks, porosity, or other defects were removed by chipping or grinding u*til sound metal was reached, and then repaired by welding. Peening was not permitted *
- i Completed welded splices were selected on a random basis and removed from the structure with suitable lengths of adjacent bars.
I I These removed splices were tensile tested for compliance with the specifications in accordance with the following schedule: lout or first 10 splices I 3 out of next 100 splices 1 out o~ each subsequent unit of 100 splices Tack we lding of special chemistry rebar was not permitted
- 1
15.5.1.11-1 12-1-69,
- 15. 5 .1.11 Missiles and Piping Rupture Interior Missiles Most of the high pressure piping and equip,ent of, the primary coolant system are located within containment cubicles protected by reinforced concrete walls and floors with a minimum thickness of 2 ft. The'control rod drive mechanisms are provided with a separate missile shield. These reinforced concrete struc-tures will terminate the flight of any conbeivable missile. Openings in the charging floor required for ventilation or access are covered by steel grating, which are designed to provide adequate missile protection. Openings in cubicle walls for overpressure blowoff protection are directed in a manner which will minimize the possibility of missiles striking the containment liner. An analysis of the missile hazard has been performed and the conclusions are as follows:
Missiles could be either concrete or steel. Because of lower density and lower strength, a concrete missile would have to be an order of magnitude heavier than a steel missile of comparable diameter and velocity for it to cause the same damage on impact with a steel shell. Also, in. the context of the Design Basis Accident, there are more potential steel missiles and these have been studied in detail. The most penetrating steel missile for a given mass and velo.city would be rod shaped, impacting end-on; therefore, rods of various diameters and weights have been investigated.
15.5,.1.11-2 12-1-69
- Missile velocities of 100 fps might be generated by rupture of a reactor coolant loop and this value has been used with penetration equations developed by D. A. Davenport 7 to estimate their penetrating capability.
Table 15.5.1.11-1 .summarizes the results of the analysis. Inspection of these results indicates that except for the containment liner, at 100 fps, the i required weight and dimensions for penetration of the metal thicknesses of interest are not credible for missile sizes which can be postulated within theI I, reactor containment. The metal thicknesses shown in the table bracket the thicknesses of interest for the containment liner and piping systems. The analysis for the containment liner does not consider the added resistance to peaetration afforded by the interaction between the concrete containment strudture I and the containment liner. This added resistance will not permit penetration byI missiles of credible weight and size. Major components, such as the steam generator, have greater shell thicknesses than the values in the table, and therefore, will not be penetrated by the postulated missiles. All potential missiles were evaluated, and _those that constitut~ a hazard to either the liner or adjacent equipment, by virtue of their velocity and/or size, are restrained by local barriers or other mechanical means
- 15.5.1.11 12-1-69 TABLE 15.5.1.11-1
_MISSILE ~IMENSIO~S _ANJ?. WEIGHTS ~EQI.JIRED TO PENETRATE PIATE_OF VARYING _THICKNESSES Material Missile Diameter, in. 1 2 3 4 5 Reactor Containment Liner Plate, 3/8 in. Weight, lb 21.1 42 64 85 106 *. Length, in. 95 48 32 24 19 4 in. Sch. 160 Pipe or 0.531 in. Wall Thickness Weight, lb 40.2 80.3 120.5 160.6 200. 8 Length, in. 181 90 60 45 36 6 in. Sch. 160 Pipe or 0.718 in. Wall Thickness Weight, lb Length, in. 68.8 309, 137.5 155 206.3 103 275.0 78 343.8 62 8 in. Sch. 160 Pipe or 0.906 in. Wall Thickness Weight, lb 109.0 218.0 327.0 436.0 545.0 Length, in. 514 245 164 123 99 10 in. Sch. 160 Pipe or 1.125 in. Wall Thickness Weight, lb 176.0 325.0 528.0 704.0 880.0 Length, in. 790 395 264 198 159
15.5.1. U-4 12;_1-69
- B. Exterior Missiles The Surry Power Station site is approximately 12 miles from the nearest com-mercial airport at Newport News, Virginia and 15 miles from Langley Air Force Base. The site is not on the normal approach path to either of these air fields.
An analysis of hypothetical aircraft accidents indicates that the most likely missile which might penetrate the reactor containment would be.a turbojet rotor. Calculations show that the 2 ft-6 in. thick containment dome would withstand witho~t penetration the direct impact of a 1,500 lb rotor impacting at a velocity of 400 mpsh. The 4 ft-6 in. thick containment walls would withstand a sim'ilar missile impacting at a velocity of 980 mph. These velocities are considerably in excess of low-level aircraft approach speeds
- Tornado generated missiles are discussed in Sectiol) 15.2.3 and include two potential'missiles:
- a. Missile equivalent to a wooden utility pole 40 ft long, 12 in.
3 diam, weighing 50 lb per ft and traveling in a vertical or horizontal direction at: 100 mph.
- b. Missile equivalent to' a ton automobile traveling at 150 mph *.
Neither of these missiles would penetrate the reactor contain-ment.
15.5.1.11-5 12-1-69 Pipe Rupture Incident The containment internal structure is designed to acconunodate the loading due to rupture of the reactor coolant and connecting piping or main steam or feed-water piping. Incident rupture was considered in only one lirte at a time. The supp'ort system was designed to pre.elude damage to or rupture of any of the lines as a result of the incident. The snubber and key systems are designed to transmit rupture thrusts from a steam generator into the. containment internal structures. In dete~ining tha steam generator support reactions, the system was reduced to a dynamic model consisting of a suitable number of masses and resistance elements under impulse loading. The structural support sy~te~ resilience and mass was
, I included in the model. The dynamic problem was solved by numerical methods, using a thrust-time history as loading. Resistance, dynamic amplification of the thrust, and rebound.forces were calculated vs. time. Design of the support system was based upon s~ress levels defined in Section 15.5.1.8. The reactor vessel and support system were similarlyt trea,ted.
The steam lines are str.apped to the crane wall at intervals, selected to prevent a whipping pipe from contacting the liner. The straps a~e designed so that no interference with the normal thermal expansion modes of the steam lines results
- 15.5 .1.12-1 12-1-69
- 15.5.1.12 Ground Water Protection and Corrosion The ground water level external to the membrane protection of the exterior surfaces of the containment structure will be kept a maximum of 6 in. below the top surface of the foundation mat by pumps as described in Section 15.5.1. 3.
If water penetrates or otherwise circumvents the membrane, it dtains to a layer of porous concrete directly below the mat and above the membrane. This 4 in.
/
thick layer of porous concrete serves as*a horizontal drain under the entire structure. The porous layer is vented by two 4 in. diam pipes which extend from the underside of the mat into a subsurface cubicle adjacent to the outside of the containment structure. This cubicle is inside the waterproof membrane. Access is provided by a concrete shaft from ground level. The floor of the cubicle is 3 ft below the mat liner, thus flooding of the cubicle would have to occur before any hydrostatic head would be applied to the steel liner. Water level alarms are installed in the 4 in. pipes, and pumps are used as necessary to remove the water~ Vertical drainage to the base of the mat is aided by three vertical inspection shafts and various tunnels and cubicles located adjacent to the exposed exterior face of the concrete containment wall, in which the concrete is exposed. The vertical edge of the mat is drained by a layer of 4 in. thick hollow concrete masonry blocks placed against the inside surface of the membrane. Cathodic protection is not provided, since adequate corros.ion protection of the embedded reinforcing is otherwise provided. Research by the National
- Bureau of Standards and other references indicates that cathodic currents damage
15.5.1.12-2 12-1-69 . the bond between the reinforcing steel and concrete. This bond softening is due to the gradual concentration of sodium and potassium ions. alkali .concentration becomes. strong enough to attack the steel. In time, the
- The surface of the steel liner in cop.tact with concrete is.not subject to corro-
/ .
sion because of the alkaline nature of the concrete. The interior exposed sur-f~ce of the liner is protected by one coat of inorganic zinc silicate paint with one top coat of interior finish paint. No other protective coatings or insulation are considered necessary. REFERENCES
- 1. Newmark, N.M., "Design Criteria for Nuclear Reactors Subjected to Earthquake Hazards , " Urbana, Illinois , May 25, 19 6 7.
- 2. Gondikas, P., Saluadori, M.G., "Wind Stresses in Domes," American Society of Civil Engineers Proceedings No *. 2616, October, 1960.
- 3. Erdei, C., Gosh, J., "The Effects of Wind on Large Diameter Chimneys and Shafts, 11 "concrete Magazine, September, 196 7.
- 4. Stone & Webster Engineering Corporation, "Nuclear Containment Structure Access Opening."
- 5. Budiansky, B., Radkowski, P.,I "Numerical Analysis of . Unsymmetrical Bending of Shells of Revolution," AIAA Journal:, August, 1963.
15.5.1.12-3 12-1-69
- 6. Stone & Webster Engineering Corporation, "Report on Pressure Testing of Reactor Containment for Connecticut Yankee Atomic Power Plant, Connecticut Yankee Atomic Power Company, Haddam, Connecticut," May, 1967.
- 7. Davenport, D.A., "Penetration of Reactor ,Containment Shells," in Nuclear Safety, December, 1960, Volume 2, No. 2.
- 8. Newmark, N.M., Fifth Rankine. Lecture in "Effects of Earthquakes.on Dams and Embankments," Geotechnique, Page 139, Volume XV, Number 2, June, 1965.
Published by The Institution of Civil Engineers, London, England.
- 9. Gere, S., Timoshenko, S., "Theory of Elastic Stability," 2nd Edition *
- 10. Gopdier, j. N. , Timoshenko, S. , "Theory of Elasticity," 2nd Edition
- 15.5.1.13-1 12-1-69 15 * .9.1.13 Testing and In-se*rvice .Surveillance General The completed containment structure was tested for structural integrity by sub-jecting the structure to an air pressure test equal to 115 percent of the design pressure. The structure was first carefully surveyed, measured, and inspected for cracks prior to the test and all measurements recorded. All mea~urements were related to an independent datum. The pressure was then raised in 10 psi increments to the 115 percent test pressure (52 psig) and held at that pressure for 1 hr. Pressure was then reduced to complete the containment liner leak rate test described in Section 5.3.
During the 48 hr period, visual examination was made of the containment exterior surface for cracks and crack patterns as well as distortion. Visual and instrumented observations at each pressure increment were made of the containment response during the test. Crack patterns were observed and their development noted. Temperature, barometric pressure, and weather conditions were recorded hourly during the test period. A further detailed dimensional survey was made of the structure on completion of the tests to record recovery of the structure.
15.5.1.13-2 12-1.:..69
- Test Instrumentation Instrumentation was designed to provide control and information on containment response during the air pressure test. Measurements were made of the radial deflection of the containment wall at selected locations from the top of the containment wall at selected locations from the top of the mat to the spring line o~ the dome. Vertical deflections were measured at the top of the mat and at the top of the d9me. Additional measurements were made around the equipment access hatch and in other areas where stresses were critical.
i Strain gages were attached to the steel liner to record strains at the junction I with the mat liner, at mid-height and at the spring line of the dome. Addi-
,1 I
tional strain gages were attached to the liner around the equipment access and I personnel hatches. I.
. I Exterior. visual *observations, above grade, were obtained using engineer's I
I scales attached to the structure and read by transits placed nearby. Transit i measurem~nts were calibrated with independent datum points. Readings obtained i by this bethod were considered accurate to within 0.10 in. Exterior deformations below grade were measured by linear variable differential transdu~ers (LVDT's) mounted in the two pits provided for this purpose. LVDT's recorde~ displacements in mils, which is an accuracy in excess of that required LVDT's ~ere I also used to measure displacements of the concrete rings surrounding i the equipment access and personnel access hatches *
- I
15.5.1.13-3 12-1-69 Electrical strain gage rosettes and conventional strain gages, reading in microinches per inch, were used to monitor strains in the liner. Since major inaccuracies with this type of gage have resulted from inadequate installation techniques, particular attention was given to the t~chnique used. Redundancy of instrumentation was obtained by multiplicity of points at which measurements were made, such that loss or damage to any one station was not critical. The range of strains and deformations expected at the 45 psig test pressure was as follows:
- a. Maximum vertical elongation of the structure, not more than 1.5 in.
b. c. Increase in containment diameter, not more than 1.4 in. The maximum width of new cracks or increase in existing cracks, not more than 0.03 in. per crack.
- d. After containment pressure was reduced to atmospheric, the residual width of new cracks or the increased width of existing cracks, not more than 0.01 in.
- e. There was no visual distortion of the liner plate.
The containment structure remained in the elastic range during the pressure test and there was permanent distortion in the liner or in the concrete once the pres-sure was reduced to atmospheric or below. However, it was fully expected that there would be small residual cracks in the concrete as a result of shrinkage in the concrete.
15.5.1.13-4. 12-1-69
- Under the test program outlined herein, 'all instruments and measuring devices were installed just prior to the test* and normal care and protection was adequate.
It.ems. damaged for any reason were readily replaced at the initiation of the test. Comparison of Tes.t Results The selection of a test pressure, which was 115 percent of design pressure, was based primarily on the fact that a similar reinforced concrete containment structure for Connecticut Yankee Atomic Power Plant has been tested and accepted I at 115 percent of its design pressure and that thereby a comparative case history of structural response has been created which permit valid comparison of similar designs *
- The selection o.f 115 percent test pressure also conforms to the ASME Boiler and Pressure Vessel Code, Section III, Nuclear Vessels, Subsection B, Require-ments for Class B Vessels, paragraph N-1312 (d). This relates to metal vessels which perform the same function as a reinforced concrete containment structure.
A comparison of stresses under 115 percent test pressure with those in the structure under incident conditions is given in Table 15.5.1.13-1. As a sensitivity analysis, the stresses associated with 125 percent pressure are also included. Incident stresses shown result,from incident pressure, dead load, loads due to temperature effects on the steel liner, and temperature gradients through the *concrete. Stresses resulting from earthquake combined with incident loads are shown separately
- 15.5.1.13-5 12-1-69 Scaled load plots comparing moments, shears, tension, and deflections resulting
- from the structural proof test pressure with moments, shea~s, tension,. and deflections due to the unfactored design incident conditions are shown in Figure 15.5.1.13-1. A comparison of the testrload with the hypothetical incident load conditions should include a review of the load plots in Figure
- 15. 5 .1. 5-2. This shows the inc;reas.e in -momex:its, r~d_ia_l:- shear, hoop tension, vertical tension, and radial deflection (deformation) imposed on the structure by the incident and test load. It is seen that the test load conditions exceed incident conditions in all cases, except that of radial deflection.
The distribution of stress varies between the structural elements under apparently similar load conditions because of the contradictory action of the containment steel liner. Under test conditions the steel liner was in a state of biaxial tension and gave considerable assistance to the reinforcing steel, particularly to the longitudinal reinforcing. Under incident conditions, the steel liner is subjected to a point where it is restrained in compression by the reinforcing steel. This effect is greater in the dome than in the cylindrical wall due to the increased th~ckness of the dome liner and the shape factor of the come. This also provides an addftional factor of safety against ultimate failure of the structure. In the event of excessive yeild in the reinforcing steel, the liner will act as a tensile membrane which would assist the reinforcing steel. This assistance would be significant since the liner will bring a considerable reserve of energy to bear for which design credit has not otherwise been claimed.
15.5.1.13-6 12-1-69
- The test pressure of 52 psig, based on 115 percent design pressure, created stresses equal to or greater than the incident stresses in the following critical areas:
- 1. Foundation mat, where test stresses are 30-40 percent above incident conditions.
- 2. Large access openings, such as equipment and personnel hatches where test st.resses are conipar~ble with incident stresses.
- It is recognized that the average stress levels attained under the test condi-tions in the principal longitudinal and circumferential steel are below these stresses resulting from incident conditions. This is considered acceptable when
- the test is associated with dimensional strain measurements and when such test provides confirmation of structural continuity, structural ductility with the concrete cracked, and where the steel is shown to carry the load in tension according to design assumptions.
An analysis of the crack pattern of the concrete under test conditions--confirms stress distribution in the structure and also reveals areas of stress concentra-tions. In fact, a pattern pf .severe local cracking would indicate structural weakness more effectively than cqnsiderations of average stress levels. Measured response of the structure, as'indicated by increase in height, diameter, and degree of recovery, together with measurements of local deformations, are all extremely important in predicting structural response to incident conditions. The structural response to the test pressure is of sufficient magnitude to allow
15.5.1.13-7 12-1-69 simple direct measurement of deformations without the need for high precision methods of measurement. In sununary, it is not possible to exactly duplicate incident stress conditions with a pressure test. An increase in the test pressure above 115 percent would only preserve and amplify the present stress anomalies without furnishing more meaningful data. In addition such a test would endanger or damage the con-tainer by seriously overstressing critical areas, or it would require a con-tainer design modification directed specifically to withstand the higher pres-sure test without proportionate improvement in withstanding the incident con-dition. Modifica~ion of the containment design to obtain closer test verifi-cation of structural integrity under the test pressure would require redesign specifically for test conditions of the critical areas in the foundation mat and at the large openings. Such redesign would not improve the capability of the containment structure to meet the incident load conditions. A design meet-ing both incident an4 test conditions is not considered practical in this type of containment design. The 115 percent pressure test provided a valid test of all critical areas with stresses equal to or greater than incident conditions; in L~ss critical areas, the pressure test provided sufficient information to permit a rel~able evalua-tion of the complete structural response under incident conditions. The average anticipated crack width at the 45 psig test pressure was 0.015 in.
15.5.1.13-8 12-1-69
- A rectangular crack pattern was anticipated, w~th vertical cracks spaced 12 to 15 in. on centers and horizontal cracks spaced approximately 2 ft on centers.
Ht>rizontal crack spacing was primarily controlled by the horizontal construction joints. The average crack width was related to the anticipated increase in containment diameter, the anticipated vertical elongation ~f the structure, and the crack spacing. It was assumed that *the total containment extension was equal to the sum of the number of cracks and the average crack width in each direction. Maximum summer temperature and minimum winter temperature difference is approxi-
*o . o
- mately 95 F. Annual average temperature variation is 40 Fat the station site *
- During unit operation, the annual maximu~ thermal cycling temperature variation is approximately 45° F.
Ambient t~mperature variations of this magnitude, +20 0 F, or even the extreme 0 . .
+45 F, will not reopen the crack pattern created in the structure by the test pressure of 45 psig, by any significant amount.
The width of thermal cycling cracks were significantly less than the 0.010 in. allowed for exterior members by ACI 318. The stresses given in Table 15.5.1.13-1 are the results obtained. from computer programs referred to in the following sections:
15.5.1.13-9 12-1-69 Section 15. 5. 1. 5 "Numerical Analysis of Unsymmetrical Bending of Shells of Revolution"
- Section 15.5.1.4 '"container Vessel Seismic Analysis" Section 15.5.1.5 "Flat Circular Mat Foundations for Nuclear Secondary Containment Structures" Section 15. 5 .1. 7 ."Nuclear Containment Structure Access Opening" -
Stone & Webster Computer Program At large openings, the stresses due to thermal load were obtained by converting the thermal effect to a pressure equivalent as described in Section 15.5.1.5
- Since all of the shears in the wall and dome were taken by the reinforcing, the effects of shrinkage and creep are not included.
In-Service Surveillance Tests Periodic structural testing of the containment structure is not planned since it would provide no more information on the containment structure capability than that obtained from the initial test. In fact, periodic testing would cum-mulatively damage the concrete in the structure to the point where the test, itself, would be the major cause of structural deterioration.
15.5.1.13-10 12-1-69
- The in-service stress and environmental conditions are not of such a nature or magnitude that any significant deterioration, of the reinforcing steel or con-crete could reasonably be expected, and periodic testing for structural purposes could be duplicated if at any time further tests were required. The minimum test level required to verify continued structural integrity would be no less than the 115 percent, or 52 psig initial test pressure.
Periodic inspection of the steel liner is accomplished by leak rate test and gas testing of all welded liner seams and gaskets. All welded joints and all penetrations of the liner are designed for periodic halogen gas testing. In summary, no basis exists for attempting to develop structural performance
- information from leak rate tests conducted at moderate pressures
- 15.5.1.13-11 12-1-69 TABLE 15.5.1.13-1
_COMPARISON OF STRESSES UNDER TEST PRESSURE WITH ST.RESSES UNDEJi INCIDENT CONDITIONS AND EARTHQUAKE PLUS INCIDENT. CONDITidNS Earthquake Plus 115 Percent 125 Percent Incident Incident Test Stress Test Stress Stress ,psi Stress ,psi psi(52 psig) psi(57 psig) Foundation Mat Top bars at cylinder wall 25,700 26,500 33,500 38,700 Bottom bars near mat 26,300 30,500 29,000 38,700 center Cylinder Wall (Approximately midheight) Circumferential reinforcing Inner 22,300 22,300 21,000 2z,soo Outer 28,600 28,600 20,000 21,700 Longitudinal reinforcing Inner 19,800 21,600 9,200 10,000 At base of wall 12,900 19,300 16,400 19,300 Outer 27,900 29,650 9,200 10,000 Diagonal reinforcing 28,300 33,600 15,500 16,900 Diagonal (radial) shear reinforcing at base of wall 15,700 16,500 20,200 23,200 Dome Radial reinforcing Inner 28,900 28,900 14,200 15,500 Outer 34,500 34,500 13,800 15,000 Circumferential reinforcing Inner 28,900 28,900 14,200 15,500 Outer 34,500 34,500 13,800 15,000
15.5.1.13-12 12-1-69
- TABLE 15.5.1.13-1 (Cont'd)
Earthquake Plus 115 Percent 125 Percent Incident Incident Test Stress Test Stress
*Stress ,psi Stress ,psi psi(52 psig) psi (57 psig)
Large Openings Equipment access hatch 32,000 33,500 31,300 33,700 Personnel access hatch 30,200 31,700 30,700 34,300 __J
FIG. 15.5.1.13 PEC. I, 1969
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15.6-1 12-1-69 *
- 15.6 OTHER CLASS I STRUCTURES Class I Structures other than the reactor containment structure are listed in Table 15~2.l-l. The major structures include the Auxiliary B~ilding, Main Control Room at~a including switchgear and relay ro0111s, Fuel Building, auxiliary generator cubicle~ Auxiliary Containment Buildings whicb contain main steam and feedwater isolation valves, recirculation spray and low head safety injection pump cubicles, auxiliary steam generator feed pump cubicle and safe-guards ventilation room, circulating water intake structures including the high level canal.
The Fuel Building and the main steam and feedwater isolation valve section of the
\
- Auxiliary Containment Buildings are supported on reinforced concrete mats on con-crete filled steel pipe piles. All othe~ structures are soil supported on rein-forced cpncrete mats or spread footings. All Class I structures designed to meet both earthquake ana tornado criteria, as listed in Table 15.2.1-1, are enclosed with heavily reinforced 2 ft thick concrete walls and roof slabs with all openings shielded against missiles.
Class I structures are designed to resist the operational basis earthquake without exceeding allowable working stresses where allowable stresses are one-third above the normal applicable code normal working stress. For concrete structures a 5 percent critical damping function is assumed and the accelerations selected from the acceleration response spectrum curves considered in conjunction with the natural frequency of each structure. A check hf.ls been made to ensure that collapse type failures will not occur under the design basis earthquake. For this
15.6-2 12-1-69 condition a 10 percent damping factor is assumed for concrete structures and stresses are limited to not moi*e than 120 percent of the minimum yield point stress. In practice, the controlling feature of the design of these structures was meeting the operational basis earthquake requirements with the limited 5 percent damping factor. The high level intake canal is formed by excavating to El.+5 ft from an average grade of approximately 35 ft. Earth fill dikes constructed on either side of the canal bring the finished height to El.+36 ft throughout the length of the canal. The interior surfaces of the canal are lined with a 4 1/2 in. thick reinforced concrete slab. Under drains and pressure relief valves are provided to prevent uplift of the concrete liner by unbalanced hydrostatic pressure. 15.6.1 OTHER STRUCTURES All other structures are designed to adequately support all dead, live and wind loads. Where necessary, subsurface walls and slabs are designed to resist the horizontal component of the soil with applicable surcharge and hydrostatic pressures. Structural steel design conforms to the 1963 issue of the "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings" of the American Institute of Steel Construction. All concrete work is designed in accordance with the Building Code Requirements for Reinforced Concrete, serial designation 318-63 of the American Concrete Institute. Access and egress require-
15.6-3 li-1-69
- ments, as well as fire ratings of walls and floor systems, satisfy the require-ments of the Basic Building Code of the Building Officials Conference of America, 1966 issue.*
Allowable soil bearing values for foundations are determined from the soil boring logs and the results of triaxial shear tests of the soil. Applicable factors of safety are applied to the test results. 15.6.2 REACTOR COOLANT SYSTEM SUPPORTS The Reactor Coolant System includes the reactor vessel, three steam generators,* three reactor coolant pumps, and a pressurizer for each unit. Structures are
- provided to support these heavy vessels and equipment to ensure system integrity during normal operation and Design Basis Accident conditions.
Design Basis All supports in the Reactor Coolant System are designed to withstand the Design Basis Earthquake acting simultaneously with an instantaneously applied pipe break. Two types of piping failures are considered: a double-ended rupture, or a longi-tudinal rupture on either the horizontal or vertical axis of the pipe. The lon-gitudinal rupture area was taken equal to the area of the double-ended rupture for these piping failures. These failures are assumed to occur in either the reactor coolant piping, pressurizer surge piping, or the main steam piping
- 15.6-4 12-1-69 The peak value of the pipe thrust for any of the reactor coolant loop pipe breaks ~
considered is approximately 1,500,000 lb; the peak value of the pipe thrust for any of the main steam piping breaks considered is approximately 620,000 lb; and the peak value for the pipe thrust for the pressurizer surge pipe break is approximately 195,000 lb. These thrust values are equal to PxA, the system pres-sure x rupture area. The load vs time transients of these breaks are computed by a computer program that analyzes the shock wave initiated at the break as it passes through the complete piping loop. Results from this program are used as forcing functions in a structural dynamic program that results in the dynamic loadings on the sup-ports. The supports are stress analyzed for these loadings simultaneously with the.design basis earthquake, and all stresses are maintained within 90 percent
- of the minimum yield point of the structural material used.
All welding is in accordance with Section IX of the ASME Code, and all welds are examined by either radiographic, sonic, dye penetrant, or magnetic particle tech-niques, depending on the material, and the state of stress at the weld. Description a) Reactor Vessel Support The reactor vessel is supported by six sliding foot assemblies mounted on the neutron shield tank. The support feet are designed to restrain lateral and rota-tional movement of the reactor vessel while allowing thermal expansion. The neu-tron shield tank is a double walled cylindrical structure which transfer~ the
15.6-5 la-1-69
- loadings to the heavy reinforced concrete mat of the containment structure.
tank also serves to minimize gamma and neutron heating of the primary concrete The shield, and.to atte~uate neutron radiation outside of the primary shield to acceptable limits (Section 11.3.2 .1). The neutron shield tank assembly and mate-rial listing are shown on Fig. 15.6-1. b) Steam Generator Support The steam generator support consists of two (upper and lower) cast rings and associated s'uspension rods and snubbers.. The lower ring which carries the steam generator weight is suspended by means of three pipe columns. Hydra~lic snubber cylinders connect the upper and lower casting to the steam generator cubicle walls
- to allow pip;ing expansion during norma.1 operation while resisting movement during seismic and pipe break conditions. The steam generator support assembly and mate-rial listing is shown on Fig. 15.6-2.
I c) Reactor Coolant Pump Support The reactor coolant pumps are supported by a four-legged suspended structure. The frame structure is secured to the cubicle side walls by tie-rods to prevent the pump from becoming a credible missile. Snubber cylinders located in the frame structure and in the tie-rods allow for thermal expansion during normal operation and resist movement during earthquake and pipebreak conditions. The reactor coolant pump support assembly and material listing are shown on Fig. 15.6-3
- 15.6-6 12-1-69 Pressurizer Support The pressurizer vessel is mounted to* a rigid support ring box girder which is suspended by four hanger columns from above. To offset the flexibility of the hanger columns, antisway brackets are welded to the shell of the pressurizer to accommodate shear blocks which are embedded in the concrete floor close to the vessels center of gravity. The shear blocks are able to take all incident loads and also allow the pressure vessel to expand vertically. The pressurizer support assembly and material listing are shown on Fig. 15.6-4.
Evaluation All pipe break forces are multiplied by a dynamic multiplication factor, and ~ach of these forces combined with the seismic loadings to ensure that the supports are conservatively designed to withstand the condition of a pipe break occurring as a result of an earthquake. Rigid Quality Assurance criteria during fabrication ensures conformance with the conservative design. 15.6.3 CONTAINMENT INTERNAL STRUCTURE The reactor containment internal structure is a reinforced concrete structure which furnishes: .1. Supports and restraints for all internal equipment and piping including the polar crane
15.6-7 12-1-69
- 2.
Missile shielding for the containment steel liner and main steam lines against internally generated missiles
- 3. Biological shielding for station operators inside the containment structure under all phases of reactor operation The structure is designed to withstand the Design Basis Earthquake together with the simultaneous loss-of-coolant accident without loss of function. Clearance is maintained between all internal structures and the steel liner of the reactor containment shell to permit differential earthquake motion. The steam generator cubicles and the pressurizer cubicle are designed to withstand an internal dif-ferential pressure load of 35 psi resulting from the postulated double-ended
- primary coolant pipe break. The primary shield is designed to withstand an internal pressure of 100 psig resulting from a hypothetical reactor coolant pipe break with-in the primary shield.
The differential pressure rise within ~he cubicles is controlled by open and shielded vent spaces in each cubicle which permit rapid pressure equalization within the containment structure. This transient pressure condition has been calculated by Stone & Webster's CUPAT Program, using input from the LOCTIC Program. Temperature differentials between c~bicles are considered coincident with the pressure differentials. The short duration of the transient accident relative to
15.6-8 12-1-69 the low thermal conductivity of the concrete is such that no significant tempera-
- ture gradient occurs across the walls. Also, the transient accident is not con-sidered as adding to the differential cubicle wall loadings.
Structural concrete design conforms to the requirements of ACI 318, Part IV-B, Ultimate Strength design. Maximum stresses are limited to the 90 percent of the minimum yi~ld point in bending or 85 percent of the minimum yield point in diagonal tension, bond, and anchorage. Special large size reinforcing steel bars No. 14 and No. 18 are controlled chem-istry steel of 50,000 psi yield point, otherwise conforming to the requirements of ASTM A408. All other reinforcing steel is steel of 40,000 psi yield point conforming to ASTM A-15 and ASTM A305. Structural steel columns placed in the annular space between the interior concrete and the steel liner support grating, concrete walk.ways, and cable trays. A stainless steel lined fuel transfer canal and reactor refueling cavity is
- incorporated in the concrete structure above the reactor vessel. A 1/4 in. thick stainless steel plate is used to prevent leakage of water from these areas into the containment structure.
Fl G. 15. 6.-:-1 DEC. I, 1969 BILL OF MATERIALS REACTOR NEUJRON SHIELD TANK ITEM NO. NAME MATERIN...
- ---t-------._ µ'/ (i) PLATE A ASTI1*A516 G~ 60 FBX !NDRWJ..lZEDl I ~ ..... *-" CD 0)
PLATE B PLATE ASTM*A516 GR 60 FBX (AUSTENIZEDl ASTIH21f0 TYPE 316L (USE FDR 7" ID TIJBESl i -- . - . ii G) PIPE A ASTM*A312 TYPE316L SST ASTM*A106 GR A OR B CARBON VESSEL SU'P'PO'QT ASSEMELY PIPE C ir"*--=---- ' __:. -~ / .. STEB.. ACTo* REACTOR VESSEL SUPPORT ASSEMBLY
~-~
ITEM NO. NAME MATER IM.. I I 6) Bhl PLATE I-IARAGED STER. 330. 000 PSI HIN Y. ~' MARAGED STE8.., 270,000 PSI M!N,Y.P.' ___._ --NEUTRON ' FORGING KARAGED STEB.., 270,000 PSI MIN Y. P. f ETECTOR TRANSFER PIPE 0 4 8 SCALE-FEET
~
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12-1-69
; APPENDIX A REPORT, SITE ENVIRONMENTAL STUDIES, SURRY POWER STATION This report is a separate bound document which has previously been submitted to the Atomic Energy Commission as Appendix A of the Surry Preliminary Safety Analysis Report .
TS-i 2-13-70
- TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION TITLE PAGE 1.0 DEFINITIONS TS 1.0-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM TS 2.1-1 SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE TS 2.1-1 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE TS 2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE TS 2.3-1 INSTRUMENT AT ION 3.0. LIMITING CONDITIONS FOR OPERATION TS 3.1-1 3.1 REACTOR COOLANT SYSTEM TS 3.1-1 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM TS 3.2-1 1 .
- 3.3 SAFETY INJECTION SYSTEM TS 3.3-1 3.4 SPRAY SYSTEMS TS 3.4-1 3.5 RESIDUAL HEAT REMOVAL SYSTEM TS 3.5-1 3.6 TURBINE. CYCLE TS 3.6-1 3.7 INSTRUMENTATION SYSTEM TS 3.7-1 3.8 CONTAINMENT TS 3.8-1 3.9 STATION SERVICE SYSTEMS TS 3.9-1 3.10 REFUELING TS 3.10-1 3.11 EFFLUENT RELEASE TS 3.11-1 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION TS 3.12-1 LIMITS 3.13 COMPONENT COOLING SYSTEM TS 3.13-1 3.14 CIRCULATING AND SERVICE WATER SYSTEMS TS 3.14-1
TS-ii 12-29-71 SECTION TITLE PAGE 3.15 CONTAINMENT VACUUM SYSTEM TS 3.15-1 3.16 EMERGENCY POWER SYSTEM TS 3.16-1 3.17 LOOP STOP VALVE OPERATION TS 3.17-1 3.18 MOVEABLE INCORE INSTRUMENTATION TS 3.18-1 4.0 SURVEILLANCE REQUIREMENTS TS 4.0-1 4.1 OPERATIONAL SAFETY REVIEW TS 4.1-1 4.2 REACTOR COOLANT SYSTEM COMPONENT TESTS TS 4.2-1 4.3 REACTOR COOLANT SYSTEM INTEGRITY TESTING TS 4.3-1 FOLLOWING OPENING 4.4 CONTAINMENT TESTS TS 4.4-1 4.5 SPRAY SYSTEMS TESTS TS 4.5-1 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING TS 4.6-1 4.7 MAIN STEAM LINE TRIP VALVES TS 4.7-1 4.8 AUXILIARY STEAM GENERATOR FEEDWATER PUMPS TS 4.8-1 4.9 EFFLUENT SAMPLING AND RADIATION MONITORING SYSTEM TS 4.9-1 4.10 REACTIVITY ANOMALIES TS 4.10-1 4.11 SAFETY INJECTION SYSTEM TESTS TS 4.11-1 4.12 VENTILATION FILTER TESTS TS 4.12-1 5.0 DESIGN FEATURES TS 5.0-1 5.1 . SITE TS 5.0-1 5.2 CONTAINMENT .TS 5.2-1 5.3 REACTOR TS 5.3-1 5.4 FUEL STORAGE TS 5.4-1 6.0 ADMINISTRATIVE CONTROLS TS 6.1-1 6.1 ORGANIZATION, SAFETY AND OPERATION REVIEW TS 6.1-1 6.2 ACTION TO BE TAKEN IN THE EVENT OF AN TS 6.2-1 e ABNORMAL OCCURRENCE IN STATION OPERATION
TS-iii 5-1-71
- STATION 6.3 TITLE ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED PAGE TS 6.3-1 6.4 UNIT OPERATING PROCEDURES TS 6.4-1 6.5 STATION OPERATING RECORDS TS 6.5-1 6.6 STATION REPORTING REQUIREMENTS TS 6.6-1
TS 1.0-1 12-29-71 1.0 DEFINITIONS The following frequently used terms are defined for the uniform interpretation of the specifications. A. Rated Power
*A steady state nuclear steam supply output of 2441 MWt.
B. Thermal Power The total core heat transferred from the fuel to the coolant. C. Reactor Operation
- 1. Refueling Shutdown Condition
- e. When the reactor is subcritical by at least 10% ~k/k and* Tavg is
<140°F and fuel is scheduled to be moved to or from the reactor core,
- 2. Cold Shutdown Condition When the reactor is subcritical by at least 1% ~k/k and T is <200°F~
avg
- 3. Intermediate Shutdown Condition When the reactor is subcritical by an amount greater than or equal to the margin as specified in Technical Specification Figure 3.12-2 and 200°F <T <547°F.
avg
- 4. Hot Shutdown Condition When the reactor is subcritical by an amount greater than or equal to the margin specified in Technical Specification Figure 3 .12-2 and -"T-
- is ~ 54,7-!?.F.
avg
TS 1.0-2 12-29-71
- 5. Reactor Critical When the neutron chain reaction is self-sustaining and keff = 1.0.
- 6. Power Operation When the reactor is critical and the neutron flux power range instrumentation indicates greater than 2% of rated power.
- 7. Refueling Operation Any operation involving movement of core components when the vessel head is unbolted or removed.
D, Operable A system or component is operable when it is capable of performing its in-
~ended function within the required range. The system or*component shall be con-sidered to have this capability when: (1) it satisfies the limiting conditions for operation defined in Section 3, and (2) it has been tested periodically in accordance with Section 4 and meets its performance requirements.
E, Protective Instrumentation Logic
- 1. Analog Channel An arrangement of components and modules as required to generate a single prot~ctive action digital signal when required by a unit condition. An analog channe_l loses its identity where single action signals are combined.
TS 1.0-3 5-1-71
- 2. Logic Channel A logic channel is a group of relay contact matrices which operate in response to the digital output signal from the analog channel to generate a protective action signal.
F. Degree of Redundancy The difference between the number of operable channels and the minimum number of channels monitoring a specific parameter which when tripped will cause an automatic system trip. G. Instrumentation Surveillance
- 1. Channel Check A qualitative determination of acceptable operability by observation of channel behavior during operation. This determination shall include comparison of the channel with other independent channels measuring the same variable.
- 2. Channel Functional Test Injection of a simulated signal into an analog channel or makeup of the logic combinations in a logic channel to verify that it is operable, including alarm and/or trip initiating action
- TS 1. 0-4 2-1-72
- 3. Channel Calibration Adjustment of channel output such that it responds, with acceptable range and accuracy, to known valu'es of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment action, alarm, or trip, and shall be deemed to include the channel functional test.
H. Containment Integrity Containment integrity is defined to exist when:
- 1. All non-automatic containment isolation valves, except those required for intermittent operation in the performance of normal operational activities, are locked closed and under administrative control. Non-automatic containment isolation valves may be opened intermittently for operational activites provided that they are under administrative control and are capable of being closed immediately if required.
- 2. Blind flanges are installed where required.
- 3. The equipment access hatch is properly closed and sealed.
- 4. At least one door in the personnel air lock is properly closed and sealed.
- 5. All automatic containment isolation valves are operable or are locked closed under administrative control.
- 6. The uncontrolled containment leakage satisfied Specification 4.4.
TS1. 0-5 2-1-72
- I. Abnormal Occurrence An abnormal occurrence is defined as:
- 1. Any unit condition that results in exceeding a safety limit or that results in safety system settings less conservative than the limiting safety system settings delineated in these Technical Specifications.
- 2. Any unit condition that results in violation of a limiting condition for operation as established in these Technical Specifications.
- 3. Any uncontrolled or unplanned release of radioactivity from the site.
- 4. Any abnormal degradation of one of the several boundaries which are designed to contain radioactive materials resulting from the fission process
- 5. Uncontrolled or unanticipated change in reactivity, except for reactor trip greater than 1%Ak/k.
- 6. Engineered Safeguard System malfunction or other component or system malfunction which rendered or could render the Engineered Safeguard System incapable of performing its intended safety function.
J, Unusual Safety Related Event Any unusual safety related event is defined as:
- 1. Discovery of any substantial errors in the transient or accident analyses, or in the methods used for such analyses, as described in the Final Safety Analyses Report or in the bases for the Technical Specifications.
- 2. Any substantial variance, in an unsafe or less conservative direction, from performance specifications contained in the Technical Specifications or from performance specifications, relevant to safety related equipment,
- contained in the Final Safety Analysis Report
- TSl.0-6 2-1-72
- 3. Any observed inadequacy in the implementation of administrative or procedural controls during the operation of the facility which would significantly affect the safety of operations.
- 4. Any condition involving a possible single failure which, for a system designed against assumed single failures, could result in a loss of the capability of the system to perform its safety function.
- 5. Occurrences or conditions involving an offsite threat to the safety of operation of the facility, including tornadoes, earthquakes, flooding, repetitious aircraft overflights, attempted sabotage, civil disturbances, etc.
K. Quadrant Power Tilt The quadrant power tilt is defined as the ratio of the maximum upper excore detector current to the average of the upper excore detector currents or the ratio of the maximum lower excore detector current to the average of the lower excore detector currents whichever is greater. If one excore detector is out of service, the three in-service units are used in computing the average. L. Low Power Physics Tests Low power physics tests are tests conducted below 5% of rated power which measure fundamental characteristics of the reactor core and related instrumen-tation. M. Interim Limits Additional limitations are imposed upon reactor core power distribution beyond
TS 1.0-7 2-1-72 previously established design bases consistent with interim bases for core cooling analysis established by the AEC in 1971. Two sets of power distribution parameters are shown; both sets are to be met.
TS 2.1-1 12-29-71 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of thermal power, Reactor Coolant System pressure, coolant temperature and coolant flow when a reactor is critical. Objective e* To maintain the integrity of the fuel cladding. Specification A. The combination of reactor thermal power level, coolant pressure, and coolant temperature.shall not:
- 1. Exceed the limits shown in TS Figure 2.1-1 when full flow from three reactor coolant pumps exist.
- 2. Exceed the limits shown in TS Figure 2.1-2 when full flow from two reactor coolant pumps exist and the reactor coolant loop stop valves in the non-operating loqp are open.
- 3. Exceed the limits shown in TS Figure 2.1-3 when full flow from two reactor coolant pumps exist and the reactor coolant loop stop valves in the non-operating loop are closed.
TS 2.1-2 12-29-71 B. The safety limit is exceeded if the combi~ation of Reactor Coolant System average temperature and thennal power level is at any time above the appropriate pressure line in TS Figures 2.1-1, 2.1-2 or 2.1-3. Basis To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the reactor coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed Departure from Nucleate Boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would -result in high clad temperatures and the possibility of clad failure. DNB is not, how-ever, an observable parameter during reactor operation. Therefore, the obser-vable parameters; thennal power, reactor coolant temperature and pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation µas been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ra.tio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNB ratio (DNBR) during steady state operation, nonnal operational transients and anticipated transients, is limited to 1.30. A DNBR
TS 2.1-3 12-29-71 of 1.30 corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. (1) The curves of TS Figure 2.1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (three loop operation) represent the loci of points of thermal power, coolant system I average temperature, and coolant system pressure for which the DNB ratio is not less than 1. 30. The area where clad integrity is assured is below these lines. In order to completely specify limits at all power levels, arbitrary constant upper limits of average temperature are shown for each pressure at powers lower than approximately 75% of rated power. The temper-ature limits at low power are considerably more conservative than would be required if they were based upon a minimum DNB ratio of 1.30 but are such that the plant conditions required to violate the limits are precluded by the.self actuated safety valves on the steam generators. The curves of TS Figures 2.1-2 and 2.1-3, which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (two loop operation), represent the loci of points of the~al power, coolant system average temperature, !'l,Ild coolant system pressure for which either the DNB ratio is equal* to 1. 30 or the average enthalpy at the exit of the core is equal to the saturation value. At low pressures or high temperatures the average enthalpy at the exit of the core reaches saturation before the DNB ratio reaches 1.30 and, thus, this arbitrary limit is conservative with respect to maintaining clad integrity. ;r:n order to completely specify limits at all
TS 2.1-4 2-1-72 power levels, arbitrary constant upper limits of average temperatures are shown for each pressure at powers lower than approximately 45% of rated power. The limits at low power as well as the limits based on the average enthalpy at the exit of the core are considerably more conservative than would be required if they were based upon a minimum DNB ratio of 1.30. The plant con-ditions required to violate these limits are precluded by the protection system and the self actuated safety valves on the steam generator. Upper limits of 70% power for loop stop valves open and 75% with loop stop valves closed are shown to completely bound the area where clad integrity is assured. These latter limits are arbitrary but cannot be reached due to the Permissive 8 protection system setpoint which will trip the reactor on high nuclear flux when only two reactor coolant plllll.ps are in service. Operation with natural circulation or with only one loop in service is not allowed since the plant is not designed for continuous operation with less than two loops in service. The curves are based on the following nuclear hot channel factors. These hot channel factors, instead of the interim ECCS criteria peaking factors, are utilized, because curves generated with the interim peaking factors would be less conservative. FN = 2.72 q N F 1.58
~H These hot channel factors are higher than those calculated at full power over the range between that of all control rod assemblies fully withdrawn to maxi-mum allowable control rod assembly insertion. The control rod assembly
TS 2.1-5 l 12-29-71 - insertion limits are covered by Specification 3.12. Adverse power distribution factors could occur at lower power levels because additional control rod assemblies are in the core; however, the control rod assembly insertion limits dictated by TS Figure 3.12-1 ens.ure that the DNBR is always greater at partial power than at full power. The hot channel factors are also sufficiently large to account for the degree of malpositioning of part-length control rod assemblies that is allowable before the reactor trip setpoints are reduced and control rod assembly with-drawal block and load runback action may be required. ( 2 ) The Reactor Control and Protection System is designed to prevent any anti-cipated combination of transient conditions for Reactor Coolant System temp-erature, pressure and thermal power level that would result in a DNB ratio of less than 1.30(3) based on steady state nominal operating power levels less than or equal to 100%, steady state nominal operating Reactor Coolant System average temperatures less than or equal to 574.4°F and a steady state nominal operating pressure of 2235 psig. Allowances are made in initial conditions assumed for transient analyses for steady state errors of +2% in power, +4°F in Reactor Coolant System average temperature and +30 psi in pressure. The combined steady state errors result in the DNB ratio at the start of a transient being 10 percent less than the value at nominal full power operating conditions. The steady state nominal operating parameters and allowances for steady state errors given above are also applicable for two loop operation except that the steady state nominal operating power level is less than or equal to 60%. (1) FSAR Section 3.4 (2) FSAR Section 3.3 (3) FSAR Section 14.2
TS FIGURE 2.1-1 e 6-30-71 SAFETY LIMITS REACTOR CORE THERMAL AND HYDRAULIC, THREE LOOP OPERATION, 100% FLOW 645
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TS FIGURE 2.1- 2 5-31-71
- SAFETY LIMITS REACTOR CORE, THERMAL AND HYDRAULIC, OPERATION' TWO LOOP LOO p STOP VALVES OPEN 650 ,-- ~--*, . I 0
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580 tiil~ ++l +
+
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Bi+/- lfiti** till :fJt + ). + .* +. 570 ffiftfi# iHl t!g J# t:.+, .
+*
0 10 20 30 ~o 50 60 70 PERCENT RATED CORE POWER
TS 2.2-1 2-13-70
- 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE Applicability Applies to the maximum limit on Reactor Coolant System pressure.
Objective To maintain the integrity of the Reactor Coolant System. Specification
- The Reactor Coolant System pressure shall not exceed 2735 psig with fuel assemblies installed in the reactor vessel.
Basis The Reactor Coolant System(!) serves as a barrier which prevents radionuclides contained in the reactor coolant from reaching the environment. In the event of a fuel cladding failure the Reactor Coolant System is the primary barrier against the release of fission products. The maximum transient pressure allowable in the Reactor Coolant System pressure vessel under the ASME Code, Section III is 110% of design pressure. The maximum transient pressure allowable in the Reactor Coolant System piping, valves and fittings under USAS Section B31.l is 120% of design pressure. Thus, the safety limit of 2 2735 psig (110% of design pressure) has been established. < )
TS 2.2-2 2....13... 70 The nominal settings of the power-operated relief valves at 2335 psig, the reactor high pressure trip at 2385 psig and the safety valves at 2485 psig are established to assure never reaching the Reactor Coolant System pressure safety limit. The initial hydrostatic test will be conducted at 3107 psig to assure the integrity of the Reactor Coolant System. (1) FSAR Section 4 (2) FSAR Section 4.3
TS 2.3-1 2-1-72 2,3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to trip and permissive settings for instruments monitoring reactor power; and reactor coolant pressure, temperature, and flow; and pressurizer level. Objective To provide for automatic protective action in the event that the principal process variables approach a safety limit. Specification A, Protective instrumentation settings for reactor trip shall be as follows:
- 1. Startup protection (a) High flux, power range (low set point) -
~ 25% of rated ~ower.
(b) High flux, intermediate range (high set point) - current equivalent to< 25% of full power. 6 (c) High flux, source range (high set point) - neutron flux< 10 counts/sec.
- 2. Core Protection (a) High flux, power range (high set point) -
..'.:. 109% of rated power.
(b) High pressurizer pressure - < 2385 psig, (c) Low pressurizer pressure - > 1860 psig.
TS2.3-2 2-1-72 (d) Overtemperature ~T
~T5_t1T 0 [K1 - Kz (T - T') + K3 (P - P') - f ( M) ]
where
~T 0 = Indicated AT at rated thermal power, °F T = Average coolant temperature, °F T' = 574.4°F P = Pressurizer pressure, psig P' = 2235 psig K1 = A constant= 1.12 (3 loop operation and 2 loop operation with the loop stop valves closed in the inoperable loop) 0.94 (2 loop operation with the loop stop valves in the inoperable loop open.)
K2 A constant= 0.0113 K3 = A constant= 0.00056 and f(~I) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be individually selected based on measured instrument response during startup tests such that: (1) for (qt - qb) within+/- 19%, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt+ qb is total core thermal power in percent of rated thermal power, f(AI) = 0. (2) for each percent that the magnitude of (qt - qb) exceeds +19% the AT trip set point shall be automatically reduced by 2% of its value at rated power.
TS 2.3-3 2-1-72 (e) Overpower 6T 6T<tiT [ K4 - K5 dT - K6 (T - T') - f (61) ]
- 0 dt where t§.0 = Indicated KI'. at rated thermal power, Op T = Average coolant temperature, Op T' = Average coolant temperature measured at nominal conditions anr rated power, OF K4 = A constant= 1.10 KS =1-0 for decreasing average temperature
-A constant, for increasing average temperature, 0.2 sec/ op K6 = A constant (for T > T') = 0. 00083; K6:0 (for T< T')
f(/il) as defined in (d) above, e (f) Low reactor coolant loop flow - ~ 90% of normal indicated loop flow as measured at elbow taps in each loop (g) Low reactor coolant pump motor frequency - ~ 57.5 Hz (h) Reactor coolant pump under voltage - > 70% of normal voltage
- 3. Other reactor trip setting (a) High pressurizer water level - < 92% of span (b) Low-low steam generator water level > 5% of narrow range instrument span (c) Low steam generator water level - > 15% of narrow range instrument span in coincidence with steam/feedwater mismatch flow - < 1. Oxl0 6 lbs/hr '
- (d)
(e) Turbine trip Safety Injection, - Trip settings for Safety Injection are detailed in T.S. Section 3.7.
TS 2.3-4 2-1-72 B. Protective instrumentation settings for reactor trip interlocks shall be as follows:
- 1. The reactor trips on low pressurizer pressure, high pressurizer level, turbine trip, and low reactor coolant flow for two or more loops shall be unblocked when power> 10% of rated power.
- 2. The single loop loss of flow reactor trip shall be unblocked when the power range nuclear flux.:::._ 50% of rated power. During two loop operation with the loop stop valves in the inactive loop open, this blocking setpoint, established by Permissive 8, may be increased to 60% of rated power only after the overtemperature ~T setpoint is adjusted to the mandatory two loop value. For two loop operation with the loop stop valves of the inactive loop closed, Permissive 8 may be increased to 65% of rated power after the stop valves are closed. The overtemperature ~T setpoint may remain at the value for three loop operation during two loop operation with the inactive loop stop valves closed.
- 3. The power range high flux, low setpoint trip and the intermediate range high flux, high setpoint trip shall be unblocked when power
< 10% of rated power,
- 4. The source range high flux, high setpoint trip shall be unblocked when. the intermediate range nuclear flux is< 5 x 10-11 am~eres.
Basis The power range reactor trip low setpoint provides protection in the power
TS 2.3-5 2-1-72 range for a power excursion beginning from low power. This trip value was used in the safety analysis. (l) The intermediate range high flux, low setpoint and source range high flux, high setpoint trips provide additional protection against uncontrolled startup excursions. As power level: increases, during startup, these trips are blocked to prevent unnecessary plant trips. The high and low pressurizer pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip is also a backup to the pressurizer code safety valves for overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The low pressurizer pressure reactor trip also trips the reactor 3 in the unlikely ev~nt of a loss-of-coolant accident. ( ) e The overtemperature ~T reactor trip provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that the transient is slow with respect to piping* transit delays from the core to the temperature detectors (about 3 seconds), and pressure is within the range between high and low pressure reactor trips. With normal axial power distribution, the reactor trip limit, with allowance (2) for errors, is always below the core safety limit as shown on TS Figure 2.1-1.
*If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip limit is auto-(4) (5) matically reduced.
In order to operate with a reactor coolant loop out of service (two-loop operation) and with the stop valves of the inactive loop open, the overtemperature
~T trip setpoint calculation has to be modified by the adjustment of the variable I
TS 2.3-6 2-1-72 K
- This adjustment, based on limits for two loop operation, provides sufficient 1
margin to DNB for the aforementioned transients during two loop operation. The required adjustment and subsequent mandatory calibrations are made in the pro-tective system racks by qualified technicians* in the same manner as adjustments before initial startup and normal calibrations for three-loop operation. For two loop operation with the inactive loop stop valves closed, the overtemperature tT trip setpoints used for three loop operation are adequate to protect against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution provided only that the transient is slow with respect to transit delays from the core to the temperature detectors. The overpower tT reactor trip prevents power density anywhere in the core from exceeding 112% of design power density as discussed in Section 7 and specified in Section 14.2.2 of the FSAR and includes corrections for axial power dis-tribution, change in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. The specified setpoints meet this requirement and include allowance (2) for instrument errors.
- The low flow reactor trip protects the core against DNB in the event of a sudden loss of power to one or more reactor coolant pumps. The setpoint (6) specified is consistent with the value used in the accident analysis, The underfrequency reactor coolant pump trip protects against a decrease in flow caused by low electrical frequency. The specified setpoint assures a reactor trip signal before the low flow trip point is reached.
- As used here, a qualified technician means a technician who meets the re-quirements of ANS-3. He shall have a minimum of two years of working experience in his speciality and at least one year of related technical training.
TS 2.3-7 2-1-72 The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief, Approximately 1154 ft 3 of water corresponds to 7 92% of span. The specified setpoint allows margin for instrument error( ) and transient level overshoot beyond this trip setting so that the trip function prevents the water level from reaching the safety valves. The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents, The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to 7 allow for starting delays for the Auxiliary Feedwater System, ( ) The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal unit operations. The prescribed setpoint above which these trips are unblocked assures their availability in the power range where needed, Above 10% power, an automatic reactor trip will occur if two or more reactor coolant pumps are lost. Above 50% power during t~ree-loop operation, an automatic reactor trip will occur if any pump is lost or de-energized. This latter trip will prevent the minimum value of the DNBR from going below 1.30 during normal operational transients and anticipated transients when only two loops are in operation and the overtemperature ~T trip setpoint is adjusted to the valve specified for three loop operation. During two loop operation with the loop stop valves in the inactive loop open, and the over-temperature ~T trip setpoint is adjusted to the value specified for two loop operation, a reactor trip at 60% power will prevent the minimum value of DNBR from going below 1.30 during normal operational transients and anticipated transients when only two loops are in operation, During two loop operation with the inactive loop stop valves closed, a reactor trip at 65% power will
TS 2. 3-8 2-1-72 prevent the minimum DNBR from going below 1.30 during normal operational transients and anticipated transients. For this latter case the overtemperature
~T trip setpoints may remain at the values used for three loop operation, Although not necessary for core protection other reactor trips provide additional protection. The steam/feedwater flow mismatch in concidence with a low steam generator water level is designed for protection from a sudden loss of the reactor's heat sink. Upon the actuation of the safety injection circuitry, the reactor is tripped to decrease the severity of the accident condition. Upon turbine trip, at greater than 10% power, the reactor is tripped to reduce the severity of the ensuing transient.
e (1) FSAR Section 14.2.1 (2) FSAR Section 14.2 (3) FSAR Section 14.5 (4) FSAR Section 7.2 (5) FSAR Section 3.2.2 (6) FSAR Section 14.2.9 (7) FSAR Section 7.2
TS 3.1-1 2-1-72 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicability I I Applies to the operating status of the Reactor Coolant System. Objectives To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe reactor operation. These conditions relate to: operational components, heatup and cooldown, leakage, reactor coolant activity, oxygen and chloride concentrations, and minimum temperature for criticality. A. Operational Components Specifications
- 1. Reactor Coolant Pumps
- a. A reactor shall not be brought critical with less than two pumps, in non-isolated loops, in operation.
- b. If an unscheduled loss of one or more reactor coolant pumps occurs while operating below 10% rated power (P-7) and results in less than two pumps in service, the affected
TS 3.1-2 12-29-71 plant shall be shutdovm and the reactor made subcritical by inserting all control banks into the core. The shutdown rods may remain withdrawn.
- c. A minimum of one pump in a non-isolated loop, or one residual heat removal pump and its associated flow path, shall be in operation during reactor coolant boron con-centration reduction.
- d. Reactor power shall not exceed 50% of rated power with only two pumps in operation and the inactive loop stop valves open unless the overtemperature tT trip setpoint, Ki, for two loop operation, has been set at 0.94, after e which power shall not exceed 60%. For two loop operation with the inactive loop stop valves closed, K1 may remain at the value for three loop operation and power shall not exceed 65%.
2 ** Steam Generator A minimum of two steam generators in non-isolated loops shall be operable when the average reactor coolant temperature is great.er than 3SOOF.
- 3. Pressurizer Safety Valves
- a. One valve shall be operable whenever the head is on the reactor vessel, except during hydrostatic tests.
TS 3.1-3 2-1-72
- b. Three valves shall be operable when the reactor coolant
~ average temperature is greater than 350°F, the reactor is critical, or the Reactor Coolant System is not connected to the Residual Heat Removal System.
- c. Valve lift settings shall be maintained at 2485 psig + 1 percent.
- 4. Reactor Coolant Loops Loop stop valves shall not be closed in more than one loop unless the Reactor Coolant System is connected to the Residual Heat Removal System and the Residual Heat Removal System is operable.
- 5. Pressurizer The reactor shall be maintained subcritical by at least 1% until the steam bubble is established and necessary sprays and heaters are operable.
Basis Specification 3.1.A-1 requires that a sufficient number of reactor coolant pumps be operating to provide coastdown core cooling flow in the event of a loss of reactor coolant flow accident. This provided flow will maintain the DNBR above 1.30. (l) Heat transfer analyses also show that reactor heat equivalent to approximately 10% of rated power can be removed with natural circulation; however, the plant is not designed for critical operation with natural cir-culation or one loop operation and will not be operated with these conditions.
TS 3.1-4 2-1-72 When the boron concentration of the Reactor Coolant System is to be reduced the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uni-form concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the equivalent of the reactor coolant system volume in approximately one half hour. One steam generator capable of performing its heat transfer function will provide sufficient heat removal capability to remove core decay heat after a normal reactor shutdown. Because of the low-low steam generator water level reactor trip, normal reactor criticality cannot be achieved without water in - the steam generators in reactor coolant loops with open loop stop valves. The requirement for two operable steam generators, combined with the requirements of Specification 3.6, ensure adequate heat removal capabilities for reactor coolant system temperatures of greater than 350°F. Each of the pressurizer safety valves is designed to relieve 295,000 lbs. per hr. of saturated steam at the valve setpoint. Below 350°F and 450 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure. There are no credible accidents which could occur when the Reactor Coolant System is connected to the Residual Heat Removal System which could give a surge rate exceeding the capacity of one pressurizer safety valve. Also, two safety v:alves have a capacity greater than the maximum surge rate resulting from complete loss of load. (Z)
TS 3.1-5 5-31-71
- The limitation specified in item 4 above on reactor coolant loop isolation will prevent an accidental isolation of all the loops which would eliminate the capability of dissipating core decay heat when the Reactor Coolant System is not connected to the Residual Heat Removal System.
The requirement for steam bubble formation in the pressurizer when the reactor has passed 1% subcriticality will ensure that the Reactor Coolant System will not be solid when criticality is achieved.
References:
- (1)
(2) FSAR Section 14.2.9 FSAR Section 14.2.10
TS 3.1-6 2-1-72 B. HEATUP AND COOLDOWN Specification
- 1. Unit 1 reactor coolant temperature and pressure and the system heatup and cooldown (with the exception of the pressurizer) shall be limited in accordance with TS Figure 3.1-1.
Heatup: The 0°F/hr, curve of Figure 3.1-1 may be used for heatup rates of up to 60°F/hr. below an indicated temperature of 257°F and 100°F/hr. above 257°F. e Cooldown: Allowable combinations of pressure and temperature for a specific cooldown rate are below and to the right of the limit lines for that rate as shown in TS Figure 3.1-1. This rate shall not exceed 50°F/hr. for temperatures at or below an indicated temperature of 263°F. For temperatures above an indicated temperature of 263°F, the rate shall not exceed 100°F/hr. The limit lines for rates between those shown in TS Figure 3.1-1 may be obtained by interpolation.
- 2. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 70°F.
TS 3.1-7 2-1-72
- 3. The pressurizer heatup and cooldown rates shall not exceed 200°F/hr. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320°F.
- 4. TS Figure 3.1-1 shall be updated periodically in accordance with the following procedures, before the calculated maximum exposure of the vessel exceeds the exposure for which TS Figure 3.1-1 applies.
- a. The curve based on 0.25% Cu, in TS Figure 3.1-2.shall be used to predict the increase in transition temperature based on integrated power unless measurements on the most recently examined irradiation specimens show that this is not appropriate, In this case,a new curve having the same slope as the original shall be constructed such that it is above all the applicable data points.
- b. At or before the end. of the integrated power period for which TS Figure 3.1-1 applies, the limit lines on the figure shall be updated for a new integrated power period as follows. The total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated neutron exposure. The predicted increase in transition temperature at the end of the new period shall then be obtained from TS Figure 3.1-2 as revised by TS 3.l.B,4a above.
TS 3.1-8 2-1-72
- c. The limit lines in TS Figure 3.1-1 shall be moved parallel to the temperature axis (horizontally) in the direction of increasing temperature a distance equivalent to the transition temperature increase obtained from TS Figure 3.1-2 as revised less the increment used for the end of the present period.
Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor system temperature and pressure (1) changes. These cyclic loads are introduced by normal unit load transients, reactor trips, and startup and shutdown operation. The number of thermal and loading cycles used for design purposes are shown in Section 4.1 of the FSAR. During unit startup and shutdown, the rates of temperature and pressure are limited. The maximum plant heatup and cooldown rate of 100°F/hr. is consistent with the design number of cycles and satisfies stress limits for (2) cyclic operation. The allowable pressure vs. temperature is based on a temperature scale relative to the RT The RT is basically the drop weight NDTT of the material, NDT NDT as determined by ASTM E208. However, to assure that this value is conservative, and to guard against the possibility that material with low upper shelf toughness, or with a low rate of increase of toughness with temperature, is not properly evaluated, Charpy tests are also performed. If 35 mils lateral
TS 3.1-9 2-1-72 expansion or 50-ft-lbs is not obtained at NDTT + 60, the RTNDT is shifted upward until this criterion is met. This procedure of selecting RT assures that the K curve used to NDT IR calculate allowable pressures will be conservatively applicable to the material. The procedure for determining the limiting RTNDT for the Reactor System is as follows:
- 1. Determine the highest RT of the material in the core region of the NDT reactor vessel, using original values and adding to this the predicted shift in RTNDT due to radiation during the service period for which this RTNDT applies. This takes into account the copper content of the material.
- 2. Examine the data for all other ferritic materials in the reactor system to assure that the RTNDT so selected is the highest in the system. If drop weight data are not available for all materials, the RTNDT of these shall be estimated in a conservative manner using trend data for the materials concerned.
- 3. For succeeding service periods, the same procedure as given in (1) above will be used unless test data from the surveillance program indicates that this will not be appropriate. In this event, the results of these tests will be used to predict the limiting RT NDT Test results on mi>t-.erial from the Surry Unit 1
- reactor vessel is presented in FSAR Table 4.A-1. Using the above procedure, the highest original RT of the core region plates is +20°F. No drop weight NDTT value is NDT
TS 3.1-10 2-1...:72 available for the core region weld material but on the basis of actual drop weight data on many similar weld materials, plus the actual Charpy values on this material, the drop weight NDTT is estimated to be 0°F. The RT for the first two years of operation will include a conservative NDT estimate of the shift in RT caused by radiation. In line with NB 23ll(d) NDT and Appendix G2000 of Section III of the ASME I fV & Boiler Code, properties at l/4T are considered. The maximum fluence after 2 years of full power operation at 1/4 Twill be 1.1 x 1018
- Shifts of the core region plates and welds considering their respective copper contents (TS Figure 3.1-2) are:
plates, 0.11 Cu,~ RTNDT = 50°F welds, 0.25 Cu,~ RTNDT = 90°F The limiting RTNDT after 2 years of full operation will be highest predicted value. For the plates, this will be +20°F + 50°F = 70°F, and for the welds, 0°F + 90°F = 90°F, so (including additional conservatism)= RTNDT of 100°F will be assumed for this initial period. In examining the data for the rest of the material in the vessel; as well as the properties for the other ferritic components of the reactor system, it is certain that all other materials will have RTNDT values significantly lower than 100°F, Since the neutron spectra at the samples and vessel inside radius are identical, the measured (RT)NDT shift for a sample can be supplied with confidence to the adjacent section of reactor vessel for some later stage
TS 3.1-11 2-:-1-72 in plant life. The maximum exposure of the vessel is obtainable from the measured sample data by appropriate application of the calculated azimuthal neutron flux variation. During cooldown, the thermal stress varies from tensile at the inner wall to compressive at the outer wall. The internal pressure superimposes a tensile stress on this thermal stress pattern, increasing the stress at the inside wall and relieving the stress at the outside wall. Therefore, the limiting stress always appears at the inside wall and the limit line has a direct dependence on cooldown rate. This leads to a family of curves for cooldown, as shown in TS Figure 3.1-1. For heatup, the thermal stress is reversed and the location of the limiting stress is a function of heatup rate. The limit lines no longer bear the simple relationship to heatup as they do to cooldown rate. The 0°F/Hr cooldown line on TS Figure 3.1-1 bounds all limit lines for heatup rates up to 60°F/Hr for indicated temperatures at or below 257°F, and 100°F/Hr above 257°F. TS Figure 3.1-1 defines stress limitations on~y. For normal operation other inherent plant characteristics~ e.g., pump parameter and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure ranges. The heatup and cooldown rate of 100°F/Hr for the steam generator is consistent with the remainder of the Reactor Coolant System, as discussed in the first paragraph of the basis. The stresses are within acceptable limits for the anticipated usage. Temperature requirements for the steam generator correspond with the measured NDT for the shell. The spray should not be used if the temperature difference
TS 3.1-12 2-1-72 between the pressurizer and spray fluid is greater than 320°F, This limit is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit,
References:
(1) FSAR, Section 4.1.5 (2) ASME Boiler & Pressure Vessel Code, Section III, N-415 e _J
TS 3.1-13 2-1-72
- c. Leakage Specifications
- 1. Detected or suspected leakage from the Reactor Coolant System shall be investigated and evaluated.
- 2. If the leakage rate, from other than controlled leakage sources, such as the Reactor Coolant Pump Controlled Leakage Seals, exceeds 1 gpm and the source of the leakage is not identified within four hours of leak detection, the reactor shall be brought to hot shutdown. If the source of leakage is not identified within an additional 48 hours, the reactor shall be brought to a cold shutdown condition.
- 3. If the sources of leakage are identified and the results of the evaluations are that continued operation is safe, operation of the reactor with a total.leakage, other than leakage from controlled sources, not exceeding 10 gpm shall be permitted except as specified in C.4 below.
- 4. If it is determined that leakage exists through a non-isolable fault which has developed in a Reactor Coolant System component body, pipe well, vessel wall, or pipe weld, the reactor shall be brought to a cold shutdown condition and corrective action taken prior to resumption of unit operation,
- 5. If the total leakage ,other t.han leakage from controlled sources, exceeds 10 gpm the reactor shall be placed in the cold shutdown condition.
TS 3.1-14 e 12-29-71 Basis Leakag~ from the Reactor Coolant System is collected in the containment or by other systems. These systems are the Main Steam System, Condensate and Feedwater System, the Gaseous and Liquid Waste Disposal Systems, the Component Cooling System, and the Chemical and Volume Control System. Detection of leaks from the Reactor Coolant System is by one or more of the following:
- 1. An increased amount of makeup water required to maintain normal level in the pressurizer.
- 2. A high temperature alarm in the leakoff piping provided to collect reactor head flange leakage.
- 3. Containment sump water level indication.
- 4. Containment pressure, temperature, and humidity indication.
If there is significant radioactive contamination of the reactor coolant, the radiation monitoring system provides a sensitive indication of primary
TS 3.1-15 6-30-71 system leakage. Radiation monitors which indicate primary system leakage include the containment air particulate and gas monitors, the condenser air ejector monitor, the component cooling water monitor, and the steam generator blowdown monitor. Ro.derences FSAR, Section 4.2.7 - Reactor Coolant System Leakage fSAR, Section 14.3.2 - Rupture of a Main Steam Pipe D. Maximum Reactor Coolant Activity ~1cc i._f ications The total specific activity of the reacto-.:- coolant due to nuclides with ltalf-lives of more than JO minutes shall not exceed (41 /E)µ Ci/cc whenever the reactor is critical or the average temperature is greater than 500°F, where Eis the average sum of the beta and gamma energies, in Mev, per disintegration, If this limit is not satisfied, the reactor shall be shut down and cooled to S00°F or less within 6 hours after detection. Should this limit be exceeded by 25%, the reactor shall be made subcritical and cooled to 500°F or less within 2 hours after detection.
TS 3.1-16 5-31-71
- Basis The specified limit provides protection to the public against the potential release of reactor coolant activity to the atmosphere, as demonstrated by th~ following analysis of a steam generator tube rupture accident.
Rupture of a steam generator tube would allow reactor coolant activity to enter the secondary system. The major portion of this activity is noble gases which would be vented to the containment on high radiation signal at the air ejector. Activity would continue to be released to the containment until the operator could reduce the primary system pressure below the set-point of the secondary relief valves and could isolate the faulty steam
- generator .
The worst credible set of circumstances is considered to be a double-ended break of a single tube, and this has been analyzed on the bases of isolation of the faulty steam generator by the operator within 30 min. after the event, see FSAR Section (14. 3 .1) . The reactor coolant was assumed to be at an equilibrium activity resulting from 1% failed fuel. The airborne release of the activity from the secondary system was assumed to be all of the noble gases that enter the secondary system and a 1.0 x 10-2 fraction of the halogens. Using a xlQ of 8.14 x 10- 4 sec/m 3 , the doses from immersion in a pulse duration cloud at the exclusion
TS 3.1-17 6-30-71 boundary would be 0.30 Rem whole body and 0.28 Rem thyroid. Thus, these doses are well below the guidelines suggested in 10CFRlOO. The basis for the 500°F temperature contained in the Specification is that the saturation pressure corresponding to 500°F, 680.8 psia, is well below the pressure at which the atmospheric relief valves on the secondary side could be actuated. Measurement of E will be performed at least twice annually. Calculations required to determine E will consist of the following:
- 1. Quantitative measurement, in units of µCi/cc, of the significant radionuclides with half lives greater than 30 minutes. These nuclides make up at least 95% of the total activity in the reactor coolant.
Table 9.1-5 of the FSAR lists significant radionuclides expected with 1% failed fuel.
- 2. A determination of the beta and gamma decay energy per disintegration of each nuclide determined in (1) above by applying known decay energies and schemes.
3. A calculation of Eby appropriate weighing of each nuclide's beta and gamma energy with its concentration as determined in (1) above. E. Minimum Temperature For Criticality Specifications
TS 3.1-18 6-30-71
- 1. Except during low power physics tests, the reactor shall not be made critical at any temperature above which the moderator temperature coefficient is positive.
- 2. In no case shall the reactor be made critical with the reactor coolant temperature below DTT + 10°F, where the value of DTT + 10°F is as determined in Part B of this specification.
- 3. When the reactor coolant temperature is below the minimum temperature as specified in E-1 above, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to primary coolant depressurization.
Basis During the early part of the initial fuel cycle, the moderator temperature coefficient is calculated to be slightly positive at coolant temperatures below (1) (2) the power operating range. The moderator coefficient at low temperatures will be most positive at the beginning of life of the initial fuel cycle, when the boron concentration in the coolant is the greatest. Later in the initial cycle and during subsequent reload fuel cycles, the boron concentrations in the coolant will be lower and the moderator coefficients will be either less positive or will be negative. At all times, the moderator coefficient is negative in the
- operating
- 2 power range. (l) ( ) The maximum
- temperature at wh"ich t h e mo d era t or coefficient is positive at the beginning of life of the initial fuel cycle with all control rod assemblies withdrawn, is determined during pre-operational physics tests. When control rod assemblies are inserted, the temperature at which the
TS 3.1-19 6-30-71 moderator coefficient becomes negative is lower so that at the temperature determined during the physics tests and with the operational control rod program, the temperature coefficient is expected to be negative. The requirement that the reactor is not to be made critical when the moderator coefficient is positive has been imposed to prevent any unexpected power ex-cursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant pressure. This requirement is waive'd during low power physics tests to permit measurement of reactor moderator coefficient and other physic~ design parameters of interest. During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler 2 3 coefficient ( ) ( ) and the small integrated 6k/k would limit the magnitude of a power excursion resulting from a reduction of moderator density. The requirement that the reactor is not to be made critical with a reactor coolant temperature below DTT + 10°F provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility transition temperature range. Heatup to this temperature is accomplished by operating the reactor coolant pumps. If a specified shutdown reactivity margin is maintained (TS Section 3.12), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure. (1) FSAR Figure 3.3-8 (2) FSAR Table 3. 3-1 (3) FSAR Figure 3.3-9
TS 3.1-20 2-1-72 F. Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration Specification 1, Concentrations of contaminants in the reactor shall not exceed any one of the following limits when the reactor coolant is above 250°F. Normal Steady-State Transients not to Exceed Contaminant Operation * (PPM) 24 Hours (PPM) a, Oxygen 0.10 1.00
- b. Chloride 0.15 1.50
- c. Fluoride 0.15 1.50
- 2. If any one of the normal steady-state operating limits as specified in 3.1.F.l above are exceeded, or if it is anticipated that they may be exceeded, corrective action shall be taken immediately.
- 3. If the concentrations of any one of the contaminants can not be controlled within the limits of Specification 3.1.F.l above, the reactor shall be brought to the cold shutdown condition, utilizing normal operating procedures, and the cause of the out-of-specification operation ascertained and corrected. The reactor may then be restarted and operation resumed if
_the maximum concentration of any of the contaminants did not exceed the permitted transient values. Otherwise, a safety review is required before startup.
TS 3.1-21 2-1-72
- 4. Concentrations of contaminants in the reactor coolant shail not exceed the following maximum limits when *the reactor coolant temperature is below 250°F:
Normal Concentration Transient not to Contaminant (PPM) exceed 24"hours (PPM)
- a. Oxygen Saturated Saturated
- b. Chloride 0.15 1. 5
- c. Fluoride 0.15 1.5 If the limits above are exceeded, the reactor shall be immediately brought to the cold shutdown condition and the cause of the out-of-specification condition shall be ascertained and corrected.
- 5. For the purposes of correcting the contaminant concentrations to meet technical specifications 3.1.F.l and 3.1.F.4 above, increase in coolant temperature consistant with operation of primary coolant pumps for a short period of time to assure mixing of the coolant shall be permitted. This increase in temperature to assure mixing shall in no case cause the coolant temperature to exceed 250°F.
- 6. If more than one contaminant or contaminants transient, which results in contaminant levels exceeding any of the normal steady state operation limits specified in 3.1.F.l or 3.1.F.4, is experienced in any seven consecutive day period, the reactor shall be placed in a cold shutdown condition until the cause of the out-of-specification operation is as-certained and corrected.
TS 3.1-22 2-1-72 Basis: By maintaining the oxygen, chloride and fluoride concentrations in the reactor coolant below the limits as specified in technical specification 3.1.F.1 and 3.1.F.4 the integrity of the reactor coolant system is assured under all operating (1) conditions. If these limits are exceeded, measures can be taken to correct the condition, e.g., replacement of ion exchange resin, or adjustment of the 2 hydrogen concentration in the volume control tank. ( ) Because of the time dependent nature of any adverse effects arising from oxygen, chloride, and fluoride concentration in excess of the limits, it is not necessary to shutdown immediately if the condition can be corrected. Thus the period of 24 hours for corrective action to restore concentrations within the limits has been established. If the corrective action has not been effective at the end of the 24 hour period, then the reactor will be brought to the cold shutdown condition and the corrective action will continue. In restoring the contaminant concentrations to within specification limits in the event such limits were'exceeded, mixing of the primary coolant with the reactor coolant pumps may be required. This will result in a small heatup of short duration which will not increase the average coolant temperature above 250°F. More than one contaminant transient, in any seven consecutive day period, that results in exceeding normal steady state operation limits, could be indicative of unforeseen chemistry control problems. Such potential problems warrant in-vestigation, correction and measures to insure that the integrity of the reactor coolant system is maintained.
TS 3.1-23 2-1-72 References (1) FSAR Section 4.2 (2) FSAR Section 9.2
UPPER PRESSURIZATION LIMITS FOR HEATUP AND COOLDOWN SURRY UNIT ONE 24-00 2200 2000 - 1800
; t*"
NOTE: C l) 0.. U.J 0::: 1600 Cl) Cl) U.J 0::: 14-00 0... I-z: <( 1200 -l 0 +/-+/-:+ ****rr 0 .+* *.+ (.) 1000 0 U.J I- <( 800 _..+* (.) ~-* .*++ 0 .+-* z: 600 0°F/HR . ~~-'
++, +-**' }fl 4-00 20°F/HR .... ~ '!
~i""**.
+!
*-t* .*,. IE_ t~.*. . ,. -*** . .,. ,. +'"i. - *H
- n::1 r+*, , **:.,_ -i-t-
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T ;}1! 'l ' fl @Rf, + '. +-i+ ** H 200 50°F /HR: . RTotmT (RT}HDT . i
!*-t*"
'_:+ +t -~
.+~* ... , _.,
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0 60 -20 0 20 4-0 60 80 100 120 14-0 160 180 200 220 24-0 260 280 300 320 INDICATED COOLANT TEMPERATURE {°F) t-3 Cl) Figure 3,1-1 Upper pressurization limit for heatup is depicted by "o°F/Hour" curve. Upper pressurization limits for cooldown are dependent as depicted, on the cooldown N lu.> rates. This figure is applicable through an nvt of r-'
- I r-'
1.1 x 1018 neutrons per cm2 (E> 1 Mev) ...._. I N r-'
TS Figure 3al*2 2-1-72 10 3 8 I-C z: I- 102 C:: IMev) Figure 3. 1-2. Radiation Induced Increase In Transition Temperature
TS 3.2-1 2-1-72 - 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Applicability Applies to the operational status of the Chemical and Volume Control System. Objective To define those conditions of the Chemical and Volume Control System necessary to ensure safe reactor operation. Specification - A. When fuel is in a reactor there shall be at least one flow path to the core for boric acid injection. The minimum capability for boric acid injection shall be equivalent to that supplied from the refueling water storage tank. B, For one unit operation the reactor shall not be critical unless the following Chemical and Volume Control System conditions are met:
- 1. Two charging pumps shall be operable.
- 2. Two boric acid transfer pumps shall be operable.
- 3. The boric acid tanks (tank associated with the unit plus the common tank) together shall contain a minimum of 4200 gallons of at least 11.5% (but not greater than 13%) by weight boric acid solution at a temperature of at least 145°F,
TS 3.2-2 2-1-72
- 4. System piping and valves shall be operable to the extent of establishing two flow paths to the core; one flow path from the boric acid tanks to the charging pumps and a flow path from the refueling water storage tank to the charging pumps.
- 5. Two channels of heat tracing shall be operable for the flow paths requiring heat tracing.
- 6. Recirculation between a unit's Boron Injection Tank and the Boric Acid Tank(s) assigned to the unit shall be maintained.
C. For two unit operation the reactor shall not be critical unless the following Chemical and Volume Control System conditions are met:
- 1. Two charging pumps shall be operable per unit.
- 2. Three boric acid transfer pumps shall be operable.
- 3. When the common tank is in service, it shall be assigned to only one unit at a time. For that unit which has usage of the common tank, the boric acid tanks (unit's tank plus common tank) together shall contain a minimum of 4200 gallons of at least 11. 5% (but not greater than 13%) by weight boric acid solution at a temperature of at least 145°F.
For that unit which does not have usage of the common tank, the unit's own tank shall contain a minimum of 4200 gallons of at least 11.5% (but not greater than 13%) by weight boric acid solution at a temperature of at least 145°F When the common tank is assigned to one unit, valves shall be positioned to establish a flow path to that unit and prevent flow to the other unit.
TS 3.2-3 12-29-71
- 4. System piping and valves shall be operable to the extent of establishing two flow paths to the core; one flow path from the boric acid tanks to the charging pumps and a flow path from the refueling water storage tank to the charging pumps.
- 5. Two channels of heat tracing shall be operable for the flow paths requiring heat tracing.
- 6. Recirculation between a unit's Boron Injection Tank and the Boric Acid Tank(s) assigned to the unit shall be maintained.
D. The requirements of Specifications Band C above may be modified to allow one of the following components to be inoperable at any one time. If the system is not restored within the time period specified, the e reactor shall be placed in the hot shutdown condition. If the requirements of Specification 3.2.B and Care not satisfied within an additional 48 hours, the reactor shall be placed in the cold shutdown condition.
- 1. One of the stipulated boric acid transfer pumps may be inoperable for a period not to exceed 24 hours provided immediate attention is directed to making repairs.
- 2. Two charging pumps may be inoperable subject to the provisions of Specification 3.3-B.
- 3. One heat tracing circuit may be inoperable for a period not to exceed 24 hours provided immediate attention is directed to making repairs.
TS 3.2-4 2-1-72 Basis The Chemical and Volume Control System provides control of the Reactor Coolant System Boron inventory. This is normally accomplished by using boric acid transfer pumps which discharge to the suction of each unit's charging pumps. The Chemical and Volume Control System contains four boric acid transfer pumps. Two of these pumps are normally assigned to each unit but valving and piping arrangements allow pumps to be shared such that 3 out of 4 pumps can service either unit. An alternate (not normally used) method of boration is to use the charging pumps taking suction directly from the refueling water storage tank. There are two sources of borated water available to the suction of the charging pumps through two different paths, one from the refueling water storage tank and one from the discharge of the boric acid transfer pumps. A. The boric acid transfer pumps can deliver the boric acid tank contents (11.5% solution of boric acid) to the charging pumps. B. The charging pumps can take suction from the volume control tank, the boric acid transfer pumps and the refueling water storage tank. Reference is made to Technical Specification 3.3. The quantity of boric acid in storage from either the boric acid tanks or the refueling water storage tank is sufficient to.borate the reactor coolant in order to reach cold shutdown at any time during core life. Approximately '4200 gallons of the 11. 5% solution of boric acid are required to meet cold shutdown conditions. Thus, a minimum of 4200 gallons in the boric acid tank is specified. An upper concentration limit of 13% boric acid in the
TS 3.2-5 12-29-71 tank is specified to maintain solution solubility at the specified low temperature limit of 145°F. For redundancy, two channels of heat tracing are installed on lines normally containing concentrated boric acid solution. Continuous recirculation between the Boron Injection Tank and the Boric Acid Tank(s) ensures that a unit's Boron Injection Tank is full of concentrated boric acid at all times. The Boric Acid Tank(s), which are located above the Boron Injection Tank(s), are supplied with level alarms, which would annunciate if a leak in the system occurred. References FSAR Section 9.1 Chemical and Volume Control System
TS 3.3-1 2-1-72 3.3 SAFETY INJECTION SYSTEM Applicability Applies to the operating status of the Safety Injection System. Objective To define those limiting conditions for operation that are necessary to provide sufficient borated cooling water to remove decay heat from the core in emergency situations. Spec if ica tions A. A reactor shall not be made c~itical unless the following conditions are met: 1, The refueling water tank contains not less than 350,000 gal. of borated water with a boron concentration of at least 2000 ppm.
- 2. Each accumulator is pressurized to at least 600 psig and contains 3 3 a minimum of 934 ft and a maximum of 939 ft of borated water with a boron concentration of at least 1950 ppm.
- 3. The boron injection tank and isolated portions of the inlet and outlet piping contains no less than 900 gallons of water with a boron con-centration equivalent to at least 11.5% to 13% weight boric acid solution
TS 3. 3-:2 I 12-29-71 at a temperature of at least 145°F.
- 4. Two channels of heat tracing shall be available for the flow paths.
- 5. Two charging pumps are operable.
- 6. Two low head safety injection pumps are operable.
- 7. All valves, piping, and interlocks associated with the above components which are required to operate under accident conditions are operable.
- 8. The Charging Pump Cooling Water Subsystem shall be operating as follows:
a) Make-up water from the Component Cooling Water Subsystem shall be available. b) Two charging pump component cooling water pumps and two charging pump service water pumps shall be operable. c) Two charging pump intermediate seal coolers shall be operable. d) The accumulator discharge valves (MOV 865A, B, & C) in non-isolated loops shall be blocked open by de-energizing the valve motor operator when the reactor coolant pressure is greater than 1000 psig.
TS 3 3-3 2- 1-72 B. The requirements of Specification 3.3-2 may be modified to"""' allow one of the following components to be inoperable at any one time. If the system is not restored to meet the requirements of Specification 3.3-A within the time period specified, the reactor shall initially be placed in the hot shutdown condition. If the requirements of Specification 3.3-A are not satisfied within an additional 48 hours the reactor shall be placed in the cold shutdown condition.
- 1. One accumulator may be isolated for a period not to exceed 4 hours.
- 2. Two charging pumps per unit may be out of service, provided immediate attention is directed to making repairs and one pump is restored to operable status within 24 hours. If one pump unit is out of service, the standby pump shall be tested before initiating maintenance and once every 8 hours to assure operability.
- 3. One low head safety injection pump per unit may be out of service, provided immediate attention is directed to making repairs and the pump is restored to operable status within 24 hours. The other low head safety injection pump shall be tested to demonstrate operability prior to initiating repair of the inoperable pump and shall be tested once every eight (8) hours thereafter, until both pumps are in an operable status or the reactor is shut down.
- 4. Any one valve in the Safety Injection Syst~m may be inoperable provided repairs are initiated immediately and are completed within
- 24 hours. Prior to initiating repairs, all automatic valves in the redundant system shall be tested to demonstrate operability.
TS 3.3-4 12-29-71
- 5. One channel of heat tracing may be inoperable for a period not to exceed 24 hours, provided immediate attention is directed to making repairs.
- 6. One charging pump component cooling water pump or one charging pump service water pump may be out of service provided the pump is restored to operable status within 24 hours.
- 7. One charging pump intermediate seal cooler or other passive component may be out of service provided the system may still operate at 100 percent capacity and repairs are completed within 48 hours.
Basis The normal procedure for starting the reactor is, first, to heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing control rods and/or diluting boron in the coolant. With this mode of startup the Safety Injection System i~ required to be operable as specified. During low power physics tests there is a negligible amount of energy stored in the system; therefore an accident comparable in severity to the Design Basis Accident is not possible, and the full capacity of the Safety Injection System is not required. The operable status of the various systems and components is to be demonstrated by periodic tests, detailed in TS Section 4.1. A large fraction of these tests are performed while the reactor is operating in the power range. If a component e is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short-time. A
TS 3.3-5 12-29-71 single component being inoperable does not negate the ability of the system to perform its function, but it reduces the redundancy provided in the reactor design and thereby limits the ability to tolerate additional equipment failures. To provide maximum assurance that the redundant component(s) will operate if required to do so, the redundant component(s) are to be tested prior to initiating repair of the inoperable component and, in some cases are to be retested at intervals during the repair period. In some cases, i.e.chargirig pumps, additional components are installed to allow a component to be inoperable without affecting system redundancy. For those cases which are not so designed, if it develops that (a) the inoperable component is not repaired within the specified allowable time period, or (b) a second component in the same or related system is found to be inoperable, the reactor will initially be put in the hot shutdown condition to provide for reduction of the decay heat from the e fuel, and consequent reduction of cooling requirements after a postulated loss-of-coolant accident. After 48 hours in the hot shutdown condition, if the malfunction(s) are not corrected the reactor will be placed on the cold shutdown condition, following normal shutdown and cooldown procedures. The Specification requires prompt action to effect repairs of an inoperable component, and therefore in most cases repairs will be completed in less than the specified allowable repair times. Furthermore, the specified repair times do not apply to regularly scheduled maintenance of the Safety Injection System, which is normally to be performed during refueling shutdowns. The limiting times for repair are based on: estimates of the time required to diagnose and correct various postulated malfunctions using safe arid proper procedures, the availability of tools, materials and equipment; health physics requirements and the extent to which other systems provide functional redundancy to the system under repair.
TS 3.3-6 12-29-71 Assuming the reactor has been operating at full rated power for at least 100 days, the magnitude of the decay heat production decreases as follows after initiating hot shutdown. Time After Shutdown Decay Heat,% of Rated Power 1 min. 3.7 30 min. 1.6 1 I hour 1.3 8 hours -o.1s 48 hours 0.48 Thus, the requirement for cor.e cooling in case of a postulated loss-of-coolant accident while in the hot shutdown condition is redµced by orders of magnitude below the requirements for handling a postulated loss-of-coolant accident occuring during power operation. Placing and maintaining the reactor in the hot shutdown condition significantly reduces the potential consequences of a loss-of-coolant accident, allows access to some of the Safety Injection System components in order to effect repairs, and minimizes the exposure to thermal cycling. Failure to complete repairs within 48 hours of going to hot shutdown condition f. is considered indicative of unforeseen problems, i.e., possibly the need of major maintenance. In such a case the reactor is to be put into the cold shutdown condition.
TS 3.3-7 2-1-72 The accumulators (one for each loop) discharge into the cold legs of the reactor coolant piping when Reactor Coolant System pressure decreases below accumulator pressure, thus assuring rapid core cooling for large breaks. The line from each accumulator is provided with a motorized valve to isolate the accumulator during reactor start-up and shutdown to preclude the discharge of the contents of the accumulator when not required. These valves receive a signal to open when safety injection is initiated. The AEC requires that these valves receive a signal to open when the Reactor Coolant Pressure exceeds a preselected value. Such a feature to satisfy this requirement will be installed prior to startup after the first refueling. In the interim period the valves will be blocked open by de-energizing the valve motor operators when the reactor coolant pressure exceeds 1000 psig. The operating pressure of the Reactor Coolant System is 2235 psig. and safety injection is initiated when this pressure drops to 650 psig. De-energizing the motor operator when the pressure exceeds 1000 psig allows sufficient time during normal start-up operation to perform the actions required to de-energize the valve. This procedure will assure that there is an operable flow path from each accumulator to the Reactor Coolant System during power operation and that safety injection can be accomplished.
TS 3.3-8 12-29-71 References FSAR Section 9.1 Chemical and Volume Control Sys.tern FSAR Section 6.2 Safety Injection System FSAR Technical Specifications Section 4.1 FSAR Supplement Addendum S6.25
TS 3.4-1 5-1-71
- 3.4 SPRAY SYSTEMS Applicability Applies to the operational status of the Spray Systems.
Objective To define those conditions of the Spray Systems necessary to assure safe unit operation. Specification
- A. A unit's Reactor Coolant System temperature or pressure shall not be made to exceed 350°F or 450 psig, respectively, or the reactor shall not be made critical unless the following Spray System conditions in that unit are met:
- 1. Two Containment Spray Subsys.tems, including containment spray pumps, piping, and valves shall be operable.
- 2. Four Recirculation Spray Subsystems, including recirculation spray pumps, coolers, piping, and valves shall be operable.
- 3. The refueling water storage tank shall contain not less than 350,000 gal of borated water at a maximum temperature as shown in TS Fig. 3.8-1.
TS 3.4-2 12-29-71 If this volume of water cannot be maintained by makeup, or the temperature maintained below that specified in *TS Fig. 3.8-1, the reactor shall be shutdown until repairs can be made. The water shall be borated to a boron concentration not less than 2,000 ppm which will assure that the reactor is in the refueling shutdown condition when all control rod assemblies are inserted.
- 4. The refueling water chemical addition ., tank shall contain not less than 3,360 gal of solution with a sodium hydroxide con-centration of not less than 18 percent by weight.
- 5. All valves, piping, and interlocks associated with the above components which are required to operate under accident conditions shall be operable.
B. During power operation the requirements of specification 3.4-A may be modified to allow the following components to be inoperable. If the com-ponents are not restored to meet the requirement of Specification 3.4-A within the time period specified below, the reactor shall be placed in the hot £hutdown condition. If the requirements of Specification 3.4-A are riot satisfied within an additional 48 hours the reactor shall be placed in the cold shutdown condition using normal operating procedures.
- 1. One Containment
,, Spray Subsystem may be out of service, provid_ed.
immediate attention is directed to making repairs and the subsystem can be restored to operable status within 24 hours. The other Con-
- tainment Spray Subsystem shall be test.ed as specified in Specification 4.5-A to demonstrate operability prior to initiating re.pair of the inoperable system.
TS 3.4-3 5-1-71
- 2. One containment spray pump drive, either motor or turbine, may 1 be out of service, provided immediate attention is directed to making repairs and the drive can be restored to operable status within 72 hours. If the inoperable drive is uncoupled from its pump, returning the effected spray pump and its remaining drive to an operable status within the time requirement of Specification 3.4.B-l, shall satisfy the requirements of Specification 3.4.A-l.
- 3. One outside Recirculation Spray Subsystem may be out of service provided immediate attention is directed to making repairs and the subsystem can be restored to operable status within 24 hours.
The other Recirculation Spray subsystems shall be tested as
- specified in Specification* 4.5-A to demonstrate operability prior to initiating repair of the inoperable system.
- 4. One inside Recirculation Spray Subsystem may be out of service provided immediate attention is directed to making repairs and the subsystem can be restored to operable status within 72 hours.
The other Recirculation Spray subsystems shall be tested as specified in Specification 4. 5-A to demonstrate operability prior to initiating repair of the inoperable subsystems. C. Should the refueling water storage tank temperature fail to be maintained
*at or below 45°F, the containment pressure and temperature shall be maintained in accordance with TS Fig. 3.8-1 to maintain the capability of
- the Spray System with the higher refueling water temperature. If the containment temperature and pressure cannot be maintained within the limits of TS Fig. 3.8-1, the reactor shall be placed in the cold shutdown condition.*
'IP
TS 3.4-4 5-1-71
- Basis The Spray Systems in each reactor unit consist of two separate parallel Contain-ment Spray Subsystems, c~ch of 100 percent capacity, and four separate parallel Recirculation Spray Subsystems, each of 50 percent capacity.
Each Containment Spray Subsystem draws water independently from the 350,000 gal capacity refueling water storage tank. The water in the tank is cooled to 45°F or below by circulating the tank water through one of the two refueling water storage tank coolers through the use of one of the two refueling water recirculation pumps. The water temperature is maintained by two mechanical refrigeration units as required. In each Containment Spray Subsystem, the water flows from the
- tank through a dual drive (motor and turbine) containment spray pump and is sprayed into the containment atmosphere through two separate sets of spray nozzles. The capability of the Spray Systems to depressurize the containment in the event of a Design Basis Accident is a function of the pressure and temperature of the containment atmosphere, the service water temperature, and the temperature in the refueling water storage tank as discussed in Specification 3.8-B.
Each Recirculation Spray Subsystem draws water from the common containment sump. In each subsystem the water flows through a recirculation spray pump and recirculation spray cooler, and is sprayed into the containment atmosphere through a separate set of spray nozzles. Two of the recirculation spray pumps are located inside the containment and two outside the containment in the
- containment auxiliary structure.
TS 3.4-5 5-1-71
- i With one Containment Spray Subsystem and two Recirculation Spray Subsystems operating together, the Spray Systems are capable of cooling and depressurizing the containment to subatmospheric pressure in less than 30 minutes following the Design Basis Accident. The Recirculation Spray Subsystems are capable of maintaining subatmospheric pressure in the containment indefinitely following the Design Basis Accident when used in conjunction with the Containment Vacuum System to remove any long term air inleakage.
In addition to supplying water to the Containment Spray System, the refueling water storage tank is also a source of water for safety injection following an accident. This water is borated to a concentration which assures reactor shutdown by approximately 10 percent ~k/k when all control rod assemblies are
- inserted and when the reactor is cooled.down for refueling .
References FSAR Section 4 Reactor Coolant System FSAR Section 6.3.l Containment Spray Subsystem FSAR Section 6.3.1 Recirculation Spray Pumps and Coolers FSAR Section 6.3.1 Refueling Water Chemical Addition Tank FSAR Section 6.3.1 Refueling Water Storage Tank FSAR Section 14.5.2 Design Basis Accident FSAR Section 14.5.5 Containment Transient Analysis
TS 3.5-1 2-1-72 3.5 RESIDUAL HEAT REMOVAL SYSTEM Applicability Applies to the operational status of the Residual Heat Removal System. Objective To define the limiting conditions for operation that are necessary to remove decay heat from the Reactor Coolant System in normal shutdown situations. Specification A. The reactor shall not be made critical unless: e 1. Two residual heat removal pumps are operable.
- 2. Two residual heat exchangers are operable.
- 3. All system piping and valves, required to establish a flow path to and from the above components, are operable.
- 4. All Component Cooling System piping and valves, required to establish a flow path to and from the above components, are operable.
B. The requirements of Specification A may be modified to allow one of th~ following components (including associated valves and piping) to be in-operable at any one time. If the system is not restored to meet the re-quirements of Specification A within 14 days, the reactor shall be shutdown.
TS 3.5-2 2-1-72
- 1. One residual heat removal pump may be out of service, provided immediate attention is directed to making repairs.
- 2. One residual heat removal heat exchanger may be out of service, provided immediate attention is directed to making repairs.
C. Electrical power to the motor operated valve in the line connecting the Residual Heat Removal System to the Reactor Coolant System, (MOV 1701 for Unit 1 and MOV 2701 for Unit 2) shall be locked out with the valve in the closed position when the reactor coolant pressure exceeds 465 psig. Basis The Residual Heat Removal System is required to bring the Reactor Coolant System fron conditions of approximately 350°F and pressures between 400*and 450 psig to cold shutdown conditions. Heat removal at greater temperatures is by the Steam and Power Conversion System. The Residual Heat Removal System is provided with two pumps and two heat exchangers. If one of the two pumps and/or one of the two heat exchangers is not operative, safe operation of the unit is not affected; however, the time for cooldown to cold shutdown conditions is extended. 'fhe AEC requires that the series motorized valves in the line connecting the RHRS and RCS be provided with pressure interlocks to prevent them from opening when the reactor coolant system is at pressure. Such a feature to satisfy the intent of this criterion shall be installed prior to startup after the first refueling. In the interim period Specification 3.5.C shall be implemented to insure that the RHR system cannot be overpressurized by de-energizing the motor operated valve which is not already pressure interlocked. The
TS 3.5-3 2-1-72 pressure at which this procedure will be implemented is the same as the set-point of the motor operated valve which is interlocked with reactor coolant pressure. References FSAR Section 9.3 Residual Heat Removal System
TS 3.6-1 12~29-71 3.6 TURBINE CYCLE Applicability Applies to the operating status of the Main Steam and Auxiliary Steam Systems Objective To define the conditions required in the Main Steam System and Auxiliary SteaI11 System for protection of the steam generator and to assure the capability to remove residual heat from the core during a loss of station power. Specification A. A unit's Reactor Coolant System temperature or pressure shall not exceed 350°F or 450 psig, respectively, or the reactor shall not be critical unless the five main steam line code safety valves associated with each steam generator in unisolated -reactor coolant loops, are operable. B. To assure residual heat removal capabilities, the following conditions shall be met prior to the commencement of any unit operation that would establish reactor coolant system conditions of 350°F and 450 psig which would preclude operation of the Residual Heat Removal System.
- 1. Two of the three auxiliary feedwat"er pumps shall be operable.
TS 3.6-2 2-1~72
- 2. A minimum of 96,000 gal of water shall be available in the tornado missile protected condensate storage tank to supply emergency water to the auxiliary feedwater pump suctions.
- 3. All main steam line code safety valves, associated with steam generators in unisolated reactor coolant loops, shall be operable.
- 4. System piping and valves required for the operation of the components enumerated in Specification B.l, 2, and 3 shall be operable.
C. Steam generator blowdown will be diverted to the Liquid Waste Disposal 3 System if radioactive contamination in the blowdown exceeds 3.5 x 10- µCi/cc.* D. The iodine - 131 activity in the secondary side of any steam generator, in an unisolated reactor coolant loop, shall not exceed 9 curies. E. The requirements of Specification B-2 above may be modified to allow utilization of protected condensate storage tank water with the auxiliary steam generator feed pumps provided the water level is maintained above 60,000 gallons, sufficient replenishment water is available in the 300,000 gallon condensate storage tank, and replenishment of the protected con-densate storage tank is connnenced within two hours after the cessation of protected condensate storage tank water consumption. Basis A reactor which has been shutdown from power requires removal of core residual heat. While reactor coolant temperature or pressure is greater than 350°F or
TS 3.6-3 12-29-71 450 psig, respectively, residual heat removal requirements are normally satis-fied by steam bypass to the condenser. If the condenser is unavailable, steam can be released to the atmosphere through the safety valves, power operated relief valves, or the 4 inch decay heat release line. The capability to supply feedwater to the generators is normally provided by the operation of the Condensate and Feedwater Systems. In the event of complete loss of electrical power to the station, residual heat removal would continue to be assured by the availability of either the steam driven auxiliary ~eedwater pump or one of the motor driven auxiliary feedwater pumps and the 100,000 gallon condensate storage tank. A minimum of 92,200 gallons of water in the 100,000 gallon condensate tank is sufficient -for 8 hours of residual heat removal following a reactor trip and loss of all off-site electrical power. If the protected condensate storage tank level is reduced to 60,000 gallons, the immediately available replenishment water in the 300,000 gallon condensate tank can be gravity-feed to the pro-tected tank if required for residual heat removal. An alternate supply of feed-water to the auxiliary feedwater pump suctions is also available from the Fire Protection System Main in the auxiliary feedwater pump cuoicle. The five main steam code safety valves associated with each steam generator have a total combined capacity of 3,725,575 pounds per hour at their individual set pressure; the total combined capacity of all fifteen main steam code safety valves is 11,176,725 pounds per hour, The ultimate power rating steam flow is 11,167,923 pounds per hour. The combined capacity of the safety valves required by Specification 3.6 always exceeds the total steam flow corresponding to the maximum steady-state power than can be obtained during:one, two or three reactor
TS 3.6-4 2-1-72 reactor coolant loop operation. The availability of the auxiliary feedwater pumps, the protected condensate storage tank, and the main steam line safety valves adequately assures that sufficient residual heat removal capability will be available when required. The concentration of radionuclides in the discharge canal shall not exceed the values specified in 10CFR20, For an unidentified mixture MPC=lo-7 µCi/cc. Maximum permissible concentration in blowdown = MPC x Discharge Flow Rate Blowdown Rate 7 µCi x 773,000 gpm cc 22 gpm
= 3.5 x 10- 3 µCi cc When the blowdown tank effluent activity becomes less than the maximum permissible concentration it is sent to the discharge canal in the normal manner.
The limit on steam generator secondary side iodine - 131 activity is based on limiting inhalation thyroid dose at the site boundary to 1.5 rem after a postulated accident that would result in the release of the entire contents of a unit's steam generators to the atmosphere. In this accident, with the halogen inventories in the steam generator being at equilibrium values, I-131 would contribute 75 percent of the resultant thyroid dose at the site boundary; the remaining 25 percent of the does is from other isotopes of iodine. In the analysis1 one-tenth of the contained iodine is assumed to reach the site boundary, making allowance for plate out and retention in water droplets.
TS 3.6-5 2-1-72 The inhalation thyroid dose at the site boundary is given by: Dose (Rem) = (C) (x/Q) (D 00 / AT) (B .R.) (
- 7 5) (P *F * )
where: C = steam generator I-131 activity (curies) x/Q = 8.14 X 10-4 sec/m 3 D /AT = 1.48 x 10 6 rem/Ci for I-131 00 B.R. = breathing rate, 3.47 x 10-4 m3/sec. from TID 14844 P.F. c plating factor, 10* Assuming* the postulated accident, the resultant thyroid dose is
- 1. 5 rem.
The steam generator's specific iodine - 131 activity limit is calculated by dividing the total activity limit of 9 curies by the water volume of a steam generator. At full power, with a steam generator water volume of 47.6 M3 , the specific iodine - 131 limit would be .18 µCi/cc; at zero power, with a steam generator water volume of 101 M3, the specific iodine - 131 limit would be .089 µCi/cc. References FSAR Section 4 Reactor Coolant System FSAR Section 9.3 Residual Heat Removal System FSAR Section 10.3.1 Main Steam System FSAR Section 10.3. 2 Auxiliary Steam System FSAR Section 10.3.5 Auxiliary Feedwater Pumps FSAR Section 10.3.8 Vent and Drain Systems FSAR Section 14.3.2.5 Environmental Effects of a Steam Line Break
TS 3.7-1 5-31-71 -* 3.7 INSTRUMENTATION SYSTEMS Operational Safety Instrumentation Applicability: Applies to reactor and safety features instrumentation systems. Objectives: To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable limits are exceeded, and to
- delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety.
Specification: A. For on-line testing or in the event of a sub-system instrumentation channel failure, plant operation at rated power shall be permitted to continue in accordance with TS Tables 3.7-1 through 3.7-3. B. In the event the number of channels of a particular sub-system in service falls below the limits given in the column entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in
- Column 4 of TS Tables 3. 7-1 through 3. 7-3.
TS 3.7-2 5-31-71
- C. In the event of sub-system instrumentation channel failure permitted by specification 3.7-B, TS Tables 3.7-1 through 3.7-3 need not be observed during the short period of time the operable sub-system channels are tested where the failed channel must be blocked to prevent unnecessary reactor trip.
D. The Engineered Safety Features initiation instrumentation setting limits shall be as stated in TS Table 3.7-4. E. Automatic functions operated from radiation monitor alarms shall be as stated in TS Table 3.7-5. Basis
- Instrument Operating Conditions During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design. This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.
Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for channel calibration and test at power. Exceptions
TS 3.7-3 5-31-71
- are backup channels such as reactor coolant pump breakers. The removal of one trip channel on process control equipment is accomplished by placing that channel bistable in a tripped mode; e.g., a two-out-of-three circuit becomes a one-out-of-two circuit. The nuclear instrumentation system channels are not intentionally placed in a tripped mode since the test signal is superimposed on the normal detector signal to test at power. Test'.ing of the NIS power range channel requires: (a) bypassing the Dropped Rod protection from NIS, for the channel being tested; and (b) placing the ~T/T protection CHANNEL avg SET that is being fed from the NIS channel in the trip mode and (c) defeating the power mismatch section of T control channels when the appropriate NIS avg channel is being tested. However, the Rod Position System and remaining NIS channels still provide the dropped-rod protection. Testing does not trip the system unless a trip condition exists in a concurrent channel.
Instrumentation has been provided to sense accident conditions and to (1) initiate operation of the Engineered Safety Features Safety Injection System Actuation Protection against a Loss of Coolant or Steam Break Accident is brought about by automatic actuation of the Safety Injection System which provides emergency cooling and reduction of reactivity. The Loss of Coolant Accident is characterized by depressurization of the Reactor Coolant System and rapid loss of reactor coolant to the containment .
- The Engineered Safeguards Instrumentation has been designed to sense these effects of the Loss of Coolant accident by detecting low pressurizer pressure
TS 3. 7-4 5-31-71
- and level and to generate signals actuating the SIS active phase based upon the coincidence of these signals. The SIS active phase is also actuated by a high containment pressure signal brought about by loss of high enthalpy coolant to the containment. This actuation signal acts as a backup to the low pressurizer pressure and level signal actuation of the SIS and also adds diversity to protection against loss of coolant.
Signals are also provided to actuate the SIS upon sensing the effects of a steam line break accident. Therefore, SIS actuation following a steam line break is designed to occur upon sensing high differential steam pressure between the steam header and steam generator line or upon sensing high steam line flow in coincidence with low reactor coolant average temperature or low
- steam line pressure
- The increase in the extraction of RCS heat following a steam line break results in reactor coolant temperature and pressure reduction. For this reason protection against a steam line break accident is also provided by coincident low pressurizer pressure and level signals actuating safety injection.
Protection is also provided for a steam lin~ break in the containment by actuation of SIS upon sensing high containment pressure. SIS actuation injects highly borated fluid into the Reactor Coolant System in order to counter the reactivity insertion brought about by cooldown of the reactor coolant which occurs during a steam line break accident .
TS 3.7-5 12-29-71
- Containment Spray The Engineered Safety Features also initiate containment spray upon sensing a high-high containment pressure signal. The containment spray acts to reduce containment pressure in the event of a loss of coolant or steam line break accident inside the containment. The containment spray cools the containment directly and limits the release of fission products by absorbing iodine should it be released to the containment.
Containment spray is designed to be actuated at a higher containment pressure (approximately 50% of design containment pressure) than the SIS (10% of design). Since .. spurious actuation of containment spray is to be avoided, it is initiated only on coincidence of high-high containment pressure sensed by 3 out of the 4 containment pre~sure signals provided fo~ its actuation. Steam Line Isolation Steam line isolation signals are initiated by the Engineered Safety Features closing all steam line trip valves. In the event of a steam line break, this action prevents continuous_, uncontrolled steam release from more than one steam generator by isolating the steam lines on high-high containment pressure or high steam line flow with coincident low steam line pressure or low reacto,r coolant average tempera-ture. Protection is afforded for breaks inside or outside the containment even when it is assumed that there is a single failure in the steam line isolation system. Feedwater Line Isolation The feedwater lines are isolated upon actuation of the Safety Injection System
TS 3. 7-6 12-29-71 in order to prevent excessive cooldown of the reactor coolant system. This mitigates the effect of an accident such as steam break which in itself causes excessive coolant temperature cooldown. Feedwater line isolation also reduces the consequences of a steam line break inside the containment, by stopping the entry of feedwater. Setting Limits
- 1. The high containment pressure limit is set at about 10% of design containment pressure. Initiation of Safety Injection protects against 2
1 oss o f coo 1 an t ( ) ors
. t earn 1 1ne
- b reak ( 3 ) acc1. d ents as d 1scusse
. d int
. he safety analysis.
e
- 2. The high-high containment pressure limit is set at about 50% of design containment pressure. Initiation of Containment Spray and Stearn Line Isolation protects agains~ large loss of coolant ( 2 ) or steam line break accidents ( 3 ) as discussed in the safety analysis.
- 3. The pressurizer low pressure setpoint for ~afety injection actuation is set substantially below system operating pressure limits. However, it is sufficiently high to protect against a loss-of-coolant accident as shown int . ( 2) h e saf ety ana 1ys1s.
,. The pressurizer low level limit is set sufficiently high to protect against a loss of coolant accident as shown in the accident analysis. 4.* The steam line high differential pressure limit is set well below
TS 3.7-7 2-1-72 the differential pressure expected in the event of a large steam (3) line break accident as shown in the safety analysis.
- 5. The high steam line flow limit is set approximately 20% of the full steam flow at no load and at 120% of full steam flow at full load, with the high steam line flow differential pressure setpoint linearly programmed between no load and full load in order to protect against large steam line break accidents. The coincident low T setting limit for SIS and avg steam line isolation initiation is set below its hot shutdown value. The coincident steam line pressure setting limit is set below the full load operating pressure. The safety analysis shows that these settings provide (3) protection in the event of a large steam line break.
Automatic Functions Operated from Radiation Monitors The Process Radiation Monitoring System continuously monitors selected lines containing or possibly containing, radioactive effluent. Certain channels in this system actuate control valves on a high-activity alarm signal. Additional 4 information on the Process Radiation Monitoring System is available in the FSAR. ( ) Reference (1) FSAR - Section 7,5 (2) FSAR - Section 14.5 (3) FSAR - Section 14.3.2 (4) FSAR - Section 11.3.3 e
I
- TABLE 3, 7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS . CANNOT BE MET
- 1. Manual 1 Maintain hot shutdown
- 2. Nuclear Flux Power Range 3 2 Low trip setting when 2 Maintain hot
*of 4 power channels greater shutdown than 10% of full power
- 3. Nuclear Flux Intermediate Range 1 2 of 4 power channels greater Maintain hot than 10% full power shutdown
- 4. Nuclear Flux Source Range 1 1 of 2 intennediate rang~lO Maintain hot channels greater than 10 shutdown amps
- 5. Overtemperature tT 2 l' Maiptain hot shutdown
- 6. Overpower tT 2 1 Maintain hot shutdown 7, Low Pressurizer Pressure 2 1 3 of 4 nuclear power channels Maintain hot and 2 of 2 turbine load shutdown channels less than 10% of rated power I-' >-3 N C/l
- 8. Hi Pressurizer Pressure 2 1 *same as Item 7 above Maintain hot N I
v.J shutdown I.O
- I .......
...._. I I-' CXl
- e TABLE 3.7-1 (Cont'd)
REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET
- 9. Pressurizer~Hi Water Level 2 1 3 of 4 nuclear power channels Maintain hot and 2 of 2 turbine load shutdown channels less than 10% of rated power
- 10. Low Flow 2/operable
- If inoperable loop channels Maintain hot loop are not in service they must shutdown be placed in the tripped mode.
- 11. Turbine Trip 2 1 Ma-intain less than 10% rated power
- 12. Lo Lo Steam Generator 2/non-iso-
- 1/non- Maintain hot Water Level lated loop isolated loop shutdown
- 13. Underfrequency 4 KV Bus 2 1 Maintain hot I-' H shutdown N C/l I
N l,..l
\0 *
- 14. Undervoltage 4 KV Bus 2 1 Maintain hot I -..J
-..J I shutdown I-' \0
- 15. Control rod misalignment Monitor***
a) rod position deviation 1 Log individual rod
e TABLE 3.7-1 (Cont'd) REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET positions once/hour, and after a load
. change > 10% or after> 30 inches of control rod motion.
b) quadrant power tilt 1 Log individual upper monitor (upper and and lower ion chamber lower excore neutron currents once/hour detectors) and after a load change 10% or after> 30 inches of control rod , motion.
- 16. Safety Injection See Item 1 of TS Table 3.7-2
- 17. Low steam generator 1/non-iso- Maintain hot shutdown water level with lated loop steam/feedwater 1/non-iso-mismatch flow lated loop
**If both rod misalignment monitors (a and b)
N 1-3 are inoperable for 2 hours or more, the I Cll f-' nuclear overpower trip shall be reset to 93 IW
--Jo percent of rated power in addition to the N--J I
increased surveillance noted. f-' 0
- e TABLE 3. 7-2 ENGINEERED SAFEGUARDS ACTION 1 2 3 4 OPERATOR ACTION MIN. IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET 1 SAFETY INJECTION
- a. Manual 1 0 Cold Shutdown
- b. High Containment Pressure 3 1 Cold Shutdown (Hi Setpoint)
- c. High Differential Pressure 2/non- 1/non-between any Steam Line and isolated isolated Cold Shutdown the Steam Line Header loop loop
- d. Pressurizer Low Pressure 2* 1 Primary Pressure Cold Shutdown an*d Low *Level less than 2000 psig except when reactor is critical
- e. High Steam Flow in 2/3 1/steamline Reactor Coolant aver- Cold Shutdown Steam Lines with Low T ***
1 age temperature less 2 Tavg signals or Low Steam Line Pres~~~e 2 Steam Pres- 1 than 547°F during
- sure Signals heatup and cooldown.
2 CONTAINMENT SPRAY I-' >-3 N Ul I
- a. Manual 2 '~* Cold Shutdown NW
\0.
I .......
-....i I
- b. High Containment Pressure 3 1 Cold Shutdown I-' I-'
I-' (Hi Hi Setpoint)
* - Each channel has two separate signals ** - Must actuate 2 switches simultaneously
- - tHth the specified minimum operable channels the 2/3 high stem'l flow is already in the triµ mode
- e TABLE 3. 7-3 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS 1 -2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET
- 1. CONTAINMENT ISOLATION
- a. Safety Injection See Item No. 1 of Table 3.7-2 Cold Shutdown
- b. Manual 1 Hot Shutdown
- c. High Containment Pressure 3 1 Cold Shutdown (Hi setpoint)
- d. High Containment Pressure
- 3 1 Cold Shutdown (Hi-Hi setpoint)
- 2. STEAM LINE ISOLATION
- a. High Steam Flow in 2/3 lines 1/steamline *** Cold Shutdown and 2/3 Low Tavg or 2 T signals 1 avg 2/3 Low Steam Pressure 2 Steam Pressure. 1 Signals
- b. High Containment Pressure 3 1 Cold Shutdown (Hi Hi Level)
- c. Manual 1/line Hot Shutdown t-' 1-'3
- 3. FEEDWATER LINE ISOLATION N Cf.l I
NW
\0 *
- a. Safety Injection See Item No. 1 of Table 3.7-2 Cold Shutdown I .......i
.......i I t-' t-'
N
- With the specified minimum operable channels the 2/3 high steam flow is already in the trip mode
- TABLE :.. 7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LDHTS lNSTRillfENT SETTING NO. FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT
- l. High Containment Pressure (High Containment a) Safety Injection 2 5 psig Pressure Signal) b) Containment Vacuum Pump Trip c) High Pressure Contain-ment Isolation d) Safety Injection Contain-ment Isolation e) F.W. Line Isolation 2 High High Containment Pressure (High High a) Containment Spray < 25 psig Containment Pressure Signal) b) Recirculation Spray c) Steam Line Isolation d) High High Pressure Con-tainment Isolation 3 Pressurizer Low Pressure and a) Safety .Injection > 1,700 psig Low Level b) Safety Injection Contain- > 5 percent instrument ment Isolation span.
c) Feedwater Line Isolation 4 High Differential Pressure Between any a) Safety Injection < 150 psi Steam Line and the Steam Line Header b) Safety Injection Contain-ment Isolation c) F.W. Line Isolation 5 High Steam Flow in 2/3 Steam Lines a) Safety Injection < 20% (at zero load) of full b) Steam Line Isolation steam flow c) Safety Injection Contain- < 120% (at full load) of full ment Isolation steam flow d) F.W.Line Isolation Coincident with Low T or Low Steam > 541°F T avg avg I-' ,_, Line Pressure > 500 µsig steam line ~ en pressure Nw I.O. I -....i
-..J I I-' I-'
v,;
I. *
- TABLE 3. 7-5 AUTOMATIC FUNCTIONS OPERATED FROM RADIATION MONITORS ALARM AUTOMATIC FUNCTION MONITORING ALARM SETPOINT MONITOR CHANNEL AT ALARM CONDITIONS REQUIREMENTS µCi/cc
- 1. Process vent particulate and Stops discharge from containment See Specifications Particulate 4xlo- 8 gas monitors vacuum systems and waste gas decay 3.11 and 4.9 Gas 9xlo-2 (RM-GW-101 & RM-GW-102) tanks (Shuts Valve Nos.
RCV-GW-160, FCV-GW-260, FCV-GW-101)
- 2. Component cooling water Shuts surge tank vent valve See Specification Twice Background radiation monitors HCV-CC-100 4.9 (RM-CC-105 & RM-CC-106)
- 3. Liquid waste disposal Shuts effluent discharge See Specifications l.5x10- 3 radiation monitor valves FCV-LW-104A and FCV-LW-104B 3.11 and 4.9 (RM-LW-108)
- 4. Condensor air ejector Diverts flow to the containment of See Specification 1.3 radiation monitors the affected unit 4.9 (RM-SV-111 & RM-SV-211) (Opens TV-SV-102 and shuts TV-SV-103 or opens TV-SV-202 and shuts TV-SV-203)
- 5. Containment particulate Trips affected unit's purge supply See Specifications Particulate 9xl0- 9 and gas monitors and exhaust fans, closes affected 3.10 and 4.9 Gas lxlo-5 (RM-RMS-159 &. RM-RMS~l60, unit's purge air butterfly valves RM-RMS-259 & RM-RMS-260) (MOV-VS-lOOA, B, C & Dor MOV-VS-200A, B, C & D)
- 6. Manipulator crane area Trips affected unit's purge supply See Specification 50 mrem/hr monitors (RM-RMS-162 & and exhaust fans, closes affected 3.10 and 4.9 RM-RMS-262) unit's purge air butterfly valves N t-3 I tr.l (MOV-VS-lOOA, B, C & Dor MOV-VS- I-'
I W 200A, B, C & D ......... N -...J I I-'
~
TS 3.8-1 2-1-72 3.8 CONTAINMENT Applicability Applies to the integrity and operating pressure of the reactor containment, Objective To define the limiting operating status of the reactor containment for unit operation. Specification A. Containment Integrity and Operating Pressure
- 1. The containment integrity, as defined in TS Section 1.0, shall not be violated, except as specified in A2, below, unless the reactor is in the cold shutdown condition,
- 2. The reactor containment shall not be purged while the reactor is operating, except as stated in Specification A.3.
- 3. During the plant startup, the remote manual valve on the steam jet air ejector suction line may be open, if under administrative control, while containment vacuum is being established. The Reactor Coolant
- System temperature and pressure must not exceed 350°F and 450 psig, respectively, until the air partial pressure in the containment has been reduced to a value equal to, or below, that specified in TS Figure 3.8-1.
TS 3. 8-2 2-1-72
- 4. The containment integrity shall not be violated when the reactor vessel head is unbolted unless a shutdown margin greater than 10 percent A k/k is maintained.
- 5. Positive reactivity changes shall not be made by rod drive motion or boron dilution unless the containment integrity is intact.
B. Internal Pressure
- 1. If the internal air partial pressure rises to a point 0.25 psi above the present value of the air partial pressure (TS Figure 3.8-1),
the reactor shall be brought to the hot shutdown condition. e 2. If the leakage condition cannot be corrected without violating the containment integrity or if the internal partial pressure continues to rise, the reactor shall be brought to the cold shutdown condition utilizing normal operating procedures. 3, If the internal pressure falls below 8.25 psia the reactor shall be placed in the cold shutdown condition. Basis The Reactor Coolant System temperature and pressure being below 350°F and 450 psig, respectively, ensures that no significant amount of flashing steam will be formed and hence that there would be no significant pressure buildup in the containment if there is a loss-of-coolant accident.
TS 3. 8-3 2-1-72 - The shutdown margins are selected based on the type of activities that are being carried out. The 10 pe~centAk/k shutdown margin during refueling precludes criticality under any circumstance, even though fuel and control rod assemblies are being moved.
\
The subatmospheric air partial pressure used for normal operation is maintained between 9.0 and 11.0 psia. The set value of this partial pressure depends on the temperature of the service water which is used to cool the recirculation spray coolers, the ambient air temperature in the containment and the heat sink capability of the refueling water storage tank. The allowable air partial pressure is given in TS Figure 3.8-1 as a function of service water temperature, the refueling water storage tank temperature, and the ambient containment air temperature. If the air partial pressure rises 0,10 psi above this set value, action shall be taken innnediately to correct the condition. If the condition cannot be corrected, the reactor will be manually shutdown and cooldown initiated. Shutdown of the reactor ensures that the contairnnent design pressure of 45 psig will not be exceeded and that depressurization can still be accomplished in the unlikely event of a loss-of-coolant accident. The contairnnent isolation system is not activated by the contairnnent high pressure alarm until the containment high pressure set point is reached. Figure 3.8-1 has been established to provide the operator with information con-cerning where the air partial pressure must be maintained as a function of RWST e
TS 3.8-4 2-1-72
- and service water temperature to insure depressurization within 40 minutes following a loss-of-coolant accident, In the event of a loss-of-coolant accident the containment is brought back to subatmospheric conditions by cooling the containment with the containment sprays and recirculation sprays.
Since the containment spray is from the refueling water storage tank (RWST) and the recirculation spray is cooled by service water, the rate at which the containment is cooled becomes a function of the RWST water temperature and the service water temperature. The pressure to which the containment returns for a given RWST water temperature and service water temperature following a LOCA depends on the initial air partial pressure, Thus, for a given combination of RWST water temperature and service water temperature, the .containment air partial pressure must be regulated below a.certain level to assure that the containment will become sub-atmospheric within 40 minutes following a LOCA. The lowest set value of air partial pressure in the containment is 9.0 psia. This corresponds to a total pressure of approximately 9.5 psia.at an*air
.. _.,,.._,.,* *-";"- .~. *~*;._..
temperature of 105°F and 80°F dew point. The shell and dome plate 1-iner of the containment are capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia. References FSAR Section 4.3.2 Reactor Coolant Pump FSAR Section 5.2 Containment Isolation FSAR Section 5.2.1 Design Bases FSAR Section 5.2.2 Isolation Design
TS FIG. 3.8-1 MARCH 15, 1971
- ~
(/) 0. 11.2 11.0
~ .......... - *
--- -- ~--- ------ - - - - - - t - - - -
~r--....K
*w a::
- )
(/) 10.8 ~ - -- ~ - - r----*----* (/) w ['-.... a:: 10.6 -- ~- .-- a..
' REFUELING WATER STORAGE TANK
""'~ r ~
__J
<( TEMPERATURE ;::: 10.4 *-**--
a:: - - - ......-45°F
<(
0.
~ - ~---55°F a:: 10.2
~
<(
(.!) z ___ i
~""I
;::: 10.0 - - ,_ t----- - I',,,
<(
"' '\i"!"'
a:: w a.. 0 9.8 I w__J CD
<( 9.6 ~ !
I\, 0 __J __J
<(
9.4 II '~\
- )
~
x 9.2 I r"\
<( ~ I '0 ~\
9.0 l I I/ I 35 45 55 65 75 85 95 105 SERVICE WATER TEMPERATURE - °F NOTES: MAXI MUM ALLOWABLE OPERATING AIR PARTIAL PRESSURE* IN THE CONTAINMENT AS A FUNCTION OF SERVICE WATER TEMPERATURE AND REFUELING WATER STORAGE TANK TEMPERATURE.
*SET POINT VALUE IN CONTAINMENT VACUUM SYSTEM INSTRUMENTATION MINIMUM AVE SERVICE CONTAINMENT WATER TEMPERATURE TEMPERATURE 95°F 95°F 95 75 85 55 75 35
- MAXIMUM ALLOWABLE OPERATING AIR PARTIAL PRESSURE SURRY POWER STATION
TS 3.9-1 2-13-70
- 3.9 STATION SERVICE SYSTEMS Applicability Applies to availability of electrical power for operation of station auxiliaries.
Objective To define those conditions of electrical power availability necessary to pro-vide for safe reactor operation. Specification
- A. A unit's reactor shall not be made critical without:
- 1. All three of the unit's 4,160 v buses energized
- 2. All six of the unit's 480 v buses energized
- 3. Both of the 125 v d-c buses energized
- 4. One battery charger per battery operating
- 5. Both of the 4,160 v emergency buses energized
- 6. Both of the 480 v emergency buses energized
- 7. Two emergency diesel generators operable as explained in Section 3.16.
TS 3.9-2 5-31-71
- B. The requirements of Specification 3.9-A above may be modified for two reactor coolant loop operation to allow one of the unit's 4160 v normal buses and the two 480 v normal buses feed from this 4160 v bus, to be un-available or iQoperable.
Basis During startup of a unit, the station's 4,160 v and 480 v normal and emergency buses are energized from the station's 34.5 kv buses. At reactor power levels greater than 5 percent of rated power the 34.5 kv buses are required to energize only the emergency buses because at this power level the station generator can supply sufficient power to the normal 4,160 v and 480 v lines to operate the unit .
- Three reactor coolant loop operation with all 4160 v and 480 v buses energized is the normal mode of operation for a unit. Equipment redundacy and bus arrangements, however, allow safe unit startup and operation with one 4160 v normal bus and the two 480 v normal buses feed from this 4160 v bus, unavailable or inoperable.
References FSAR Section 8.4 Station Service Systems FSAR Section 8.5 Emergency Power Systems
TS 3.10-l 5-1-71
- 3.10 REFUELING Applicability Applies to operting limitations during refueling operations.
Objective To assure that no accident could occur during refneling operations that would affect public health and safety. Specification A. During refueling operations the following conditions are satisfied:
- 1. The equipment door and at least one door in the personnel air lock shall be properly closed. For those systems which provide a direct path from containment atmosphere to the outside atmosphere, all automatic containment isolation valves in the.unit shall be operable or at leas 4 one valve shall be closed in each line penetrating the containment.
- 2. The Containment Vent and Purge System and the area and airborne radiation monitors which initiate isolation of this system, shall be tested and verified to be operable innn.ediately prior to refueling operations.
TS 3.10-2
- 3.
2-1-72 At least one source range neutron detector shall be in service at all times when the reactor vessel head is unbolted. Whenever core geometry or coolant chemistry is being changed, subcritical neutron flux shall be continuously monitored by at least two source range neutron detectors, each with continuous visual indi-cation in the Main Control Room and one with audible indication within the containment.During core fuel loading phases, there shall be a minimum neutron count rate detectable on two operating source range neutron detectors operating source range neutron detectors with the exception of initial core loading, at which time a mini-mum neutron count rate need be established as soon as there are eight (8) fuel assemblies loaded into the reactor vessel.
- 4. Manipulator crane area radiation levels and airborne activity levels within the containment and airborne activity levels in the ventilation exhaust duct shall be continuously monitored during refueling. A manipulator crane high radiation alarm or high airborne activity level alarm within the containment will auto-matically stop the purge ventilation fans and automatically close the containment purge isolation valves.
- 5. Fuel pit bridge area radiation levels and ventilation vent exhaust airborne activity levels shall be continuously monitored during refueling. The fuel building exhaust will be continuously bypassed through the iodine filter bank during refueling procedures, prior to discharge through the ventilation vent *
- 6. At least one residual heat removal pump and heat exchanger shall be operable to circulate reactor coolant.
TS 3.10-3 12-29-71
- 7. When the reactor vessel head is unbolted, a minimum boron concen-tr at ion of 2,000 ppm shall be maintained in any filled portion of the Reactor Coolant System and shall be* checked by sampling at least once every 8 hours, 8, Direct communication between the Main Control Room and the refueling cavity manipulator crane shall be available whenever changes in core geometry are taking place.
9, No movement of irradiated fuel in the reactor core shall be accomplished until the reactor' has been subcritical for a period of at least 100
- hours.
10, A spent fuel* cask or other heavy loads exceeding 110 percent of the weight of a fuel-assembly shall not be moved over spent fuel, and only one spent fuel assembly will be handled at one time over the reactor or the spent fuel pit, B, If any of the specified limiting conditions for refueling are not met,
- refueling of. the reactor shall cease, work shall be initiated to correct the conditions so that the specified limits are met, and no operations which increase the reactivity of the core shall be made.
C, After initial fuel loading and after each core refueling operation and prior to reactor operation at greater than 75% of rated power, the moveable incore detector system shall be utilized to verify proper power distribution. Basis Detailed instructions, the above specified precautions and the design of the
TS 3.10-4 6-30-71 fuel handling equipme~t, which incorporates built-in interlocks and safety features, provide assurance that an accident, N"hich would result in a hazard to public health and safety, will not occur during 1'."efueling operations. When no change is being made in core geometry, one neutron detector is Rufficient to monitor the core and permits maintenance of the out-of-function instrumentation. Continuous monitoring of radiation levels and neutron flux provides irmn.ediate indication of an unsafe condition. Containment high radiation levels and high airborne activity levels automatically stop and isolate the Containment Purge System. The fuel building ventilation exhaust is .diverted through charcoal filters whenever refueling is in progress. At least one flow path is required for cooling and mixing the coolant contained in the reactor ves~el so as to maintain a uniform boron concentration and to remove residual heat. The shutdown margin established by Specification A-7 maintains the core sub critical, even with all of the control rod assemoJ.ies withdrawn from the core. During refueling, the reactor refueling water cavity is filled with approximately 220,000 gal of water borated to at least 2,000 ppm boron. The boron concentration of this water is sufficient to maintain the reactor subcritical by approximately 10% /J. k/k in the cold shutdown condition with all control rod assemblies inserted and also to maintain.the core subcritical by approximately 1% with no control rod assemblies inserted into the reactor. Periodic checks of refueling water boron concentration assure the proper shutdown margin. Specification A--8 allows the Control Room Operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.
TS 3.10-5 5-1-71 In addition to the above safeguards, interlocks are used during refueling to assure safe handling of the fuel assemblies. An "!Xcess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. The spent fuel transfer mechanism can accommodate only one fuel assembly at a time. Upon each completion of core loading and installation of the reactor vessel head, specific mechanical and electrical tests will be performed prior to initial criticality. The fuel handling accident has been analyzed based on the activity that could be released from fuel rod gaps of 204 rods of the highest power assembly with a 100 hour decay period following power operation at 2550 MWt for 23,000 hours. The requirements detailed in Specification 3.10 provide assurance that refueling plant conditions conform to the operating conditions assumed in the accident analysis. Detailed procedures and checks insure that fuel assemblies are loaded in the proper locations in the core. As an additional check, the moveable incore detector system will be used to verify proper power distribution. This system is capable of revealing any assembly enrichment error or loading error which could cause power shapes to be peaked in excess of design value. Ref ere.rices
- FSAR Section 5.2 Containment Isolation
TS 3.10-6 5-1-71 FSAR Section 6.3 Consequence Limiting Safeguards FSAR Section 9.12 Fuel Handling System FSAR Section 11.3 Radiation Protection FSAR Section 13.3 Tahle 13.3-1 FSAR Section 14.4.1 Fuel Handling Accidents FSAR Supplement: Volume I: Question 3~2
TS 3.11-1 5-1-71 3.11 EFFLUENT RELEASE Applicability Applies to the release of radioactive liquids and gases from the station during normal operation. Ohjective. To define. the limiting conditions for release of radioactive wastes from the circulating water dis-chP,rge, the process vent, and the ventilation vent to pre.vent off-si.te doses from exceeding limits of 10CFR20 . Specification A. Liquid Wastes
- 1. The. controlled release rate of radioactive liquid effluents from the station shall be s.uch that the concentration of radionuclides leaving the circulating water discharge canal will not exceed the limits specified in 10CFR20, Appendix B, for unrestricted areas.
- 2. friar to release of liquid waste, a sample shall be taken and analyzed to demonstrate. compliance with A-1 above, using the
- existing circulating water discharge rate .
TS 3.11-2 5-31-71
- 3. Liquid waste gross activity and flow rate shall be continuously monitored and recorded during release by the Liquid Waste Disposal
*Radiation Monitor and the liquid waste flow recorder._ During liquid waste release the Circulating Water Discharge Tunnel Radiation monitor for Unit No. 1 shall be operating.
- 4. Steam generator blowdown shall be continuously monitored by the Steam Generator Blowdown Sample Monitor System. If radioactive contamination exceeds the value in Specification 3.6-C, steam generator blowdown shall be diverted to the Liquid Waste Disposal System.
- 5. Prior to and during liquid waste releases, circulating water flow shall be established in the discharge tunnel for.Unit No. 1.
B. Gaseous Wastes
- 1. The controlled release rates of gaseous and airborne particulate wastes originating from station operation from the process vent and ventilation vent together shall be limited as follows:
Qi < 2.0 x 10 5 , m3/sec (MPC) i where Qi is the controlled release rate (Ci/sec) of any radioisotope i and (MPC)i' in units of µCi/cc, are defined in Column 1, Table II of Appendix B to 10CFR20, except that for
TS 3 .11-3 5-31-71
- halogen and particulate isotopes with half lives greater than 8 days, the values of (MPC)i shall be reduced by a factor of 1/700.
- 2. Gaseous waste gross and particulate activity and flow rate shall be continuously monitored and recorded during release of radioactive gaseous wastes to the process vent.
- 3. During release of radioactive gaseous waste to the process vent, the following conditions shall be met:
- a. At least one process vent blower shall be operating.
- b. The Process Vent Gas Monitor and the Particulate Monitor shall be operating.
- 4. All effluents to be discharged to the atmosphere from the waste gas decay tanks of the Gaseous Waste Disposal System shall be sampled and analyzed to demonstrate compliance with B-1 above prior to release via the process vent.
- 5. Whenever the air ejector discharge monitor is inoperable and the steam generator blowdown monitors indicate an increase in secondary side activity, samples shall be taken from the air ejector discharge and analyzed for gross activity on a daily basis.
Basis Liquid waste from the Liquid Waste Disposal System is diluted in the Circulating
TS 3.11-4 6-30-71 - Water System prior to release to the James River. Each unit has four circulating water pumps, each having a rated capacity of approximately 210,000 gpm to main-tain the level in the high level intake canal. From the canal, the circulating water flows by gravity to the unit condenser, through a discharge tunnel and then through a discharge canal, common to both units, to the river. The radioactive liquid waste is discharged to the Circulating Water System in the discharge tunnel for Unit 1. Because of the low radioactivity levels in the circulating water discharge, the concentrations of radionuclides in the circulating water discharge canal are calculated from the gross activity of the Liquid Waste Disposal System effluent and the steam generator blowdown effluent, the flow of this effluent, and the flow in the Circulating Water System. The concentration is also measured by waste effluent sample analysis. Prior to release to the atmosphere, gaseous waste from the Gaseous Waste Disposal System is mixed in the process vent with flow from at least one of the process vent blowers. Further dilution occurs in the atmosphere. The formula prescribed in Specification B-1 takes atmospheric dilution into account and assures that at the point of maximum ground concentration at the site boundary the requirements of 10CFR20 will not be exceeded. The limit is based on the highest annual average value of X/Q which will occur at the site boundary and is 5.0 x 10- 6 sec/m 3 . The requirement to reduce the MPC's of 10CFR20 by a factor of 700 for halogen and particulate isotopes with half lives greater than 8 days conservatively limits exposures from airborne radioactive materials that may enter terrestrial food chains.
TS 3.11-5 5-31-71
- References FSAR Section 2.2.3 FSAR Section 11. 2. 3 Average Atmospheric Dilution Liquid Waste Disposal System FSAR Section 11. 2. 5 Gaseous Waste Disposal System FSAR Section 11. 3. 3 Process Radiation Monitor Systems FSAR Section 11. 3 .4 Area Radiation Monitor Systems-
TS 3.12-1 2-1-72 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS Applicability Applies to the operation of the control rod assemblies and power distribution limits. Objective To ensure core subcriticality after a reactor trip, a limit on potential reactivity insertions from a hypothetical control rod assembly ejection, and an acceptable core power distribution during power operation. Specification A. Control Bank Insertion Limits
- 1. Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the shutdown control rods shall be fully withdrawn.
- 2. Whenever the reactor is critical, except for physics tests and control rod assembly exercises,. the control rod assembly groups shall be no further inserted than the limits shown on TS Fig.
3.12-1, 3.12-2, or 3.12-3 for three loop operation and TS Fig. 3.12-4, 3.12-5 or 3.12-6 for two loop operation.
TS 3.12-2 2-1-72
- 3. The limits shown on TS Figures 3.12-1 through 3.12-6 may be revised on the basis of physics calculations and physics data obtained during unit startup and subsequent operation, in accordance with the following:
- a. The sequence of withdrawal of the controlling banks, when going from zero to 100% power, is A, B, C, D.
- b. An overlap of control banks, consistent with physics calculations and physics data obtained during unit startup and subsequent operation, will be permitted.
- c. The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown on TS Figure 3.12-7 under all steady-state operating conditions, except for physics tests, from zero to full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at hot shutdown conditions (T >547°F) if all control rod assemblies were tripped, avg-assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon, boron, or part-length rod position.
TS 3.12-3 2-1-72
- 4. Whenever the reactor is subcritical, except for physics tests, the critical rod position, i.e., the rod position at which criticality would be achieved if the control rod assemblies were withdrawn in normal sequence with no other reactivity changes, shall not be lower than the insertion limit for zero power.
- 5. The part length rod bank may be moved over the entire travel range, full out to full in, as required by axial power dis-tribution control.
- 6. Insertion limits do not apply during physics tests or during periodic exercise of individual rods. However, the shutdown margin indicated in TS Figure 3.12-7 must be maintained except for the low power physics test to measure control rod worth and shutdown margin. For this test the reactor may be critical with all but one full length control rod~expected to have the highest worth, inserted and part length rods fully withdrawn.
B. Power Distr.ibution Limits 1, If the power tilt ratio exceeds 1.10, except for physics test, or if a part-length or full-length control rod is more than 15 inches out of alignment with its bank, then within eight hours:
- a. The situation shall be corrected, or
------1 TS 3.12-4 2-1-72
- b. The hot channel factors shall be determined and maximum allowable power shall be reduced one percent for each percent the hot channel factor exceeds the values of Design Limits Interim Limits N N F = 2. 72 FN = 1. 58 FN = 2.52 FL'iH - 1.50 q L'iH q or
- c. Power shall be limited to 75% of rated power for 3 loop operation or 45% of rated power for 2 loop operation.
- 2. If after a period of 24 hours, the power tilt ratio in 1. above is not corrected to less than 1.10, and
- a. If design hot channel factors are not exceeded, an evaluation as to the cause of the discrepancy shall be made and reported to the Atomic Energy Commission, within 30 days.
- b. If design hot channel factors are exceeded and power is greater than 10%, the nuclear overpower.and overpowerAT trips shall be reduced one percent for each percent the hot channel factor exceeds the design values in the Technical Specifications.
- 3. If the power tilt ratio exceeds 1.25, except for physics tests, then the reactor shall be put in the hot shutdown condition,
TS 3.12-5 2-1-72 and the Atomic Energy Commission notified per Specification 6.6,Power operation for the purpose of testing is permitted, provided that maximum power levels are less than 50% of rated thermal power.
- I
- 4. Upon determination of FN and F~, as required by Specification q 1.1H 4.10.B, allowable power shall be reduced one percent for each per-cent these hot channel factors exceed the limiting values specified in these Technical Specifications. This restriction on allowable power will remain in effect until the anomaly is corrected.
C. Inoperable Control Rods
- 1. A control rod assembly shall be considered inoperable if the assembly cannot be moved by the drive mechanism, or the assembly remains misaligned from its bank by more than 15 inches. A full-length control rod shall be considered inoperable if its rod drop time is greater than 1.8 seconds to dashpot entry.
- 2. No more than one inoperable control rod assembly shall be permitted when the reactor is critical.
- 3. If more than one control rod assembly in a given bank is out of service because of a single failure external to the individual rod drive mechanisms, i.e. programming circuitry, the provisions of Specification Cl and 2 shall not apply and the reactor may remain critical for a period not to exceed two hours provided immediate attention is directed toward making the necessary repairs. In the event, the affected assemblies cannot be returned
TS 3.12-6 2-1-72
- -- to service within this specified period the reactor will be brought to hot shutdown conditions.
- 4. The provisions of Specifications Cl and 2 shall not apply during physics test in which the assemblies are intentionally misaligned.
- 5. If an inoperable full-length rod is located below the 200 step level and is capable of being tripped, or if the full-length rod is located below the 30 step level whether or not it is capable of being tripped, then the insertion limits in TS Figure 3 .12-2 apply.
- 6. If an inoperable full-length rod cannot be located, or if the in-operable full-length rod is located above the 30 step level and cannot be tripped, then the insertion limits in TS Figure 3.12-3 apply.
7, No insertion limit changes are required by an inope~able part-length rod.
- 8. If a full-length rod becomes inoperable and reactor operation is continued the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days. The analysis shall include due allowance for nonuniform fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically determined part power level which is consistent with the safety analysis.
TS 3.12-7 2-1-72 D. If the reactor is operating above 75% rated power with one excore nuclear channel out of service, the core quadrant power balance shall be determined
- 1. Once per day, and
- 2. After a change in power level greater than 10% or more than 30 inches of control rod motion.
The core quadrant power balance shall be determined by one of the following methods:
- 1. Movable detectors (at least two per quadrant)
- 2. Core exit thermocouples (at least four per quadrant).
E. Inoperable Rod Position Indicator Channels
- 1. If a rod position indicator channel is out of service then:
a) For operation between 50% and 100% of rated power, the position of the RCC shall be checked indirectly by core instrumentation (excore detector and/or thermocouples and/or movable incore detectors) every shift or subsequent to motion, of the non-indicating rod, exceeding 24 steps, whichever occurs first. b) During operation below 50% of rated power no special monitoring is required.
-1 TS 3.12-8 2-1-72
- 2. Not more than one rod position indicator (RPI) channel per group nor two RPI channels per bank shall be permitted to be inoperable at any time.
Basis The reactivity control concept assumed for operation is that reactivity changes accompanying changes in reactor power are compensated by control rod assembly motion. Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature {operating temperature to cold shutdown) are compensated for by changes in the soluble boron concentration. During power operation, the shutdown groups are fully withdrawn and control of power is by the control groups. A reactor trip occurring during power operation will place the reactor into the hot shutdown condition. The control rod assembly insertion limits provide for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod assembly remains fully withdrawn, with sufficient margins to meet_ the assumptions used in the accident analysis. In addition, they provide a limit on the maximum inserted rod worth in the unlikely event of a hypothetical assembly ejection, and pro"vide for acceptable nuclear peaking factors. The limit may be determined ., on the basis of unit startup and operating data to provide a more realistic limit which will allow for more flexibility in unit operation and still assure compliance with the shutdown requirement. The maximum shutdown margin
TS 3.12-9 2-1-72 requirement occurs at end of core life and is based on the value used in analysis of the hypothetical steam break accident. The rod insertion limits are based on end of core life conditions. Early in core life, less shutdown margin is required, and TS Figure 3.12-7 shows the shutdown margin equivalent to 1.77% reactivity at end-of-life with respect to an uncontrolled cooldown. All other accident analyses are based on 1% reactivity shutdown margin. Relative positions of control rod banks are determined by a specified control rod bank overlap. This overlap is based on the considerations of axial power shape control. It is not based on safety criteria, but minimizes possible changes in axial offset accompanying control rod motion. Positioning of the part-length assemblies is governed by the requirement to maintain the axial power shape within specified limits or to accept an automjtic cutback of the overpower ~T and overtemperature ~T set points (see Specification 2.3). Thus, there is no need for imposing a limit on the physical positioning of the part-length assemblies. The various control rod assemblies (shutdown banks, control banks A, B, C D and part-length rods) are each to be moved as a bank, that is, with all assemblies in the bank within one step (5/8 inch) of the bank position. Position indication is provided by two methods: a digital count of actuating pulses which shows the demand position of the banks and a linear position indicator, Linear Variable Differential Transformer, which indicates the actual assembly position. The position indication accuracy of the pulse
TS 3.12-10 2-1-72 count is within one step (5/8 inch). The accuracy of the Linear Variable Differential Transfonner is approximately +5% of span (+/-7,5 inches) under steady state conditions.(l) The relative accuracy of the linear position indicator is such that, with the most adverse errors, an alarm is actuated if any two assemblies within a bank deviate by more than 14 inches. In the event that the linear position indicator is not in service, the effects of malpositioned control rod assemblies are observable from nuclear and process information displayed in the Main Control Room and by core thermocouples and in-core movable detectors. Below 50% power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment(part-length or full length control rod assembly 12 feet out of alignment with its bank) operation at 50% steady state power does not result in exceeding core limits. The specified control rod assembly drop time is consistent with safety analyses that have been perfonned.(2) An inoperable control rod assembly imposes additional demands on the operators. The permissible number of inoperable control rod assemblies is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemblies upon reactor trip.
TS 3.12-11 2-1-72 A quadrant to average power tilt will be indicated by the excore detectors. The excore current tilt is indicated by the arrangement of the current recorders on the control board. Four 2-pen recorders are provided, the pens are grouped so that, in the absence of a tilt, the two ink traces coincide. Any divergence in the traces indicates a power tilt. Furthermore, a maximum-to-average alarm is provided for the upper and lower sets of excore currents. A power tilt ration of 1.1 is not expected to result in peaking factors greater than design under expected operating conditions. If, instead of using movable in-core detectors to measure the hot channel factors, the operator chooses simply to reduce power, the specified limit of 75% power maintains the minimum design margin to core safety limits for up to 1.25 tilt ratio. Resetting of the overpower trip set points ensures that the protection system basis is maintained for sustained unit operation. A tilt ratio of 1.25 or more is indicative of a serious performance anomaly and a unit shutdown is prudent. References (1) FSAR Section 7.2. (2) FSAR Section 14.
TS FIGURE 3.12-1 2-1-72 FIGURE 3.12-1 CONTROL BANK INSERTION LIMITS FOR NORMAL 3 LOOP OPERATION (0.2) (0.78) 0.0 --~~~--~~---~~------~---------..;...,--;..._~~--, BANK B Cl I.J.J 0.2 Ix LL.I V) z: z: {0.35) 0
...... 0.4 t-u BA~K C
~
0:. u. z: a I-1/) 0.6 0 Q.
~
z: - - BANK D
~ (0.76) 0.8 1.0 L-----------.L--------,1&......!i-----------'-----------~---------~
0.0 0.2 (0.34)0.4 0.6 0.8 1.0 FRACTION OF RATED POWER e I j* i
TS FIGURE 3.12-2 2-1-72 FIGURE 3.12-2 CONTROL BANK INSERTION LIMITS FOR 3 LOOP OPERATION WITH ONE BOTTOMED ROD 0.0 17
/
/
I/ 0.2 V V BANKC /
/ /
0 w
~v /
~
)7 /
a: 0.4 V w en / V z / 17
/v z /v V BANKO 0 /
~ V fl7 ~ 0.6
/
17 a: LL / V
/
/
0.8 / i/ I/ I/ 1.0 V 0.0 0.2 0.4 0.6 0.8 1.0 FRACTION FULL POWER
TS FIGURE 3.12-3 2-1-72 FIGURE 3.12-3 CONTROL BANK INSERTION LIMITS FOR 3 LOOP OPERATION WITH ONE INOPERABLE ROD 0.0
/ i/
/ V
/ BANKC 7 0.2 / V V V V V 0
/ /
IJJ er: 0.4 t- / z IJJ / ~- en z BANKO --
/ 0::
f2-z / 0 t- / ~- u 0.6 !i- <t ff / . a:
~-
/ c
/
0.8 /
/
/1/
V 1.0 0.0 0.2 0.4 0.6 0.8 1.0 FRACTION FULL POWER
TS FIGURE 3.12-4 2-1-72 FIGURE 3.12-4 CONTROL BANK INSERTION LIMITS FOR 2 LOOP NORMAL OPERATION 0.0 I/ V
~ANKC
/
0.2 V
/
y C / I.LI .... / ffi 0.4 V
/
en / z V
/
z / 0 / .... V BANKO
/
~ 0.6 /
/
/
/
0.8 /
/
1.0 0.0 0.2 0.4 0.6 FRACTION FULL POWER
FIGURE 3.12-5 CONTROL BANK INSERTION LIMITS FOR 2 LOOP OPERATION FIGURE 3.12-6 e WITH ONE BOTTOMED ROD WITH ONE INOPERABLE ROD 0.0 / 0.0
/ V V
V
/ /
/
/
BANKC I/ 0.2 / / 0.2 / V
/ /
/v / BANKO
~
/ I/
C / 0 / w V w /v t-a:: w u, 0.4
/
V
/
BANKO- -- t-a:: 0.4 w Cl) /
/
z V z V
/
z V z 0 / 0 t-u 0.6
/v f-(.)
0.6 <( a:: I.L
,v <(
ff V I/ 0.8 V 0.8 1.0 1.0 o.o 0.2 0.4 0.6 0.0 0.2 0.4 0.6 t-3 t-3 Cfl Cfl FRACTION FULL POWER FRACTION FULL POWER 1-rj 1-rj HH G') G') qq
- ,;:1 :,;:1 trj trj NWW I *
- t-' t-' t-'
INN
-..J I I N 0\ \JI
FIGURE 3 .12-2 REQUIRED _SHUTDOWN REACTIVITY AS l\FUNCTION. OF_ REACTOR. COOLANT .BORON CONCENTR.A.TION 2.0 1.8 1.6
~ 1.4 I-
- > 1.2 I-u ct:
w 0:: 1.0 z 3 0 C I- 0.8
=>
- c en C
LLJ 0.6 0:: a LI.I 0:: 0.4 0.2 0 0 I 500 I l 000 l 1500 G ") C
;::o rr, Nw BORON CONCENTRATION (ppm) -I _
~N N..!..J
TS 3.13-1 12-29-71 3.13 COMPONENT COOLING SYSTEM Applicaliility Applies. to the operational status of all subsystems of the Component Cooling Sys:tem.. The Component Cooling System consists of the Component Cooling Water Subsystem, Chilled Component Water Subsystem, Chilled Water Subsystem, Neutron Shield Tank Cooling Water Subsystem and Charging Pump Cooling Water Subsystem. Objective To define limiting conditions for each subsystem of the Component Cooling Syst~ necessary to assure safe operation of each reactor unit of the station during startup, power operation, or cooldown * . Speci:Ucations A. When a unit's Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively, or when a unit's reactor is critical operating conditions for the Component Cooling Water Subsystem shall be l as follows:
- 1. For one unit operation, two component cooling water pumps and
__J
TS 3.13-2 12-29-71 heat exchangers shall be operable.
- 2. For two unit operation, three component cooling water pumps and heat exchangers shall be operable.
- 3. The Component Cooling Water Subsystem shall be operable for immediate supply of cooling water to the following components, if required:
~- Two operable residual heat removal heat exchangers.
- h. Seal water and stuffing box jacket of two operable residual heat remo_val pumps.
- 4. Durip.g power operation, Specification A-1, A-2, or A-3 above may be J modified to allow one of the_required components to be inoperable provided immediate attention is directed . to making repairs. If the system is not restored within 24 hours to the requirements of Specification A-1, A-2, or A-3, an operating'reactor shall be placed I in .the hot shutdown condition. If the repairs are not completed within an additional 48 hours, the affected reactor shall be placed in the cold shutdown condition.
B. For each unit whose Reactor Coolant System exceeds a temperature of 350°F and a pressure of 450 psig, or when a unit's reactor.is critical*
TS 3.13-3 2-1-72 operating conditions for the Charging Pump Cooling Water Subsystem shall be as follows:
- 1. Make-up water from the Component Cooling Water Subsystem shall be available.
- 2. One charging pump component cooling water pump and one charging pump service water pump shall be operating. The spare charging pump component cooling water pump and the spare charging pump service water pump shall be operable.
- 3. One charging pump intermediate seal cooler shall be operating and the spare charging pump intermediate seal cooler shall be operable.
- 4. During power operation the requirements of B-1, -2, and -3 above may be modified to allow one of the following components to be inoperable at any one time. If the system is not restored to meet the conditions of B-1 above within the time period specified below, the reactor shall be placed in the hot shutdown condition. If the system is not restored to meet the conditions of B-1, -2, and -3 within an additional 48 hours, the reactor shall be placed in the cold shutdown condition.
- a. One charging pump component cooling water pump or one
TS 3 .13-4 12-29-71 charging pump service water pump may be out of service provided the pump is restored to operable status within 24 hours.
- b. One charging pump intermediate seal cooler or other passive component may be out of service provided the system may still operate at 100 percent capacity and repairs are completed within 48 hours.
Basis The Component Cooling System is an intermediate cooling system which serves both reactor units. It transfers heat from heat exchangers containing reactor coolant, other radioact{ve liquids, and other fluids to the Service Water System. The Component Cooling System is designed to (1) provide cooling water for the removal of residual and sensible heat from the Reactor Coolant System during shutdown, cooldown, and startup, (2) cool the containment recirculation air coolers and the reactor coolant pump motor coolers, (3) cool the letdown flow in the Chemical and Volume Control System during power operation, and during residual heat removal for continued purification, ('4) cool the reactor coolant pump seal water return flow, (5) provide cooling water for the neutron shield tank and (6) provide cooling to dissipate heat from other reactor unit componenta. The Component Cooling Water Subsystem has four component cooling water pumps - and four component cooling water heat exchangers. Each of the component cooling water heat exchangers is designed to remove during normal operation the entire
TS 3.13-5 12-29-71 heat load from one unit plus one half of the heat load common to both units. Thus, one component cooling water pump and one component cooling water heat exchanger are required for each unit which is at power operation. Two pumps and two heat exchangers are normally operated during the removal of residual and sensible heat from one unit during cooldown. Failure of a single component may extend the time required for cooldown but does not effect the safe operation of the station. A Charging Pump Cooling Water Subsystem is provided for each reactor unit. The subsystem has two charging pump component cooling water pumps, two charging pump service water pumps, two charging pump intermediate seal coolers, a surge tank, and interconnecting valves and piping to provide cooling water to the charging pump lubricating oil coolers and charging pump mechanical seal coolers. Each of the charging pump component cooling water pumps, service water pumps, and intermediate seal coolers has full capacity, providing 100 percent_redundancy for this sybsystem. References FSAR Section 5.3 Containment Systems FSAR Section 9.4 Component Cooling System FSAR Section 15.5.1.2 Containment Design Criteria
TS 3.14-1 I-~ 12-29-71 3.14 CIRCULATING AND SERVICE WATER SYSTEMS Applicability Applies to the operational status of the Circulating and Service Water Systems. Objective To define those limiting conditions of the Circulating and Service Water Systems necessary to assure safe station operation .
- Specification A. The Reactor Coolant System temperature or pressure of a reactor unit shall not exceed 350°F or 450 psig, respectively, or the reactor shall
.not be critical unless: I.
- 1. The high level intake canal is filled to at least El.+ 18.0 ft at the high level intake structure.
- 2. Unit subsystems, including piping and valves, shall be operable to the extent of being able to establish the following:
- a. Flow to and from one bearing cooling water heat exchanger.
TS 3.14-2 2-1-72
- b. Flow to and from the comronent cooling heat exchangers required by Specification 3.13.
- 3. At least two circulating water pumps are operating or are operable.
- 4. At least two emergency service water pumps are operable; these two pumps will service both units simultaneously.
- 5. Two service water flowpaths to the charging pump service water subsystem are operable.
- 6. Two service water flowpaths to the recirculation spray subsystems are operable.
B. There shall be an operating service water flow path to and from one operating control area air conditioning condenser and at least one operable service water flow path to and from at least one operable control area air conditioning condenser whenever fuel is loaded in reactor core. C. The requirements of Specifications A-5 and A-6 may be modified to allow unit operation with only one operable flow path to the charging pump service water subsystem and to the recirculation spray subsystems. If the affected systems are not restored to the requirements of Specifications A-5 and A-6 within 24 hours, the reactor shall be placed in a hot shutdown condition. If the requirements of Specifications A-5 and A-6 are not me*t within an additional 48 hours, the reactor shall be
TS 3.14-3 2-1-72 placed in a cold shutdown condition. Basis The Circulating and Service Water System are designed for the removal of heat resulting from the operation of various systems and components of either or both of the units. Untreated water, supplied from the James River and stored in the high level intake canal is circulated by gravity through the recircu-lation spray coolers and the bearing cooling water heat exchangers and to the charging pumps lubricating oil cooler service water pumps which supply service water to the charging pump lube oil coolers. In addition, the Circulating and Service Water Systems supply cooling water to the component cooling water heat exchangers and to the control area air conditioning condensers. The component cooling water heat exchangers are not required during loss-of-coolant accident conditions for unit safety; however, they are normally required for power operation and during a loss-of-station power accident. At least one operating and one operable control area air conditioning condenser is required whenever monitoring of reactor conditions is required. The long term service water requirement for the loss-of-coolant accident in one unit with simultaneous loss-of-station power and second unit being main-tained in a safe condition is 15,000 gpm. Three diesel driven emergency service pumps with a design capacity of 15,000 gpm each, are provided to supply water to the high level intake canal during a loss-of-station power incident.
TS 3.14-4 2-1-72 Thus, one operating emergency service water pump is required for both units plus one operable spare, and the third pump can be down for maintenance. A minimum level of El. +18 ft in the high level intake canal is required to provide design flow of service water through the recirculation spray coolers during a loss-of-coolant accident. If the water level falls below El. +18 ft, the turbines will be automatically tripped and the reactors will be automatically tripped as a result of the turbines being tripped.
References:
FSAR Section 9.9 Service Water System FSAR Section 10.3.4 Circulating Water System FSAR Section 14.5 Loss-of-Coolant Accidents, Including the Design Basis Accident
TS 3.15-1 12-29-71 3.15 CONTAINMENT VACUUM SYSTEM Applicability Applies to the operational status of the Containment Vacuum System. Objective To define those conditions of the Containment Vacuum System necessary to assure safe station operation. Specification The unit Reactor Coolant System shall not be made to exceed a temperature or pressure greater tha~ 350° For 450 psig, respectively, or the reactor shall not be ccitical unless the following unit Containment Vacuum System conditions I are met: A~ Following a period when the containment was at atmospheric pressure and containment integrity has been achieved, the Containment Vacuum *System steam ejector and its associated piping and valves shall have been operating and have reduced the containment internal pressure to the subatmospheric operating pressure corresponding to an air partial pressure of between 9.0 and 11.0 psia as established in Specification 3. 8. e
TS 3 .15-2 12-29-71 B. One mechanical vacuum pump and one associated flow path shall be operable. C, The steam ejector and its associated piping and valves shall be secured and isolated once the containment internal pressure is at the subatmospheric pressure used for normal operation. D. The vacuum pump shall not be operated if the air partial pressure rises 0.25 psi above the preset value of the air partial pressure as determined by Technical Specification 3.8. Basis The Containment Vacuum System consists of q. steam ejector and two mechanical vacuum pumps with the required piping, valves, and instrumentation. It is designed to perform the following functions: A. Evacuation of the containment from atmospheric pressure to the subatmospheric pressure used for normal operation. B. Removal of air from the containment to compensate for containment inleakage during normal operation The system, through the use of the steam ejector, is designed to reduce the containment pressure from atmospheric press~re to the subatmospheric pressure used for normal operation in approximately 4 hr, compatible with the unit startup schedule. Each of two mechanical vacuum pumps has the capacity re-quired to.maintain the normal subatmospheric operating pressure. However,
TS 3~15-3 12-29-71 - due to the low leakage characteristics of the containment, neither pump will operate for long periods of time. The vacuum pumps are capable of being operated from the emergency buses and discharge to the gaseous waste disposal system. Technical Specification 3.8.B requires that the reactor be brought to the hot shutdown condition if the internal air partial pressure rises 0.25 psi above the preset value of the air partial pressure. Accordingly, the vacuum pump shall not be operated if this condition exists to insure that, in the event of an accident, there will be no release through the vacuum system. If containment vacuum is not established, the requirement that Reactor Coolant System temperature and pressure be no greater than 350°F and 450 psig,
.respectively, and that the reactor not be critical, will insure that, if a loss-of-coolant accident does occur, no significant pressure buildup in the containment would occur.
References FSAR Section 6.3.2 Containment Vacuum System Technical Specification 3.7 Containment
TS 3.16-1 12-29-71 3.16 EMERGENCY POWER SYSTEM Applicability Applies to the availability of electrical power for safe operation of the station during an emergency. Objective To define those conditions of electrical power availability necessary to shutdown the reactor safely, and provide for the continuing availability of Engineered Safeguards when normal power is not available. Specification A, A reactor shall not be made critical nor shall a unit be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350°F, respectively, without:
- 1. Two diesel generators (the unit diesel generator and the shared backup diesel generator) operable with each g~nerator's day tank having at least 290 gallons of fuel and with a minimum on-site supply of 35,000 gal of fuel available.
- 2. Two 4,160 v emergency buses energized.
- 3. Two 480 v emergency buses energized *.
TS 3.16-2 2-1-72
- 4. Two available sources of 4,160 v normal shutdown power from outside the station for the emergency buses.
- 5. Two operable flow paths for providing fuel to each diesel generator.
- 6. Two batteries, two chargers, and the d.c. distribution systems operable.
B, During power operation or the return to power from hot shutdown conditions 1 the requirements of specification 3.16-A may be modified to allow one of the fqllowing to be unavailable or inoperable.
- 1. One diesel generator and its associated fuel oil pumps and flow paths provided the operability of the other diesel generator and its associated fuel oil pumps and flow paths is demonstrated daily. If this diesel generator is not returned to an operable status within 7 days, the reactor shall be brought to a cold shutdown condition.
- 2. One of the normal shutdown power sources from outside the station.
The sources of normal shutdown power are the reserve station trans-formers which are supplied from the high voltage substation.
- 3. One battery may be inoperable for 24 hours provided the other battery and battery chargers remain operable with one battery charger carrying the d.c. load of the failed battery's supply system. If the
TS 3.16-3 2-1-72 battery is not returned to operable status within the 24 hour period the reactor shall be placed in the hot shutdown condition. If the battery is not restored to operable status within an additional 48 hours, the reactor shall be placed in the cold shutdown condition. C. The continuous running electrical load supplied by an emergency diesel generator shall be limited to 2750 kw. Basis The Emergency Power System is an on-site, independent, automatically starting power source. It supplies power to vital unit auxiliaries if a normal power source is not available. The Emergency Power System consists of three diesel generators for two units. One generator is used exclusively for Unit 1, the second for Unit 2, and the third generator functions as a backup for either Unit 1 or 2. The diesel generators have a continuous and 2,000 hour rating of 2750 kw and a two hour rating of 2850. The actual loads using conservative ratings for accident conditions, require approximately 2,320 kw. Each unit has two emergency buses, one bus in each unit is connected to its exclusive diesel generator. The second bus in each unit will be connected to the backup diesel generator as required. Each diesel generator has 100 percent capacity and is connected to independent 4,160 v emergency buses. These emergency buses are
TS 3.16-4 12-29-71 normally fed from two independent reserve 4,160 v systems, with the diesel generators functioning as an on-site backup system. Each emergency bus provides power to the following operating Engineered Safe-guards equipment:' A. O~e containment spray pump B. One charging pump C. One lbw head safety injection pump D. One recirculation spray pump inside containment E. One recirculation spray pump outside containment F. One containment vacuum pump G. One motor control center for valves, instruments, control air compressor, fuel oil pumps, etc. H. Control area air conditioning requipment - four air recirculating units, one water chilling unit, one service water pump and one chilled water circulating pump e* I. One charging pump service water pump for charging pump intermediate seal coolers and lube oil coolers. J. One charging pump cooling water pump for charging pump se'al coolers.
TS 3.16-5 12-29-71 The day tanks are filled by transferring fuel from any one of two buried tornado missile protected fuel oil storage tanks, each of 20,000 gal capacity. Two 100 percent capacity fuel oil transfer pumps _per diesel generator are powered from the emergency buses to assure that an operating diesel generator has a continuous supply of fuel. The buried fuel oil storage tanks contain a seven day supply of fuel, 35,000 gal minimum, for the full load operation of one diesel generator; in addition, there is an above ground fuel oil storage tank on-site with a capacity of 410,000 gal which is used for trans-ferring fuel to the buried tanks. If a loss of normal power is not accompanied by a loss-of-coolant accident, the safeguards equipment will not be required. ~nder this condition the following additional auxiliary equipment may be operated from each emergency bus: A. One component cooling pump B.* One residual heat removal pump C. One motor-driven auxiliary steam generator feedwater pump The emergency buses in each unit are capable of being interconnected under strict administrative procedures so that the equipment which would normally be operated by one of the diesels could be operated by the other diesel, if required.
TS 3.16-6 12-29-71 References FSAR Section 8,5 Emergency Power System FSAR Section 9.3 Residual Heat Removal System FSAR Section 9.4 Component Cooling System FSAR Section 10.3.2 Auxiliary Steam System
TS 3.17-1 12-29-71 3.17 LOOP STOP VALVE OPERATION Applicability Applies to the operation of the Loop Stop Valves. Objective To specify those limiting conditions for operation of the Loop Stop Valves which must be met to ensure safe reactor operation. Specifications
- 1. Whenever a reactor coolant loop is isolated, the boron con-centration in the isolated loop shall be maintained at a value gre~ter than or equal to the boron concentration in the active loops. The boron concentration in an isolated loop shall be measured and logged at least 5 days per week.
- 2. Whenever startup of an isolated reactor coolant' loop is initiated, the following conditions sh~i1 be met:
- a. All the channels, including redundant channels, of the Loop Stop Valve Interlock System of the isolated loop are operable.
In the event this condition is not satisfied, the loop must remain isolated.
TS 3.17-2 12-29-71 b, The unit shall be in a shutdown condition prior to opening either stop valve and throughout the timing interval required prior to opening the cold leg stop valve. c, Prior to opening the hot leg valve and again prior to opening the cold leg valve, the boron concentration of the isolated loop must be verified as greater than or equal to the boron concentration in the operating loops.
- d. The count rate, as given by the nuclear instrumentation shall be logged every five minutes during the timing interval required prior to opening the cold leg stop valve. Should the count rate increase by more than a factor of two over the initial count rate, the hot leg stop valve should be re-closed a~d no attempt made to open the stop valves until the reason for the count rate increase has been determined,
- e. All three reactor. coolant pumps must be running during the time a hot leg stop valve is being opened, when a hot leg stop valve is open and the cold leg stop valve in the same loop is closed, and during the time a cold leg stop valve is being opened; except when the reactor coolant temperature and pressure are equal to or less than 350°F and 450 psig, respectively.
Basis The above specified precautions and the design of the Loop Stop Valve Interlock (1) System provide assurance that no accidental reactivity addition to the core
TS 3.17-3 12-29-71 could occur during the startup operation of an inactive coolant loop. The Loop Stop Valve Interlock System will eliminate the possibility of adding reactivity to the core at any significant rate by preventing the cold leg valve opening unless a mixing flow between the isolated loop and the remainder of the Reactor Coolant System has existed for 90 minutes, and the temperatures in the cold and hot leg of the isolated loop are respectively within 20°F of the highest cold and hot leg temperatures of the other loops. The boron concentration in the isolated loop is maintained continuously at least equal to that in the active loops. The verification of the boron con-centration in the inactive loop prior to opening the hot leg and the cold leg valves provides a re-assurance of the adequacy of -the boron concentration. The continuous monitoring of the neutron flux by the nuclear instrumentation will immediately indicate a rapid change in the reactivity status of the core during the mixing flow phase. A slow reactivity change will be noticeable by comparing the count rates which will be logged every five minutes with the initial count rate. The fully withdrawn shutdown rods maintain at any time the capability to shut the unit down if this is required. Reference (1) FSAR Section 4.2
TS 3.18-1 12-29-71 3.18 MOVABLE IN-CORE INSTRUMENTATION Applicability Applies to the operability of the movable detector instrumentation system. Objective To specify functional requirements on the use of the in-core instrumentation systems, for the recalibration of the excore symmetrical off-set detection system. Specification A. A minimum of 16 total accessible thi~bles and at least 2 per quadrant 1 each of which.will cont~in a movable incore-detector,shall be operable during re-calibration of the excore synunetrical off-set detection system. B. Power shall be limited to 90% of .rated power for three loop operation, 54% of rated power for two loop operation with the loop stop valves closed, and 50% of rated power for two loop operation with the loop stop valves open if re-calibration requirements for the excore symmetrical off-set detection system, identified in Table 4.1-1, are not met.
TS 3.18-2 5-1-71 Basis The Movable In-core Instrumentation System (l)has five drives, five detectors, and 50 thimbles in the core. Each detector can be routed to twenty or more thimbles. Consequently, the full system has a great deal more capability than would be needed for the calibration of the excore detectors. To calibrate the excore detectors system, it is only necessary that the Movable In-core System be used to determine the gross power distribution in the core as indicated by the power balance between the top and bottom halves of the core . After the excore system is calibrated initially, recalibration is needed only infrequently to compensate for changes in the core, due for example to fuel depletion, and for changes in the detectors. If the recalibration is not performed, the mandated power reduction assures safe operation of the reactor since it will compensate for an error of 10% in the excore protection system. Experience at Beznau No. 1 and R. E. Ginna plants has shown that drift due to the core on instrument channels is very slight. Thus limiting the operating levels to 90% of the rated two and three loop powers is very conservative for both operational modes. Reference
- (1) FSAR - Section 7.6
TS 4. 0-1 5-1-71
- 4.0 SURVEILLANCE REQUIREMENTS Surveillance requirements provide for testing, calibrating, or inspecting those systems or components which are required to assure that operation of the units or the station will be as prescribeq in the preceding sections.
Specified time intervals may be adjusted plus or minus 25 percent to accommodate normal test schedules. References FSAR Section 12.9 Inservice Inspection FSAR Section 14.1 Operational Safety Review
TS 4.1-1 2-1-72 4.1 OPERATIONAL SAFETY REVIEW App li cab ili ty Applies to items directly related to safety limits and limiting conditions for operation. Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions. Specification A, Calibration, testing, and checking of instrumentation channels shall be performed as detailed in Table 4.1-1. B. Equipment tests shall be conducted as detailed in Table 4.l-2A. C. Sampling tests shall be conducted as detailed in Table 4.2-2B. D. Whenever containment integrity is not required, only the asterisked items in Table 4.1-1 and 4.1-2 are applicable. E, Flushing of sensitized stainless steel pipe sections shall be conducted as detailed in Table 4.1-3.
TS 4.1-2 5-1-71 Basis Check Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in "upscale" or "downscale" indication can be .easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action, and a periodic check supplements this type of built-in surveillance. Based on experience in operation of both conventional and nuclear unit systems,
- when the unit is in operation, the minimum checking frequencies set forth are deemed adequate for reactor and steam system instrumentation.
Calibration Calibration shall be performed to ensure the presentation and acquisition of accurate information. The nuclear flux (power level) channels shall be calibrated daily .against a heat balance standard to account for errors induced by changing rod patterns and core physics parameters. Other channels are subject only to the "drift" errors induced within the
- instrumentation itself and, consequently, can tolerate longer intervals between calibration, Process systems instrumentation errors resulting from drift within the individual instruments are normally negligible.
TS 4.1-3 5-1-71
- During the interval between periodic channel tests and daily check of each channel, a comparison between redundant channels will reveal any abnormal condition resulting from a calibration shift, due to instrument drift of a single channel.
During the periodic channel test, if it is deemed necessary, the channel may be tuned to compensate for the calibration shift. However, it is not expected that this will be required at any fixed or frequ~nt interval. Thus, minimum calibration frequencies of once-per-day for the nuclear flux (power level) channels, and once each refueling shutdown for the process system channels is considered acceptable
- Testing The minimum testing frequency for those instrument channels connected to the safety system is based on an average unsafe failure rate of 2.5 x 10- 6 failure/hr per channel. This is based on operating experience at conventional and nuclear units. An unsafe failure is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or when it attempts to respond to a proper signal.
For the specified one month test interval, the average unprotected time is 360 hrs in case of a failure occurring between test intervals. Thus, the 6 probability of failure of one channel between test intervals is 360 x 2.5 x 10-or .9 x 10- 3
- Since two channels.must fail in order to negate the safety function, the probability of simultaneous failure of two-out-of-three channels
TS 4.1-4 2-1-72 6 is 3(.9 x 10- 3 ) 2 = 2.4 x 10- . This represents the fraction of time in which each three-channel system would have one operable and two inoperable channels
-6 and equals 2.4 x 10 x 8760 hours per year, or (approximately) 1 minute/year.
It must also be noted that to thoroughly and correctly test a channel, the channel components must be made to respond in the same manner and to the same type of input as they would be expected to respond to during their normal operation. This, of necessity, requires that during the test the channel be made inoperable for a short period of time. This factor must be, and has been, taken into consideration in determining testing frequencies. Because of their greater degree of redundancy, the 2/4 logic arrays provide an even greater measure of protection and are thereby acceptable for the same testing interval. Those items specified for monthly testing are associated with process components where other means of verification provide additional assurance that the channel is operable, thereby requiring less frequent testing. Flushing During construction of the facility, stress relieving of some of the cold bent type 316 stainless steel piping, resulted in its becoming sensitized to potential .stress corrosion cracking under certain conditions e.g. low pH in conjunction with high chlorides. The systems containing sensitized pipe which remain wet during normal operation and have no flow, e.g. safety injection, may contribute to the buildup of those contaminants which could cause accelerated corrosive
TS 4.1-A 2-1-72 -* attack of the pipe. In order to insure the continued integrity of the pipe throughout plant life, the affected lines are flushed periodically to remove stagnant water which may contain contaminants. The flushing requirements delineated in TS Table 4.1-3 insure that a buildup of contaminants will not occur. The specified minimum flush durations, with expected flow rates during flushing, insures that a volume of water greater than the volume contained in the stagnant flow paths listed in Table 4.1-3 will be flushed.
- TABLE 4.1-1 e
MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND
. TESTS OF INSTRUMENT CHANNELS Channel Description Check Calibrate Test Remarks
- 1. Nuclear Power Range s D (1) BW (2) 1) Against a heat balance standard M (3) 2) Signal to 6T; bistable action AP (3) (permissive, rod stop, trips)
- 3) Upper and lower chambers for symetric offset by means of the moveable incore detector system.
- 2. Nuclear* Intermediate Range .*S (1) N.A. p (2) 1) Once/shift when in service
- 2) Log level; bistable action (permissive, rod stop, trip)
- 3. Nuclear Source Range . *S (1) N.A. p (2) 1) Once/shift when in service
- 2) Bistable action (alarm, trip)
- 4. Reactor Coolant Temperature *S R BW (1) 1) Overtemperature-6T BW (2) '2) Overpower - 6T I
- 5. Reactor Coolant Flow s R M
- 6. Pressurizer Water Level s R M
- 7. Pressurizer Pressure(High and low) s R M
- 8. 4 Kv Voltage & Frequency s R M Reactor protection circuits only
- 9. Analog Rod Position *S (1,2) R M (3) 1) With step counters (4) 2) Each six inches of rod motion f-3 t--' Cl) when data logger is out.of N I .i:,-.
service N *
\0 t--'
- 3) Rod bottom bistable action I I
'-I V,
- 4) NA When reactor is in cold shut- t--'
down
e TABLE 4.1~1 (Continued) Channel Description *check Calibrate Test Remarks
- 10. Rod Position Bank Counters S (1, 2) N .A. N.A. 1) Each six inches of rod motion when data logger is out of service
- 2) With analog rod position
- 11. Steam Generator Level S R M
- 12. Charging Flow N.A. R N.A.
- 13. Residual Heat Removal Pump Flow N .A. R N.A.
- 14. Boric Acid Tank Level *D R N.A.
- 15. Refueling Water Storage Tank Level W R N.A.
- 16. Boron Injection Tank Level W R N.A.
- 17. Volume Control Tank Level N.A. R N.A.
- 18. Reactor Containment Pressure-CLS *D R M (1) 1) Isolation Valve signal and spray signal
- 19. Process and Area Radiation Monitor-ing Systems *D R M
- 20. Boric Acid Control N.A. R N.A.
- 21. Containment Sump Level N.A. R N.A.
- 22. Accumulator Level and Pressure s R N.A.
- 23. Containment Pressure-Vacuum Pump t-3 System s R N.A. I-' C/.l N
I N ~
- 24. Steam Line Pressure s R M \0 I I-'
-..J I I-' 0\
e TABLE 4.1-1 (Continued) Channel Description Check Calibrate Test Remarks
- 25. Turbine First Stage Pressure s R M
- 26. Emergency Plan Radiation Instrtllllents *M R M
- 27. Environmental Radiation Monitors *M N.A. N.A. TLD Dosimeters
- 28. Logic Channel Testing N.A. N.A. M
- 29. Turbine Overspeed Protection N.A. R R Trip Channel (Electrical)
- 30. Turbine Trip Set Point N.A. R R Stop valve closure or low EH fluid pressure r.
- 31. Seismic Instrumentation M SA M S - Each Shift M - Monthly D Daily P - Prior to each startup if not done previous week W - Weekly R - Each Refueling Shutdown NA - Not applicable BW - Every two weeks SA - Semiannually AP - After each startup if not done previous week
- See Specification 4.lD NH I tr.i I-'
I
...i~
N* I-' I
. -...J
TS 4.1-8 12-29-71 TABLE 4.l-2A. MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR Section Description Test Frequency Reference
- 1. Control Rod Assemblies Rod drop times of Each refueling 7 all full length rods shutdown or after at hot and cold con- disassembly or ditions. maintenance requiring the breech of the Reactor Coolant System integrity
- 2. Control Rod Assemblies Partial movement Every 2 weeks 7 of all rods
- 3. Refueling Water Chemical Functional Each refueling 6 Addition Tank shutdown
- 4. Pressurizer Safety Setpoint Each refueling 4 Valves shutdown e 5. Main Steam Safety Valves Setpoint Each refueling shutdown 10
- 6. Containment Isolation *Functional Each refueling 5 Trip shutdown
- 7. Refueling System *Functional Prior to refueling 9.12 Interlocks
- 8. Service Water System *Functional Each refueling 9.9 shutdown
- 9. Fire Protection* Pump Functional Monthly 9.10 and Power Supply
- 10. Primary System Leakage
- Evaluate Daily 4
- 11. Diesel Fuel Supply *Fuel Inventory 5 days /week 8.5
- 12. Boric Acid Piping
- Operational Monthly 9.1 Heat Tracing Circuits
- 13. Main Steam Line Trip Functional Monthly 10 Valves
~ 14. Service Water System Functional Each Refueling 9.9 Values in Line Supplying Recirculation Spray Heat Exchangers
- See Specification 4.1.D
TS 4.1-9 2-1-72 TABLE 4.l-2B MINIMUM FREQUENCIES FOR SAMPLING TESTS FSAR Section Description Test Frequency Reference
- 1. Reactor Coolant Liquid Radio-chemical Monthly Samples Analysis (1)
*Gross Activity (2) 5 days/week 9.1
*Tritium Activity Weekly 9.1
*Chemistry (Cl, F&0 ) 5 days/week 4
*Boron Concentration2 Twice/week 9.1 E Determination Semiannually (3)
- 2. Refueling Water Storage Boron Concentration Weekly .6 Tank Water Sample
- 3. Boric Acid Tanks *Boron Concentration Twice/week 9.1
- 4. Boron Injection Tank Boron Concentration Twice/week 6
- 5. Chemical Additive NaOH Concentration Monthly 6 Tank
- 6. Spent Fuel Pit *Boron Concentration Monthly 9.5 7, Secondary Coolant Fifteen minute degassed Weekly 10.3 (3 and'/ activity (4)
- 8. Stack Gas Iodine and
- I-131 and particulate Weekly Particulate Samples radioactive releases(5)
- 9. Accumulator Boron Concentration Monthly 6.2
- See Specification 4.1.D (1) A radiochemical analysis will be made to evaluate the following corrosion products:
Cr 51, Fe 59, Mn 54, Co 58, and Co 60. (2) A gross beta-gamma degassed activity analysis shall consist of the quantitative measurement of the total radioactivity of the primary coolant in units of µCi/cc. (3) E determination will be started when the gross gamma degassed activity of radionucl:ides with half-lives greater than 30 minutes* analysis indicates > lOµCi/cc. \ (4) If the fifteen minute degassed beta and gamma activity is 10% of that given in - (5) Specification 3.6.C, an 1-131 analysis will be performed. If the activity of the samples is 10% or greater of that given in Specification 3.11.B.1. the frequency shall be increased to daily.
TABLE 4.1-3 MINIMUM FREQUENCIES FOR FLUSHING SENSITIZED PIPE FLUSH FLOW PATH MINIMUM FLUSH GENERAL DESCRIPTION DURATION FREQUENCY REMARKS
- 1. From C.S. Pump CS-P-lA 15 minutes Monthly Run in conjunction with to M.O. Isolation Valves or immediately after pump test required by Specification 4.5.A.l
- 2. From C.S. Pump CS-P-lB 20 minutes Monthly Run in conjunction with to M.O. Isolation Valves or immediately after pump test required by Specification 4.5.A.l
- 3. From L.H.S.I.Pump, SI-P-lA, 60 minutes Monthly Run in conjunction with or Discharge Line to MDV 863A immediately after pump test required by ~pecifica-tion 4.11.B.1
- 4. s. I. line, from charging pump Flushes to be performed discharge loop fill header to only when R.C. System containment missile barrier, for pressure is greater than flow to 1500 psig.
- a. R.C. hot leg loop 1 15 minutes Monthly
- b. R.C. hot leg loop 2 10 minutes Monthly
- c. R.C. hot leg loop 3 15 minutes Monthly
- 5. s. I. line, from charging pump discharge Flushes to be performed header to containment missile barrier, only when R.C. System for flow to pressure is greater than
- a. R.C. cold leg loop 1 5 minutes Monthly 1500 psig.
- b. R.C. cold leg loop 2 5 minutes Monthly
- c. R.C. cold leg loop 3 5 minutes Monthly N~
I
- f-J f-J I I "N f-J 0
L_ _ _
TS ,,.:2.-J. I 0-15-70
- 4.2 1
REACTOR COOLANT SYSTEM COMPONENT I NSPECTI0NS Applicability Applies to in-service inspection of the reactor vessel and the reactor coolant system pressure boundary. Objectives To provide assurance of the continued integrity of the reactor vessel and reactor coolant system press.ure boundary.
- Specifications A. Prior to initial unit operation a survey usip.g ultrasonic, visual, and surface techniques shall be made to establish pre-operational system integrity and establish baseline data.
B. Following initial unit operation, nondestructive inspections shall be performed as specified in Table 4. 2-1. The results obtained from these inspections shall be evaluated after five years of commercial operation and reviewed in light of the technology available at that time. C. The normal inspection interval is 10 years .
TS 4.2-2 3-15-71
- D, Detailed records of each inspection shall be maintained to allow a continuing evaluation and comparison with future inspections.
Bases The in-service inspection program is based on Section XI of the ASME Code for Inservice Inspection of Nuclear Reactor Coolant Systems dated January, 1970, Winter 1970 Addenda, (and the rules and guidelines given in the AEC document, "In-service Inspection Requirements for Nuclear Power Plants Constructed with Limited Accessibility for In-service Inspection"). Since each unit was designed and partially constructed without the benefit of the ASME In-Service Inspection Code, 100-percent compliance may not be feasible
- or practical. However, the inspection program was developed by adopting, insofar as practicable, the principles and intent embodied in the Code. It must be recognized that equipment and techniques to perform the inspection are still in development, and to that extent are speculative. In most areas scheduled for test, a detailed pre-service mapping will be conducted using techniques which are believed to be appropriate.for the later operational inspections. The areas specified for inspection are representative of those experiencing relatively high strain and therefore will serve to indicate potential problems before significant flaws could develop, As more experience is gained in operation of pressurized-water reactors, the recommended time schedule and location of inspection might be altered, or should new techniques be developed, consideration will be given to incorporate these into this inspection program
- TS 4.2-3 3-15-71
- The program defines the examinations to be performed within the first 5 years of unit operation. The inspection sampling and frequency during this period shall be governed by the requirements as specified in the ASME Code for the first third of the 10 year inspection interval. A tentative 10 year program is also included, although the operating experience and results of the first five years will constitute the bases for re-evaluation to take into account changes in technology, equipment, and the Code. This program complies with the intent of the Code but has some variations to take into account unit design as discussed under each inspection category.
The techniques anticipated for in~service inspection include visual inspections, ultrasonics, radiographic, magnetic particle, and liquid penetrant testing of selected parts during refueling periods and necessary maintenance outages
- The primary pressure boundary covered by this inspection will include the primary reactor coolant system and branch lines 2 inches in diameter and greater. However it should be noted that excluded from volumetric examination will be that piping 4 inches in diameter and smaller and on which no meaningful f volumetric examination results can*be obtained. The inspection programs for the reactor vessel, the pressurizer, the primary side of the steam generators, piping, pumps, and valves are outlined in Table 4.2-1. Each component system is discussed in categories corresponding to Section XI of the ASME Code
- TS
- The following discussions in each category cover the general type of exami-1 nation planned in Table 4.2-1 and the technical reasons for any exceptions to the Code.
A. Reactor Pressure Vessel Category A - Pressure Containing Welds in Reactor Vessel Belt-Line Region Depending upon development of appropriate equipment, it is intended that these welds be volumetrically examined, when required, using automated ultrasonic techniques. The mechanized device could be used only when the core and lower internals have been removed from the vessel. For this reason, examination may be made at or near the end of the 10-year interval or when the core and lower internals are removed at an earlier date for other reasons .
- Category B - Pressure Containing Welds in Remainder of Vessel Dependent upon development of appropriate equipment, the same mechanized ultrasonic techniques used for examination of the welds in the core region would be used to examine the other longitudinal and circumferential welds in the barrel section of the reactor pressure vessel and would be performed at the same time and manner as the welds in Category A. There are no meridional seam welds.
Category C - Pressure Containing Welds, Vessel-to-Flange and Head-to-Flange The head flange weld can be examined using either mechanized or manual ultrasonic techniques,. Mechanized techniques are preferred due to the repeatability of such techniques and the ability to record the data
- TS 4.2-5 10-15-70
- The vessel flange weld can be examined using mechanized ultrasonic, techniques.
This examination would also include the ligaments in the flange between the bolt holes. Category D - Pressure Containing Nozzles in Vessels Depending upon development of appropriate equipment, it is intenped that these welds be volumetrically examined from the ID using automated ultrasonic techniques. Examination of the coolant outlet nozzle-to-shell welds would be performed from the ID without removal of the core barrel. This examination would also include the integral extension of the nozzle inside the vessel and the inner ...
- nozzle radii.
Examination of the coolant inlet nozzles-to-shell welds from the ID requires removal of the core and lower internals. Therefore, these nozzle welds, including the inner radii, will be examined at the same time as the welds in Category A. Category E Pressure Containing Welds in Vessel Penetrations The only penetrations in this category are the control rod drive housings in the upper head and the in-core instrumentation penetrations in the lower head. The control rod drive tubes are welded to the upper head with a partial penetration weld. A thermal sleeve is permanently affixed to the inside of the 0 stub tube prohibiting volumetric inspection from the ID. Geometry and the partial
- penetration weld do not allow a meaningful inspection from the O.D.
welds would be visually examined as specified in Category E-2. These
TS 4.l-b 12-29-71 Volumetric inspection of the in-core instrumentation penetrations is dependent upon development of acceptable equipment and techniques. Consequently, these welds will be examined as specified in Category E-2, Category E Pressure Containing Welds in Vessel Penetrations A visual examination for evidence of leakage would b~ made of the penetrations in the upper head at the ti.me of the system hydrostatic tests as required by section IS-520 of the Code. The penetrations in the lower head will also be examined for leakage during this test if volumetric inspection techniques are not developed and if radiation levels permit access. Category F - Pressure Containing Dissimilar Metal Welds Depending upon development of appropriate equipment, it is intended that these welds be volumetrically examined from the ID using automated ultrasonic techniques. Experience with other units has shown that, in general, these welds can be volumetrically examined with ultrasonic techniques. The feasibility of this examination would be determined on the pre-operational examination, *and this will determine the acceptable method to.use in the in~ervice inspection. Examination of the coolant outlet nozzle dissimilar metal weld (pipe-to-nozzle weld) would be performed from the ID without removal of the core barrel. Examination of the coolant inlet nozzle dissimilar metal weld would require removal of the core and lower internals. Therefore these welds would be examined at the same time as the welds in Category A. Exception is taken to performing a visual and surface examination from the O.D. on these welds due to anticipated radiation levels and physical access problems.
TS 4.2-7 10-15-70
- Category G - Pressure Retaining Bolting All of the reactor vessel studs are scheduled to be removed at each refueling cycle. They are, thus, available for a volumetric, surface, and visual examination as may be required. A visual examination and an ultrasonic volumetric examination would be made of each stud on an examination schedule as shown in Table 4.2-1 for the 5-year program and the tentative 10-year program.
The c~asure stud nuts woulj- be examined with techniques similar to those for the studs, and the washers would be visually inspected only. The ligaments between the bolt holes of the reactor pr~ssure vessel would be volumetrically examined at the same time the vessel flange weld is examined
- Category H - Vessel External Supports The reactor pressure vessel is supported on pads, integrally welded to the coolant nozzles. In accordance with Category Hof Table IS-251 of the Code, this examination is covered by Category D, Category I - Vessel Interior Clad Surfaces Those areas of the vessel cladding normally accessible during refueling periods shall be visually examined during the inspection interval. However, whenever the core barrel and lower internals are removed, the patches in the vessel interior would be visually examined. The examination scheduled in the first 5-year period and tentatively scheduled for the 10-year period is given in Table 4.2-1.
TS 4.2-8 10-15-70
- Selected areas of the cladding in the upper closure head will be examined during normal refueling periods when the head is removed.
Category N - Interior Surfaces and Internal Components of Reactor Vessels It is proposed that examinations in this category be made by remote television or borescopic examination. A critical examination would be made at the first refueling to detect if any changes have occurred due to initial operation. The amount of examination to be performed at subsequent refueling outages would depend upon the results of the first examination and those made on comparable pressurized-water systems. B. Pressurizer Category B - Pressure Containing Welds in Vessels
- rhe primary heads on the pressurizer are of cast material and do not contain meridional welds. There are circumferential welds joining the heads to the barrel section and circumferential and longitudinal welds in the barrel section that require examination. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in Table 4.2-1.
Category D - Pressure Containing Nozzles in Vessels The nozzles are integrally cast into the head and there are, therefore, no welds in this category
- TS 4.2-9 3-15-71
- Category E - Pressure Containing Welds in Vessel The heater connections and instrument penetrations of the pressurizer meet the exclusion criteria of Section IS-121 of the Code, These penetrations shall be given a visual examination for evidence of leaking at the time of the system hydrostatic test as required by IS-520 of the Code.
Category F - Pressure Containing Dissimilar Metal Welds There are dissimilar metal welds on the nozzles of the pressurizer. Experience with other units has shown that, in general, these welds can be volumetrically examined with ultrasonic techniques, The feasibility of this examination would be determined on* the pre-operational examination, and this will determine the acceptable method to use in the in-service inspection. The examinations scheduled for the first 5-year period and tentatively scheduled for the 10-year period are given in Table 4.2-1. Category G - Pressure Retaining Bolting Bolting on the pressurizer below 2 inches.in diameter would require visual examination. Excluded from examination are* bolting of a single connection whose failure results in conditions that satisfy the exclusion criterion of IS-121 of the Code. The examinations scheduled for the 10-year period_ are given in Table 4.2-L Category H - Vessel External Supports There are no integrally welded vessel supports on the Surry pressurizer
- TS 4.2-10 10-15-70
- Category I - Vessel Interior Clad Surface A visual inspection of a 36-square-inch area of the clad surface of the pressurizer. would be made at such time as the pressurizer is opened for necessary maintenance. The examinations scheduled for the first 5-year period and tentatively scheduled for the 10-year period are given in Table 4.2-1 C. Steam Generators Category B - Pressure Containing Welds in Vessels The circumferential welds joining the heads to the tube sheet are the only welds that require examination'. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in Table
- 4. 2-1.
Category D - Pressure Containing Nozzles in Vessels The nozzles are integrally cast with the head and there are no nozzle-to-shell welds. The inner nozzle radii of the heads would be volumetrically examined from the ID when the generators are opened for necessary maintenance. Category F - Pressure Containing Dissimilar Metal Welds The dissimilar metal welds between the nozzles and the main coolant piping join cast carbon steel to cast stainless steel. Volu,metric inspection of these welds will be dependent upon meaningful examination techniques and results of the pre-service inspection
- TS 4.2-11 3-15-71
- Category G - Pressure Retaining Bolting There is bolting on the steam generators of less than 2 inches in diameter.
Excluded from examination are bolting of a single connection whose failure results in conditions that satisfy the exclusion criterion of IS-121 of the Gode. Category H - Vessel External Supports The steam generators are supported by pads integrally cast with the lower head. There are no welds that require examination, Category I - Vessel Interior Clad Surface A visual inspection of the clad surface on the primary head of the steam
- generators from the ID wi11 be made at such time* as the generators are opened for necessary maintenance. The examinations scheduled for the first 5-year period and tentatively scheduled for the 10-year period are given in Tabel 4, 2-1.
D. Piping Pressure Boundaries Category F - Pressure Containing Dissimilar Metal Welds The main coolant piping of this plant is stainless st*eel, as are the pump and valve nozzles. Thus, the only dissimilar metal welds in the main coolant pipe are the attachments between the coolant pipe and the vessels. There are a few other dissimilar metal welds in the piping system, such as piping connections to the pressurizer .
TS 4.2-12 10-15-70
- These inspections and the pipe-to-nozzle connections on the steam generator are specified in Table 4.2-1, along with inspection of the reactor vessel pipe-to-nozzle welds. The examinations scheduled for the first 5-year period and tentatively scheduled for the 10-year period are given in Table 4,2-1.
Category J - Pressure Containing Welds in Piping The circumferential welds in the main coolant piping, branch, and auxiliary piping systems, will be examined in compliance with the Code assuming access and radiation levels allow. Exception i~ taken for socket welds and the longitudinal welds in the cast elbows of the main coolant piping until development of acceptable voltnnetric inspection techniques *
- Category G - Pressure Retaining Bolting All bolting in the piping system is below 2 inches in diameter and wil,l be inspected in compliance with the Code.
Category K - Support Members and Structures for Piping Systems The piping systems that contain supports and restraints integrally welded to the pressure containing boundary will be inspected in compliance with the Code, Members and structures whose structural integrity is relied upon to withstand the design loads and seismic-induced displacements will be inspected as specified in Item 4.6 of Table 4.2-1, The examinations scheduled for the 5-year period and tentativeiy scheduled for the 10-year period are given in Table 4.2-1.
TS 4 2-13 10-15-70
- E. Pump Pressure Boundary Category F - Pressure Containing Dissimilar Metal Welds There are no nozzle-to-safe end welds between the main coolant piping and the pump suction and discharge nozzles.
Category G - Pressure Retaining Bolting There are bolts 2 inches and larger in diameter on the pumps. If the connection is not broken during the inspection interval, a visual exami-nation would be made and an ultrasonic examination would be made with the bolting under tension. If the bolting connection is broken for any reason, a visual and volumetric examinatlon of the bolting would be made and, in addition, a volumetric inspection of the ligaments of the base material
- would be made insofar as .is possible due to the cast structure. The examinations scheduled for the 5-year period and tentatively scheduled for the IO-year period are given in Table 4.2-1.
Category K -* Support Members and Structures for Pumps The support members for the pumps consist of a cast foot welded to the cast casing. These supports and the pump supporting structure would be visually examined since there are no acceptable techniques for ultrasonically inspecting these welds. Category L - Pressure Containing Welds in Pump Casing These welds were volumetrically examined using radiographic techniques during the pump manufacture. A similar radiographic technique is theoretically
TS 4.2-14 12-29-71
- possible when the pump is disassembled and the impeller has been removed.
However, it is not known if such radiography can be performed in practice due to pump design and the radiation level that will probably exist in this component as well as due to the interference of the swirl guide vanes. A study would be made to see if this is possible and, in addition, experience with other pressurized-water unit in-service inspections would be solicited to determine the feasibility. Thus, this examination is not scheduled, as is shown in Table 4.2-1. A visual examination of the available internal surfaces of the pump would be made at such time as the pump is disassembled for maintenance, Item 5,8 - Primary Pump Flywheel All flywheels shall be visually examined at the first and third refueling. An in-place ultrasonic voiumetric examination of the areas of higher stress concentration at the bore and keyway of each flywheel shall be performed at approximately three (3) year intervals, during the refueling or maintenance shutdown coinciding with the inservice inspection schedule as required by the ASME Boiler and Pressure Vessel Code Section XI, A surface examination. of all exposed surfaces and complete ultrasonic volumetric examination shall be conducted at approximately ten (.10) year intervals. Removal of the flywheel is not required to perform these examinations. F. Valve Pressure Boundaries Category F - Pressure Containing Dissimilar Metal Welds There are no dissimilar metal welds between the valves and the piping system in this facility,
-\
TS 4.2-15 3-15-71
- Category G - Pressure Retaining Bolting There is bolting 2 inches and larger in diameter on the main coolant stop valves. All other valve bolting is below 2 inches in size. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year per~od are given in Table 4.2-1.
Category M Pressure Containing Welds in Valve Bodies There are no valves in this facility with pressure containing valve body welds. Category M-2 *- Valve Bodies The internal pressure boundary surfaces of one disassembled valve (3 inches and over in size) in each category and type shall be visually examined during each inspection interval. If the valve shall be disassembled for maintenance any time during the inspection interval, it may be inspected at this time. Otherwise the examination may be performed at or near the end. of each in-spection interval. Cate&ory K Support M*embers and Structures for Valves There are no integrally-welded supports on the valves subject to this examination. Category K Supports and Hangers Any valve support whose structural integrity is relied upon to withstand the design loads and seismic-induced displacements would be examined. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in Table 4.2-L
TS 4.2-15a 2-1-72 G. Miscellaneous Inspections Item 7.1 - Materials Irradiation Surveillance Specimens The reactor vessel surveillance program includes eight specimen capsules to evaluate radiation damage based on pre-irradiation and post-irradiation testing of specimens. Capsule No. 1 shall be removed and examined at the first region refueling. Capsule No. 2 shall be removed and examined after five years. Capsule No. 3 shall be removed and examined after ten years. Capsule No. 4 shall be removed and examined after twenty years. Capsules No. 5-8.are spares for complementary or duplicate testing. H. Sensitized Stainless Steel Piping This piping is subject to augmented inspection to assure piping integrity. Item 8.1 Sensitized stainless steel piping which is part of section D, category J will be inspected at twice the frequency required by the code. Item 8.2 Sensitized stainless steel piping which is not subject to section XI of the ASME Code, will be undergo visual and surface examination. The containment and recirculation spray rings, which are located in the overhead of the containment, will be visually inspected after two years of station operation, and thereafter at the major inservice inspection shutdown scheduled near the end of each ten year interval. Additionally, sections of the piping will be examined by liquid penetrant inspection when the piping e** is visually inspected.
TS 4.2-15b 2-1-72 - All other piping included in item 8.2 will be visually inspected at least every two years. Sections of this piping will be examined by liquid penetrant inspection when the piping is visually inspected.
TABLE 4.2~*1 SECTION A. REACTOR VESSEL AND CLOSURE HEAD Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 1.1 A Longitudinal and circum- Volumetric 0% 5% of the length of the The required examina-ferential shell welds in circumferential welds; tions may be made at or core region 10% of the length of the near the end of the 10-longitudinal welds year inspection interval. When the longitudinal and circumferential weld have received an ex-oosure to neutron fluence in excess of 1019 nvt (En of 1 Mev or above), the length of the weld in the high fluence region to be examined shall be in-creased*to 50 per cent. 1.2 B Longitudinal and circum- Volumetric 0% 5% of the length of the The required examinations ferential welds in shell circumferential welds; 10% may be made at or near the (other than those of of the length of the end of the 10-year in-Category A and C) and longitudinal welds spection interval. meridional and circum-ferential seam welds in bottom head and closure head (other than those of Category C) 1.3 C Vessel-to-flange and head- Volumetric 33% of the length of 100% of the length of Both of these welds are to-flange circumferential the circumferential the circumferential welds available for examina-welds welds tion during normal re-fueling operations. 1.4 D Primary nozzle-to-vessel Volumetric 100% coverage of a 100% of the remaining The coolant inlet nozzles-welds and nozzle-to- coolant outlet nozzle- coolant outlet nozzle-to- to-shell welds and radius ~ vessel inside radiused to-shell weld and inner shell welds and inner sections will be ex- 6 section nozzle radii radius sections and 100% amined at the same time ~-:" of the coolant inlet 2s the welds in Category Y' 't nozzles-to-shell welds and A. c3 ~ radius sections
TABLE 4.2-1 SECTION A, REACTOR VESSEL Ai'l'D CLOSURE HEAD (Continued) Required Required Extent of Examination Tentative Inspec-Ite:n Examination Examination Planned During First tion During No. Category Areas* Methods 5-Year Interval 10-Year Interval Remarks
- 1. 5 E-1 Vessel penetrations, in- Volumetric 0% 25% No meaningful volumetric eluding control rod drive examinations can be per-penetrations and control* formed on the control rod rod housing pressure drive penetrations at this
- boundary welds time. Volumetric examination of the instrument penetrations in the bottom head will be performed dependent upon de-velopment of acceptable equip-ment.
1.6 E-2 Vessel penetrations, in- Visual* 0% 25% The examination will be a cluding control rod drive yisual examination for evidence penetrations and control of leaking at the time of the rod housing pressure system hydrostatic test at or boundary welds near the end of the inspection interval. *
- 1. 7 F Primary nozzles to safe* Visual and 100% of the dissimilar 100% of the re- The dissimilar metal welds of end welds surface and metal weld on one maining coolant out- each nozzle will be volumetri-volumetric coolant outlet nozzle let nozzle dissimilar cally examined at the same time will be volumetrically metal welds and 100% as the nozzle-to-shell weld, examined of the coolant inlet as specified in Category D.
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TABLE 4.,2-1 SECTION A. REACTOR VESSEL AND CLOSURE HEAD (Continued) Required Required Extent of Examination Tentative Inspec-Examination . Examination Planned During First tion During Item Remarks Catego!I Areas Methods 5-Year Interval 10-Year Interval ~ Volumetric 33% 100% Vessel studs and nuts 1.8 G-1 Closure studs and nuts would be inspected and visual or surface volumetrically and by surface techniques as required. The ligaments of the vessel flange would be examined at the same time as the flange welds, Volumetric 33% of the vessel-to- 100% of the vessel The ligaments will be 1.9 G-i Ligaments between threaded stud examined at the same time holes flange bolt ligaments flange bolt ligaments will be examined will be examined as the flange weld of Item No. 1.3. Closure washers, bushings Visual 33% 100% None 1.10 G-1 Pressure retaining_bolting Visual Not applicable Not *applicable There are no pressure 1,11 G-2 retaining bolts less than 2 inches on the Surry Reactor vessels, Integrally welded vessel Volumetric (See remarks) (See remarks) These welds are covered 1.12 K by the examinations of supports Category D,
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TABLE 4.2-1 SECTION A. REACTOR VESSEL AND CLOSURE HEAD (Continued) Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Catego!l Areas Methods 5-Year Interval 10-Year Interval Remarks 1.13 I-1 Closure head cladding Visual and 2 patches 6 patches During the 10-year period, surface or at least 6 patches (each volumetric 36 square inches) evenly distributed in the vessel head would be examined. 1.14 I-1 Vessel cladding Visual None 6 patches During the 10-year period, at least 6 patches (each 36 square inches) evenly distributed in the vessel head would be visually inspected. 1.15 N Interior surfaces and internals Visual A critical examination The inspections made at The examination will include and integrally welded internal will be made of the the 4th refueling cycle internal support attachments supports interior surfaces made will be repeated at the welded to the vessel whose available by normal 7th and 10th refueiing failure may adversely affect refueling operations at cycle core integrity provided the 1st refueling cycle, these are available for This will be repeated at visual. examination by the 4th refueling cycle components removed during with the amount of the normal refueling operations. inspection being dependent upon results of the 1st inspection and that made on other pressurized-water sysc;ems. I-'~ o. IN I-' I
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TABLE 4.2-1 SECTION B. PRESSURIZER Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During ~ Category Areas Methods 5-Year Interval 10-Year Interval Remarks 2.1 B Longitudinal and circumferential Visual and 5% of the length of the By the end of the 10-year None welds volumetric circumferential we~d period, 10% of the length joining the lower head of the longitudinal and to the barrel section 5% of the length of the and 10% of the length circumferential weld of the adjoining longi- would be inspected tudinal weld would be examined. 2.2 D Nozzle-to-vessel welds Visual and Not applicable Not applicable There are no nozzle-to-volumetric head welds, as the nozzles are integrally cast with the heads. Instrument and sample nozzles are included in Category E. 2.3 E-1 Heater connections Visual and (See remarks) (See remarks) These connections are surface considered in Item 2.4. 2.4 E-2 Heater connections Visual (See remarks) A cumulative total of 25% The examination will be a of the heater connections visual examination for would be visually examined evidence of leaking at the time of the system hydrostatic test at or near the end of the inspection interval. I-'~ o. IN I-' I V1N 10
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TABLE 4.2-1 SECTION B. PRESSURIZER (Continued) Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks F Pressure containing dissimilar Volumetric 100% of the weld joining 100% of the weld joining Although this item is metal welds the surge line to the the surge line to the not required in the Code, outlet nozzle would be outlet nozzle and of the it is felt that dissimilar volumetrically examined weld joining the spray metal welds and stainless by the end of the 5-year line to the nozzle in the steel safe ends on the interval uppe_r head would be pressurizer should be examined examined, 2.5 G-1 Pressure retaining bolting Visual and Not applicable Not applicable There is no bolting 2 volumetric inches and larger in diameter on the Surry pressurizer. G-2 Pressure retaining bolting Visual 33% by end of 5-year Cumulative 100% by end The bolting below 2 interval of interval inches in diameter would be visually examined, either in place if the bolted connection is not disassembled during the inspection interval, or whenever the bolted connection is disassembled. The bolting to be examined would include studs and nuts. 1-'.i,,, O* IN 1-'J VIN I I-' 0
TABLE 4.2-1 SECTION B. PRESSURIZER (Continued) Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 2.7 H Integrally welded vessel supports Visual and Not applicable Not applicable There are no integrally volumetric welded vessel supports on the Surry pressurizer. 2.8 I-2 Vessel cladding Visual None 1 patch (36 square inches) A selected patch of the pressurizer cladding would be visually examined when entry into the vessel is made available for necessary maintenance. SECTION C. STEAM GENERATORS 3.1 B Longitudinal and circumferential Visual and 5% of the length of By the end of the interval, None welds, including tube sheet-to-- volumetric the circumferential 5% of the length of all head or shell welds on the weld joining the of the welds.joining the primary sidP. lower primary head to primary heads to the tube the tube sheet of one sheets would be inspected ate/ml generator would be inspected 3.2 D Primary nozzle-to-vessel head Visual and (See remarks) (See remarks) There are no nozzle-to-welds and nozzle-to-head inside volumetric head welds in the primary radiused section heads of the steam generators as the nozzles are integrally cast with the head,
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TABLE 4.2-1 SECTION C, STEAM GENERATORS Required Require'd Extent of Examination Tentative Inspec-Itea Examination Examination Planned During First tion During Ho, Category Areas Het:hods 5-Year Interval 10-Year Interval Remarks 3.3 F ~rimary nozzle-to-safe end welda Visual and (See remarks) (See remarks) The primary nozzles are surface and a buttered end pre-volumetric paration and the weld is located between the cast I)Ozzle and a cast elbow, Volumetric inspection is dependent upon meaningful examina-tion techniques, 3.4 G-1 Pressure retaining bolting V:!-sual and Not applicable Not applicable There is no pressure-volumetric retaining bolting greater than two inches in diameter on the steam generators. 3,5 G-2 Pressure retaining bolting Visual 33% by end of 5-year Cumulative 100% by end The bolting below 2 inches interval of interval in diameter would be visually examined, either in place if the bolted connection is not disassembled during the inspection interval, or whenever the bolting connection is disassemble.a. The bolting to be examined would include studs and nuts.
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e TABLE 4.2-1 SECTION C. STEAM GENERATORS (Contlnued) Required Required Extent of Examination Tentative Inspec-Item Examination Examinatio11 Planned During First tion During ~ Category Areas Methods 5-Year Interval 10-Year Interval Remarks 3.6 H Integrally welded vessel supports Visual anrl Not applicable Not applicable There ar~ no supp,rts volumetric integrally welded to the steam generator pressure boundary. 3.7 1-2 Vessel cladding Visual None 1 patch (36 square inches) A visual examination would be made only at such time as the primsry side is opened for necessary maintenance. SECTION D. PIPING PRESSURE BOUNDARY 4.1
- F Vessel-, pump-, and valve-safe Visual. and (See remarks) By the end of the inter- Examination of these ends-to-primary pipe welds and surface and *val, a cumulative 100% welds is covered under safe ends in branch piping welds volumetric of the welds would have Section A, B, and C and been examined explained in the discussion.
4.2 J Circumferential and longitudinal .visual and 5% of the circum- By the end of the inter- None pipe welds and branch pipe volumetric ferential welds, val, a cumulative 25% connection welds larger than 4 including one foot of of the circumferential inches in diameter any longitudinal weld welds in the piping on either side of the system would have been butt weld, would be examined, including one examined. foot of any longitudinal weld on either side of the butt welds I-' N
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TABLE 4.2-1 SECTION D, PIPING PRESSURE BOUNDARY Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 4.3 G-1 Pressure retaining bolting Visual and Not applicable Not applicable There is no bolting volumetric 2 inches and larger in the piping system, 4.4 G-2 Pressure retaining bolting Visual 33% of the bolting By the end of the inter- All bolting is below would be examined val, 100% of the bolting 2 inches in diameter and would be examined would be visually examined, either in place if the bolted connection ia not disassembled during the inspection interval, or whenever the bolted connection is dis-assembled. The bolting to be examined would include studs and nuts, 4.5 K-1 Integrally welded supports Visual and 10% of supports would By the end of the inter- Areas subject to volumetric be examined val, a cumulative 25% examination would include of the supports would any integrally welded be examined external support attach-ment which includes the welds to the pressure containing boundary, the base metal beneath the weld zone, and along the support attachment for distance of two base metal thicknesses.
TABLE 4,2-1 SECTION D. PIPING PRESSURE BOUNDARY (Continued) Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During_ No. Category Areas Methods 5-Yeat Interval 10-Year Interval Remarks 4.6 K-2 Piping support and hanger Visual 33% of the supports By the end of the inter- The support members and would be examined val, a cumulative 100% structures subject to of the supports would examination would include be examined those supports within the system whose structural integrity is relied upon to withstand the design loads and seismic-induced displace-ments. The support settings of constant and variable spring-type hangers, snubbers, and shock absorbers would be in-spected to verify proper dis.tribution of design loads among the associated support components. 4.7 J-2 Circumferential and long- Visual 50% of the welds By the end of the in-itudinal pipe welds and would be examined spection interval a cumu-branch pipe connections lative 100% of the welds and pipe branch connections would be examined. 4.8 J-1 Socket welds and pipe Visual and 5% of the circum- By the *end of the inspection branch connections welds Surface ferential welds and interval, a cumulative 25% ..,, I-' :n 4 inches in diameter and 5% of the pipe branch of the circumferential N I .i,. smaller connection welded. welds in the piping system N
\0 N joints would be ex~ a,d 25% of the pipe branch I
..._, N I
amined. connections welded "joints I-' 0\ would be examined.
TABLE 4.2-1 SECTION E,
- PUMP PRESSURE BOUNDARY Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No, Category Areas *-* Methods 5"-Year Interval 10-Year Interval Remarks 5.1 L-1 Pump casing welds Visual and None None The only feasible method volumetric. known to date to volumetrically inspect these pump weld casings is radiography. It is not known if such radio-graphy can be performed in service due to the design and the radiation level in the co~ponent.
If experience or a study indicates such radiography is possible, the inspection would be performed at the frequency specified in the Code. In any case, a visual inspection would be made at such time as the pump is opened for any purpose. 5,2 L-2 Pump casings Visual None By the end of the inter- The only pumps involved
*val, a cumulative 100% in this program are the of the available inner coolant pumps.
surfaces of the required pumps would be examined if the pumps are dis" assembled for maintenance 0.
~
UN 0
TABLE 4.2-1 SECTION E. PUMP PRESSURE BOUNDARY Required Required Extent of Examination Tentative Inspec""- Item Examination Examination Pl~nned During First tion During No. Cat ego.El. Areas Methods 5-Year Interval 10-Y.ear Interval Remarks 5.3 F Nozzle-to-safe end welds Visual and Not applicable Not applicable There are no nozzle-volumetric to-safe.end welds on the pumps. 5.4 G-1 Pressure retaining bolting Visual and (See remarks) By the end of the inter- Bolting 2 inches and volumetric val, a cumulative 100% larger in diameter would of the bolting would be be examined either in examined place under tension, or when the bolting is removed or when the bolting connection is disassembled for maintenance purposes. The bolting and areas to be examined would include the studs, nuts, bushings,
-threads in the base material, and the flange ligaments between thr-eaded stud holes.
TABLE 4.2-1 SECTION E. PUMP PRESSURE BOUNDARY (Continued) Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion *During No. Categorr Areas Methods 5-Year Interval 10-Year Interval Remarks 5.5 G 2 Pressure retaining *Visual (See remarks) By the end of the interval, Bolting below 2 inches in bolting a cumulative 100% of the diameter would be visually bolting would be examined examined, whenever the bolted connection is disserr~led for maintenance purposes. The bolting to be examined would includr. ; tuds ar.d nuts, 5_.6 K-1 Integraily welded Visual arid 10% of the supports By the end of the interval, None supports v9lumetric would be visually a cumulative 25% of the examined supports would be visually examined 5.7 K-2 Supports and hangers Visual 33% of the supports Cumulative 100% of the No.ne would be examined supports would be examined 5.8 Primary Pump Flywheel Visual and (See remarks)* 100% examination at or near All flywheels shall 1'e visu-allv volumetric* the end of the inspection examined at the first refueling interval and third refueling. Everv three (3) years an in-place ultrasonic volumetric examination of the areas of higher stress concentra-tion at the hare and kevwav of* each flywheel shail he performed. At the end of every 10 vears, all flywheels will be subject to a 100% ultrasonic inspection. f-' :n N I N
\0 N
I I
-...J N f-' '°
TABLE 4.2-1 SECTION F. V.Af...VE PRESSURE BOUNDARY Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned .During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 6.1 M-1 Valve body welds Visual and Not applicable Not applicable All the valves subject volumetric to this inspection* are cast valves, thus this item is not applicable. 6.2 M-2 Valve bodies Visual (See remarks) By the end of the The valves subject to this in-interval, a cu.,iula- spection may be examined during tive 100% of the maintenance or at the end of each available inner interval. I surfaces of the re-quired valves would be.examined. 6.3 F Valve-to-safe end Visual and Not applicable Not applicable There are no valves in this welds volumetric system with dissimilar metal welds. 6.4 G-1 Pressure retaining Visual and (See remarks) By the end of the Bolting.2 inches and larger in bolting volumetric interval, a cumu- diameter would be exa1I1ined either lative 100% of the in place under tension or when the bolting 2 inches and bolting is removed or when the larger would be bolting connection is disassembled examined for maintenance purposes. The bolting in areas to be exa~ined would include the studs, nuts, bushings, threads in the flange ligaments between threaded stud holes.
,-,3 I-' *tll 0
I .i,- 1-'
- l11 N I 1 00
TABLE 4.2-1 SECTION F. VALVE PRESSURE BOUNDARY (Continued) Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During .J!2.:_ Category Areas Methods 5-Year Interval 10-Year Interval Remarks 6.5 G-2 Pressure retaining bolting Visual and 33% By the end of the inter- All bolting is below 2 volumetric val, a cumulative 100% inches in diameter and of the bolting would be would be visually examined examined, either in place if the bolting connection is not disassembled at the inspection interval, or whenever the bolting
*connection is disassembled, The bolting to be examined would include studs and nuts.
6.6 . K-1 Integrally welded supports Visual and Not applicable Not applicable There are no integrally-volumetric welded supports on the valves subject to this examination, 6.7 K-2 Supports and hangers Visual 33% of the supports By the end of the inter- The support members and and hangers would be val, a cumulative 100% structures subject to examined of the supports and examination would include hangers would be those supports for piping, examined valves, and pumps within the system boundary, whose structural integrity is relied upon to withstand the design loads and seismic-induced dis-placements. I-'~ O* IN viw 0
TABLE 4.2-1 SECTION F. VALVE PRESSURE BOUNDARY (Continued) Required Required Extent of Examination Tentative Inspec-Item . Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 6.7 (Continued) The support settings of constant and variable spring-type hangers, snubbers and shock absorbers would be in-spected to verifv proper dis-tribution of design loads among the associated support components. SECTION G. MISCELLANEOUS INSPECTIONS 7.1 Materials Tensile and Capsule 1 shall be Capsules shall Capsule 4 shall be removed Irradiation Charpy V removed and examined be removed and and examined after 20 years. Surveillance Notch (Wedge at the first region examined after Capsules 5-8 are extra Open Loading) replacement. Capsule 10 years capsules for complimentary 2 shall be examined or duplicate testing. after 5 years. 7.2 Low Head SIS Visual (See Remarks) Not applicable This pipe shall be visually Piping Located inspected at each refueling in Valve Pit* shutdown. 7.3 Low Pressure Visual and 100% of blades Not applicable Turbine Rotor magnetic Blades particle or dye penetrant
"'"~
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SECTION H. SENSITIZED STAINLESS STEEL Required Required Extend of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 8.1.1 J Circumferential Visual and and Longitudinal Volumetric Twice the extent Twice the extent pipe welds and of category D.4.2 of category D.4.2 branch pipe connections larger than 4 inches in diameter 8.1. 2 J-2 Circumferential Visual Twice the extent Twice the extent and longitudinal of category D.4.7 of category D.4.7 pipe welds and branch pipe connections. 8.1. 3 J-1 Socket welds and Visual and Twice the extent Twice the extent pipe branch Surface of category D.4.8 of category D.4.8 connections welds 4 inches in dia-meter and smaller 8.2.1 Containment and Visual and (See Remarks) (See Remarks) The piping will be inspected after two Recirculation Surface years of station operation, and thereafter Spray Rings near the end of each inspection interval. The inspection will include 100% of the piping by visual examination. Surface examination will include 6 patches (each 9 inches square) evenly distributed around ,-, each spray ring. en N <" I
- 8.2.2 Remaining Visual and (See Remarks) (See Remarks) The piping would be inspected every two I-' N I I sensitized stain- Surface '1 w years. The inspection will include 100% NW less steel piping of the piping by visual examination, Surface examination will include a strip one inch wide and one foot long located on each piping bend.
TS 4.3-1
- 4.3 REACTOR COOLANT SYSTEM INTEGRITY TESTING FOLLOWING OPENING 5-1-71 Applicability Applies to test requirements for Reactor Coolant System integrity. In this context, closed is defined as that state of system integrity which permits pressurization and subsequent normal operation after the system has been opened.
Objective
- To specify tests for Reactor Coolant System integrity after the system is closed following normal opening, modification or repair.
Specification A. Each time the Reactor Coolant System is closed, the system will be leak tested at not less than 2335 psig in conformance with NDT I requirements. B. When Reactor Coolant System modifications or repairs have been made which involved new strength welds on piping and components. greater than 2 in. diameter, the new welds will receive both a surface and 100% volumetric non-destructive examination and meet applicable code
- .req.uiremen ts.
TS 4.3-2 5-1-71 .
- c. When Reactor Coolant System modifications or repairs have been made which involve new strength welds on piping and components 2 in.
diameter or smaller, the new welds will receive a surface examination. Basis For normal opening the integrity of the system, in terms of strength, is unchanged. If the system does not leak at 2335 psig (operating pressure plus 100 psi), it will be leaktight during normal operation. For repairs on piping and components greater than 2 in. diameter, the thorough nondestructive testing gives a very high degree of confidence in
- the integrity of the system, and will detect any significant defects in and near the new welds.
Significance of repairs on piping and components 2 in. diameter or smaller are relatively minor in comparison. The surface examination assures an adequate standard of integrity. In all cases, the leak test will ensure leaktightness during normal operation .
TS 4.4-1 12-29-71 4.4 CONTAINMENT TESTS Applicability Applies to containment leakage testing.
*Objective To assure that leakage of the primary reactor containment and associated systems is held within allowable leakage rate limits; and to assure that periodic surveillance is performed to assure proper maintenance and leak repair during the service life of.the containment * . Spedf ication A. Periodic and post-operational integrated leakage rate tests of the containment shall be performed .in accordance with the requirements of proposed 10 CFR 50 ,. Appendix J, "Reactor* Containment Leakage Testing For Water Cooled Power Reactors," as published in the Federal Register,. Volume_ 36, No, 167, August 27, 1971.
B. Testing. Requirements
. Type A tests will be performed in accordance with the* reduced pressure test program as defined in paragraph III, A.l. (a) of Appendix J~
- a. The reference volume method of leakage rate
- testing will be used as the method for performing the test. The absolute method of leakage rate testing will be used for verification. Test will be conduct.ed in accordance with the provisions of ANS 7.60.
TS 4. 4-2 12-29-71 I
- b. An initial leakage rate te'st will be performed at a pressure of 25 psig (Pt) and a second test at 39,2 psig (Pp)*
c, The measured leakage rate Lpm shall not exceed the design basis accident leakage rate (La) of 0.1 weight percent per 24 hours at pressure Pp.
- d. The maximum allowable test leakage rate Lt will be computed in accordance with paragraph III. A.4.
(a)(iii) of Appendix J.
- 2. Type Band C tests will be performed at a pressure of 39.i psig (Pp) in accordance with the provisions of Appendix J, section III. B. and C.
- c. Acceptance Criteria Type A, Band C tests will be considered. to be satisfactory if the acceptance criteria delineated in Appendix J, Sections III.A.5, III, A.7., III.B.3., and III.C.3 are met.
D. Retest Schedule The retest schedules for Type A, B, and C tests will be in accordance with Section III-D of Appendix J. E. Inspection and Reporting of Tests Inspection and reporting of tests will be in accordance with Section V of Appendix J,
TS 4.4-3 2-1-72 e Basis The leaktightness testing of all liner welds was performed during construction by welding a structural steel test channel over each weld seam and performing
~oap bubble and halogen leak tests.
The containment is designed for an accident pressure of 45 psig. The containment is maintained at a subatmospheric air partial pressure which varies between 9 psia and 11 psia depending upon the cooldown capability of the Engineered Safeguards and is not expected to rise above 42.5 psig for any postulated loss-of-coolant accident. All loss-of-coolant accident evaluations have been based on an integrated containment leakage rate not to exceed 0.1 percent of containment volume per 24 hr. The above specification satisfies the conditions of 10 CFR 50.54(0) which states that primary reactor containments shall meet the containment leakage test re-quirements set forth in Appendix J. References FSAR Section 5.4 Design Evaluation of Containment Tests and Inspections of Containment FSAR Section 7.5.1 Design Bases of Engineered Safeguards Instrumentation FSAR Section 14.5 Loss-of-Coolant Accident 10 CFR 50 Appendix J (Proposed), "Reactor Containment Leakage Testing for Water Cooled Power Reactors," as published in the Federal Register, Volume 36, No. 167, August 27,1971. i
TS 4.5-1 5-1-71
- 4.5 SPRAY SYSTEMS TESTS Applicability Applies to the testing of the Spray Systems.
Objective To verify that the Spray Systems will respond promptly and perform their design function, if required. Specification
- A. Test and Frequencies 1, The containment spray pumps shall be flow tested at a reduced flow rate at least once per month.
- 2. All inside containment recirculation spray pumps shall be dry tested at least once per month.
- 3. The recirculation spray pumps outside the containment shall be flow tested at a reduced flow rate at least once per month
- TS 4.5-2 2-1-72
- 4. The weight loaded check valves within the containment in the various subsystems shall be tested by pressurizing the pump dis-charge lines with air at least once each refueling period. Verifi-cation of seating the check valves shall be accomplished by applying a vacuum upstream of the valves.
- 5. All motor operated valves in the containment spray and recircula-tion spray flow path shall be tested by stroking them at least once per month.
- 6. The containment spray nozzles and containment recirculation spray nozzles shall be che.cked for proper functioning at least every five years.
- 7. The spray nozzles in the refueling water storage shall be checked for proper functioning at least monthly.
B, Acceptable Criteria
- 1. A dry-test of a recirculation spray pump shall be considered satis-factory if the motor and pump shaft rotates, starts on signal, and the ammeter readings for the motor are comparable to the original dry test ammeter readings.
- 2. A flow-test of a containment spray pump or an outside recirculation spray pump shall be considered satisfactory if the pump starts, and the discharge pressure and flow rate determine a point on the head curve. A check will be made** to* deter-
TS 4.5-3 12-29-71 mine that no particulate material from the refueling water storage tank clogs the test spray nozzles located in the refueling water storage tank.
- 3. The test of each of the weight loaded check valves shall be conside~ed satisfactory if air flows through the check valve, and if sealing is achieved.
- 4. A test of a motor operated valve shall be considered satis-factory if its limit switch operates a light on the main control board demonstrating that the valve has stroked.
- 5. The test of. the containment spray nozzles shall be considered satisfactory if the measured air flow through the nozzles indicates that the nozzles are not plugged.
- 6. The test of the spray nozzles in the refueling water storage tank shall be considered satisfactory if the monitored flow rate to the nozzles, when compared to the previously ~stablished flow rate obtained with the new nozzles, indicates no appreciable reduction in flow rate.
Basis The flow testing of each containment spray pump is performed by opening the normally closed valve in the containment spray pump recirculation line re-turning water to the refueling water storage tank. The containment spray
TS 4.5-4 12-29-71 pump is operated and a quantity of water recirculated to the refueling water storage tank, The discharge to the tank is divided into two fractions, one for the major portion of the recirculation flow and the other to pass a small quantity of water through test nozzles which are identical with those used iri the containment spray headers. The purpose of the recirculation through the test nozzles is to assure that there is no particulate material in the refueling water storage tank small enough to pass through pump suction strainers and large enough to clog spray nozzles. Due- to the physical arrangement of the recirculation spray pumps inside the containment, it is impractical to flow-test them periodically. These pumps are capable of being operated dry for 60 sec and it can be determined that the,pump shafts are turning by rotation sensors which indicate in the Main Control Room. Motor current is indicated on an ammeter in the Control Room, and will be compared with readings recorded during preoperational tests to
,ascertain that no degradation of pump operation has occurred. The recircula-tion spray pumps outside the containment have the capability of being dry-run and flow-tested, The flow-test of an outside recirculation spray pump is per~
fo~ed *by closing the suction line valve and the isolation valve between the ,_* pump discl),arge and the cot1tainnient penetration. This allows the pump casing to be filled with water and the pump to recirculate water through a test line from the pump discharge to the pump casing. With system flush conducted to remove particulate matter prior to the installa-tion of spray nozzles and with corrosion resistant nozzles and piping, it is not considered credible that a significant number of nozzles would plug during the life of the unit to reduce the effectiveness of the subsystems; therefore,
TS 4.5-5 12-29-71 provisions to air test the nozzles every five years is sufficient to indicate that plugging of the nozzles has not occurred. The spray nozzles in the refueling water storage tank provide means to ensure that there is no particulate matter in the refueling water storage tank and the Containment Spray Subsystems which could plug or cause deterioration of the spray nozzles. The nozzles in the tank are identical to those used on the containment spray headers. The monthly flow test of the containment spray pumps and recirculation to the refueling water storage will indicate any plugging of the nozzles by a re-duction of flow through the nozzles. References FSAR Section 6.3.1 Containment Spray Pumps FSAR Section 6.3.1 Recirculation Spray Pumps e*
TS 4.6-1 2-1-72 - 4,6 EMERGENCY POWER SYSTEM PERIODIC TESTING Applicability Applies to periodic testing and surveillance requirements of the Emergency Power System. Objective To verify that the Emergency Power System will respond promptly and properly when required. Specification The following tests and surveillance shall be performed as stated: A, Diesel Generators
- 1. Tests and Frequencies
- a. Manually initiated start of the diesel generator, followed by manual synchronization with other power sources and assumption of load by.the diesel generator up to 2750Jiuir. Thia teat will be conducted monthly on each diesel generator for a duration of 30 minutes, Normal station operation will not be affected by this test.
TS 4. 6-2 2-1-72
- b. Automatic start of each diesel generator, load shedding, and restoration to operation of particular vital equipment, initiated by a simulated loss of off-site power together with a simulated safety injection signal. This test will be conducted at approximately one year intervals normally during reactor shutdown for refueling to assure that the diesel generator will start within 10 sec and assume load in less than 30 sec after the engine starting signal.
- c. Availability of the fuel oil transfer system shall be verified by operating the system in conjunction with the monthly test.
- d. Each diesel generator shall be given a thorough inspection during e* each refueling interval utilizing the manufacturer's recommendations I for this class of stand-by service.
- 2. Acceptance Criteria The above tests will be considered satisfactory if all applicable equipment operates as designed.
B. Fuel Oil Storage Tanks for Diesel Generators
- 1. A minimum fuel oil storage of 35,000 gal shall be maintained on-site to assure full power operation of one diesel generator for seven days.
TS 4.6-3 12-29-71 C. St.at ion Batteries
- 1. Tests and Frequencies a, The specific gravity,electrolytic temperature,cell voltage of the pilot cell in each 60 cell battery, and the D.C. bus voltage of each battery shall be measured and recorded weekly.
- b. Each month the voltage of each battery cell in each 60 cell battery shall be measured to the nearest 0.01 volts and recorded.
- c. Every 3 months the specific gravity of each battery cell, the temperature reading of every fifth cell, the height of electrolyte of each cell, the amount of water added to any cell shall be measured and recorded.
d, Twice a year, during normal opera_tiori., t_he pattery charger shall be turned off for approximately 5 min and _the battery voltage and current shall be recorded at the beginning and end of the test. e, Once a year during the normal .refueling shutdown each battery shall be subjected to a simulated load test without battery charger. The battery voltage and current as a function of time _shall be monitored.
TS 4.6-4 12-29-71
- f. Annually connections shall be checked for tightness and anti-corrosion coating shall be applied to interconnections.
- 2. Acceptance Criteria
- a. Each test shall be considered satisfactory if the new data when compared to the old data indicate no signs of abuse or deterioration.
- b. The load test in (d) and (e) above shall be considered satis-factory if the batteries perform within acceptable limits as established by the manufacturers discharge characteristic curves.
Basis The tests specified are designed to demonstrate that the diesel generators will provide power for operation of essential safeguards equipment. They also assure that the emergency diesel generator system controls and the control systems for the safeguards. equipment will function automat1cally in the event of a loss of all normal station service power. The testing.frequency specified will be often enough to identify and correct any mechanical or electrical deficiency before it can result in a system failure *. The fuel supply and starting* circuits and controls are continuously e monitored and any faults are alarm indicated. An abnormal condition.in these systems would be signaled without having to place the diesel generators them-selves on test.
TS 4.6-5 5-1-71
- Station batteries may deteriorate. with time, but precipitous failure is extremely unlikely. The surveillance specified is that which has been demon-strated by experience to provide an indication of a cell becoming unserviceable long before it fails. In addition alarms have Qeen provided to indicate low battery voltage and low current from the inverters which would make it extremely unlikely that deterioration would go unnoticed.
The equalizing charge, as recommended by the manufacturer, is vital to main-taining the ampere-hour capability of the battery. As a check upon the effectiveness of the equalizing charge, the battery shall be loaded rather heavily and the voltage monitored as a function of time. If a cell has deteriorated or if a connection is loose, the voltage under load will drop excessively indicating the need for replacement or maintenance
- FSAR Section 8.5 provides further amplification of the basis.
References FSAR Section 8.5 Emergency Power System
TS 4. 7-1 5-1-71
- 4.7 MAIN STE.AM LINE TRIP VALVES Applicability Applies to periodic testing of the main steam line trip valves.
Objective To verify the ability of the main steam line trip valves to close upon signal. Specification A. Tests and Frequencies
- 1. Each main steam trip valve shall be tested for full closure under cold conditions approximately once a year during each refueling shutdown.
- 2. Each main steam trip valve shall be in-service tested for partial closure at least once a month.
B. Acceptance Criteria
- 1. A full closure test of main steam trip valve shall be considered satisfactory if the valve closes fully in 5 sec or less
- TS 4. 7-2 12-29-71
- 2. A partial closure in-service test of a main steam trip valve shall be considered satisfactory if the valve can be stroked at least 5 degrees from its full open position.
Basis The main steam trip valves serve to limit an excessive Reactor Coolant System cooldown rate and resultant reactivity insertion following a main steam line break accident. Their ability to close fully shall be verified at each scheduled refueling shutdown. A closure time of 5 sec was selected since this is the closure time asstlllled in the safety evaluation. The in-service testing of partial valve stroke will take place to verify the freedom of the valve disc to function as required. A limit switch in the test circuit prevents the valve disc from entering the flow stream and slamming the valve shut during in-service testing.
TS 4.8-1 12-29-71 4.8 AUXILIARY FEEDWATER SYSTEM Applicability Applies to periodic testing requirements of the Auxiliary Feedwater System. Objective To verify the operability of the auxiliary steam generator feedwater pumps and their ability to respond properly when required. Specification A. Tests and Frequency
- 1. Each motor driven auxiliary steam generator feedwater pump shall be flow tested for at least 15 minutes on a monthly basis to demonstrate its operability.
- 2. The turbine *driven auxiliary steam generator feedwater pump shall be flow tested for at least 15 minutes on a monthly basis to demonstrate its operability.
3*, The auxiliary steam generator feedwater pump dis charge valves shall be exercised on a monthly basis.
TS 4.8-2 5-1-71
- B. Acceptance Criteria These tests shall be considered satisfactory if control board indication and subsequent visual observation of the equipment demonstrate that all components have operated and sequenced properly.
Basis On a monthly basis the auxiliary steam generator feedwater pumps will be tested to demonstrate their operability by recirculation to the 100,000 Gallon Condensate Storage Tank. The capacity of any one of the three feedwater pumps in conjunction with the water inventory of the steam generators is capable of maintaining the plant
- in a safe condition and sufficient to cool the unit down.
Proper functioning of the steam turbine admission valve and the ability of the feedwater pumps to start will demonstrate the integrity of the system. Verification of correct operation can be made both from instrumentation within the Main Control Room and direct visual observation of the pumps. References FSAR Section 10.3.1 Main Steam System FSAR Section 10.3.2 Auxiliary Steam System
TS 4.9-1 5-1-71
- 4.9 EFFLUENT SAMPLING AND RADIATION MONITORING SYSTEM Applicability Applies to the periodic monitoring and recording of radioactive effluents.
Objective To ascertain that radioactive releases are maintained as low as practicable and within the limits set forth in 10CFR20. Specification
- A, Procedures shall be developed and used, and equipment which has been installed to maintain control over radioactive materials in gaseous and liquid effluents produced during normal reactor operations, including expected operational occurrences, shall be maintained and used to keep levels of radioactive materials in effluents released to unrestricted areas as low as practicable.
B. All effluents to be discharged to the atmosphere from the waste gas decay tanks* of the Gaseous Waste Disposal System shall be sampled prior to release via the process vent. Effluent from the Liquid Waste Disposal System shall be continuously monitored by the circulating water discharge tunnel monitor and, periodically, be sampled upstream of the point where it is discharged into the circulating water discharge tunnel .
TS 4.9-2 5-1-71
- C. The gross activities of all gaseous and airborne particulate ~ffluents released from the Gaseous Waste Disposal System and the Ventilation Vent System and the gross activity of all liquid effluent released from the Liquid Waste Disposal System and steam generator blowdown shall be measured and recorded continuously while they are being discharged.
D. All radiation monitor channels shall be checked, calibrated and tested as indicated in Table 4.1-1. E. The environmental program given in Table 4.9-1 shall be conducted. Basis
- The test and calibration requirements are specified to detect possible equip-ment failures and to show that maximum permissible release rates are not exceeded. All the radiation monitors except the recirculation spray cooler service water outlet monitors operate continuously and the operator observes that these instruments are performing daily. In addition, the check source for each operating channel is tripped daily from the Main Control Room to verify instrument response. Ali the monitors for a particular unit will be calibrated on a periodic basis, and normally during the refueling shutdown of that unit. Experience with instrument drift and failure modes indicates that the above specified test and calibration frequencies are adequate.
The environmental survey incorporates measurements to provide background data and measure possible plant effects. Samples collected at points where concentrations of effluents in the environment are expected to be the greatest
TS 4.9-3 5-1-71
- will be compared with samples collected' concurrently at points expected to be essentially uneffected by station effluents. The latter samples will provide background measurements as a basis for distinguishing significant radioactivity introduced into the environment by the operation of the station from that due to nuclear detonations and other sources.
This schedule will ensure that changes in the environmental radioactivity can be detected. The materials which first show changes in radioactivity are sampled most frequently. Those which are less affected by transient changes but show long-tenn accumulations are sampled less frequently. Data on the composition, quantity, frequency, etc. of releases, dilution factors obtained, and measured concentrations in food and other organisms (if any are observed) should make it desirable to review and re-evaluate this program periodically .
e TABLE 4.9-1 ENVIRONMENTAL MONITORING PROGRAM Sample Type Number of Sampling Points Frequency Type of Analysis
- l. WATERS A. James River 2 Semi-Annual Gamma Isotopic & Tritium on composit of Bimonthly Samples, Upstream & Downstream of Station B. Wells 4 Semi-Annual Gross Alpha, Gross Beta, Tritium C. Surface Water 4 Semi-Annual Gross Alpha, Gross Beta, Tritium D. Precipitation 2 Semi-Annual Gross Beta & Tritium on composit of Monthly Samples.
- 11. AIR A. Particulate 6 Monthly Gross Alpha, Gross Beta I B. Rad i ogas 6 Quarterly mrem Exposure 111. BIOTA A. Crops* Annual Gamma Isotopic B. Fowl Annual Gamma Isotopic N -I I en C. Oyster 3 Quarterly Gamma Isotopic I-'
I -l:-
-..J.
NI.O D. Clam 3 Quarterly Gamma Isotopic I
-l:-
E. Crab Annual Gamma Isotopic F. Fish*i' Semi-Annual Gamma Isotopic
*Three different crops are sampled - corn, peanuts and soybeans.
**Two types of fish are sampled - carnivore arid bottom feeder.
e TABLE 4.9-1 CONT'D. ENVIRONMENTAL MONITORING PROGRAM Sample Type Number of Sampling Points Frequency Type of Analysis IV. SI LT 5 Semi-Annual Gamma Isotopic V. SOIL 6 Annual Gamma Isotopic VI. MILK 4 Quarterly 1-131 , Cs-137, Sr-90, Calcium N -I I C/l t--' I -t:-
-...J.
N I..O I V1
TS 4.10-1 2-1-72 4.10 REACTIVITY ANOMALIES Applicability Applies to potential reactivity anomalies. Objective To require evaluation of applicable reactivity anomalies within the reactor. Specification A, Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be compared monthly with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of one percent in reactivity, an evaluation as to the cause of the discrepancy shall be made and reported to the Atomic Energy Commission per Section 6.6 of these Specifications. B. During periods of power operation at greater than 10% of power, design N N peaking factors, F and FLiH' shall be determined monthly using data from q limited core maps. If these factors exceed values of Design Limits Interim Limits FN = 2.72 N FN = 2.52 FN = 1.50 q FliH .= 1.58 q liH art evaluation as to the cause of the anomaly shall be made.
TS 4.10-2 2-1-72 Basis BORON CONCENTRATION To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions. When full power is reached initially, and with the control rod assembly groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration, and the slope of the curve relating burnup and reactivity is compared with that predicted. This process of normalization should be completed after about 10% of the total core burnup. Thereafter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated. Any reactivity anomaly greater than 1% would be un-expected, and its occurrence would be thoroughly investigated and evaluated. The value of 1% is considered a safe limit since a shutdown margin of at least 1% with the most reactive control rod assembly in the fully withdrawn position is always maintained. PEAKING FACTORS A thermal criterion in the reactor core design specifies that "no fuel melting during any anticipated normal operating condition" should occur, To meet the above criterion during a thermal overpower of 112% with additional margin for design uncertainties, a steady state maximum linear power is
TS 4.10-3 2-1-72 selected. This then is an upper linear power limit determined by the maximum central _temperature of the hot pellet. The peaking factor is a ratio taken between the maximum allowed linear power density in the reactor to the average value over the whole reactor. It is of course the average value that determines the operating power level, The peaking factor is a constraint which must be met to assure that the peak linear power density does not exceed the maximum allowed value. During normal reactor operation, measured peaking factors should be significantly lower than design limits. As core burnup progresses, measured designed peaking factors are expected to decrease. A monthly determination of N N F and F, is adequate to ensure that core reactivity changes with burnup have q LlH not significantly altered peaking factors in an adverse direction.
TS 4.11-1 5-1-71
- 4.11 SAFETY INJECTION SYSTEM TESTS Applicability Applies to operational testing of the Safety Injection System.
Objective To verify that the Safety Injection System will respond promptly and perform its design functions; if required. Specification A. Safety Injection System
- 1. System tests shall be performed during reactor shutdowns for refueling. The test shall be performed in accordance with the following procedure:
With the Reactor Coolant System pressure less than or equal to 450 psig and temperature less than or equal to 350°F, a test safety injection signal will be applied to initiate operation of the system. The charging and low head safety injection pumps are immobilized for this test
- TS 4.11-2 2-1-72
- 2. The test will be considered satisfactory if control board indication and/or visual observations indicate that all the appropriate components have received the safety injection signal in the proper sequence. That is, the appropriate pump breakers shall have opened and closed, and all valves, required to establish a safety injection flow path to the Reactor Coolant System and to isolate other systems from this flow path, shall have completed their stroke, B. Component Tests Pumps
- 1. The low head safety injection pumps and charging pumps shall be operated at intervals not greater than one month.
- 2. Acceptable levels of performance for the low head safety injection pumps shall be that the pumps start, reach their required developed head on recirculation flow and the control board indications and/or visual observations indicate that the pumps are operating properly.
- 3. In addition to the Safety Injection System, the charging pumps form an integral part of the Chemical and Volume Controi System (CVCS),
and are operated on a routine basis as part of this system. If these pumps have performed their design function as part of the routine operation of the eves, their level of performance will be deemed acceptable as related to the Specification.
TS 4.11-3 2-1-72 Valves
- 1. The refueling water storage tank outlet valves shall be tested in performing the pump tests.
- 2. The accumulator check valves shall be checked for operability during each refueling shutdown.
- 3. All valves required to operate on a safety injection signal shall be tested for operability each refueling shutdown.
Basis Complete system tests cannot be performed when the reactor is operating be-cause a safety injection signal causes containment isolation. The method of assuring operability of these systems is therefore to combine systems tests to be performed during refueling shutdowns, with more frequent component tests, which can be performed during reactor operation. The systems tests demonstrate proper automatic operation of the Safety Injection System. With the pumps blocked from starting, a test signal is applied to initiate automatic action and verification is made that the components receive the safety injection signal in the proper sequence, The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry. During reactor operation, the instrumentation which is depended on to initiate safety injection is checked periodically and the initiating circuits are tested in accordance with Specification 4.1. In addition, the active components (pump
TS 4.11-4 12-29-71 and valves) are to be tested monthly to check the operation of the starting circuits and to verify that the plllllps are in satisfactory running order. The test interval of one month is based on the judgment that more frequent testing would not significantly increase the reliability (i.e. the probability that the component would operate when required), yet more frequent testing would result in increased wear over a long period of time. Other systems that are also important to the emergency cooling function are the accumulators and the Containment Depressurization System, The accumulators are a passive safeguard. In accordance with Specification 4.1 the water volume and pressure in the accumulators are checked periodically. Reference FSAR Section 6.2 Safety Injection System
TS 4.12-1 12-29-71 -- 4.12 VENTILATION FILTER TESTS Applicability Applies to the testing of particulate and charcoal filters in safety related air filtration systems. Objective To verify that leakage efficiency and iodine removal efficiency are within acceptable limits. Specification A. Tests and Frequencies
- 1. The charcoal filters in the Auxiliary Building filter banks, control room emergency filter banks, and relay room emergency filter banks shall be tested for leakage efficiency at least once every 12-18 month period, normally during refueling shutdown using an in-place Freon~ll2 (or equivalent) test method.
- 2. The particulate filters. in the Auxiliary Building filter, c.ontrol room emergency filter banks, and relay room emergency filter banks shall be tested for leakage efficiency at least once every 12-18 month period, normally during refueling shutdown usirig an in-place DOP test method,
- 3. A carbon sample will be removed from one of the banks once every third year and subjected to chemical analysis to determine the iodine removal capability.
4 *. Instrumentation, equipment, a11d procedures shall generally conform to
TS 4.12-2 2-1-72 the recormnendations in ORNL-NSIC-65, "Design, Construction and Testing of High-Efficiency Air Filtration Systems for Nuclear Application", C. A. Burchsted and A. B. Fuller, Oak Ridge National Laboratory, USAEC, January, 1970. B. Acceptance Criteria
- 1. The in-place leakage tests on charcoal units using Freon-112 (or equivalent) shall be conducted on each filter bank. Removal of 99.0%
of the Freon-112 (or equivalent) shall constitute acceptable performance.
- 2. The in-place leakage tests on the particulate filters using DOP shall be conducted on each train. Removal of 99.5% of the DOP shall constitute acceptable performance.
- 3. Chemical analysis of the carbon sample shall be performed to demonstrate the iodine adsorption capability of the charcoal. Verification of an elemental iodine removal efficiency of 99.0% shall constitute acceptable performance.
Basis The purpose of the Auxiliary Building Filter Banks is to provide standby capa-bility for removal of particulate and iodine contaminants from any of the venti-lation systems in the auxiliary building, fuel building, decontamination building, safeguards area adjacent to the containments, and the reactor con-tainments (during shutdown) which discharge through the ventilation vent and could require filtering prior to release. The exhausts of the above systems can be diverted, if required, through the Auxiliary Building filter banks remotely from the control room. The Safeguards Area exhaust is automatically diverted through the filter banks in the event of a LOCA (diverted on high-high contai~
TS 4.12-3 2-1-72 ment pressure). The fuel building exhaust is aligned to continuously exhaust through the filters during spent fuel handling in the spent fuel pool. The purpose of the control and relay room emergency filter banks is to provide emergency ventilation for the control and relay rooms during accident conditions. The off-site dose calculations for LOCA and fuel handling accidents, assume only 90% iodine removal efficiency for the air passing through the charcoal filters. Therefore, demonstration of 99% efficiency once each refueling cycle will assure the required capability of the filters is met or exceeded. System components are not subject to rapid deterioration, having lifetimes of many years, even under continuous flow conditions. The tests outlined above provide assurance of filter reliability and will insure early detection of conditions which could cause filter degradation. Instrumentation, equipment, and procedures for testing shall generally conform to the recommendations in ORNL-NSIC-65, which is considered to be the best available guide for design, construction, and testing of particulate and iodine removal filter systems. Reference FSAR Section 9.13, Auxiliary Ventilation Systems
TS 5-1.1 5-1-71
- 5.0 DESIGN FEATURES 5.1 SITE Applicability Applies to the location and boundaries of the site for the Surry Power Station.
Objective To define those aspects of the site which will affect the overall safety of the installation *
- Specification The Surry Power Station is located in Surry County, Virginia, on property owned by Virginia Electric and Power Company on a point of land called Gravel Neck which juts into the James River. It is approximately 46 miles SE of Richmond, Virginia, 17 miles NW of Newport News, Virginia, and 25 miles NW of Norfolk, Virginia. The minimum distance from a reactor centerline to the site exclusion boundary as defined in 10CFRlOO is 1,650 ft. This is the
- distance for Unit 1, which is controlling.
References FSAR Section 2.0 Site FSAR Section 2.1 General Description
TS 5.2-1 5-1-71
- 5.2 CONTAINMENT Applicability Applies to those design features of the reactor containment structures and con-tainment systems relating to operational and public safety.
Objective To define ths significant design features of the reactor containment structures and containment systems. Specifications A. Structure
- 1. A containment structure completely encloses each reactor and Reactor Coolant System and assures that an acceptable upper limit for leakage of radioactive materials to the environment is not exceeded even if gross failure of a Reactor Coolant System occurs. Each structure provides biological shielding for both normal operation and accident situations. Each containment structure is designed for an internal subatmospheric pressure of 8 psia .
TS 5.2-2 12-29-71
- 2. Each containment structure is designed for a reactor operating at the ultimate rated thermal power of 2546 Mwt.
- 3. Each containment structure is designed to withstand an internal design pressure of 45 psig acting simultaneously with: (1) loads resulting from an Operational Basis Earthquake having a hori-zontal ground acceleration of 0.07 g at zero period with an assumed structural damping factor of 5 percent, or (2) loads resulting from a Design Basis Earthquake having a horizontal ground acceleration of 0.15 g at zero period with an assumed structural damping factor of 10 percent.
*B Containment Penetrations
- 1. All penetrations through the containment structure for pipe, electrical conductors, ducts, and access hatches are of the double barrier type.
- 2. The automatically actuated isolation valves _ar~ designed to close as.outlined below. The actuation system is designed such that no single component failure will prevent contain-ment isolation if required. Refer to Table 3.7-4 in the Technical Specifications for set point values of the signals.
- a. A safety injection signal closes all trip valves which are located in normally open lines connecting the reactor e coolant loops and penetrating the containment.
TS 5.2-3 5-1-71
- b. A high containment pressure isolation signal closes the automatic trip valves in all normally open lines penetrating the containment which are not required to be open to control containment pressure to perform an orderly reactor shut down without actuation of the consequence limiting safe-guards in case of a small Reactor Coolant System leak.
- c. A further rise in containment pressure, indicating a major loss-of-coolant accident, produces a containment high-high pressure_ isolation signal which closes all normally open lines which penetrate the containment which have not been closed by 2-b above *
- d. Isolation can be accomplished manually from the control in the Main Control Room if any of the automatic signals fail to actuate the above valves.
- c. Containment Systems
- 1. Following a loss-of--coolant accident, the Containment Spray Subsystems distribute at least 2,600 gpm borated water spray containing sodium hydroxide for iodine removal within the containment atmosphere. The Recirculation Spray Subsystem recirculate at least 3,500 gpm of water from the containment sump
- TS 5.2-4 5-1-71
- 2. No part of the Containment Ventilation System is designed for continued operation during a total loss-of-coolant accident.
It may, however, continue to operate with small Reactor Coolant System leaks until the Containment Spray System is initiated. References FSAR Section 5.2 Containment Isolation FSAR Section 5.3 Containment Systems FSAR Section 5.4 Design Evaluation FSAR Section 7.2.2 Safeguards Initiation and Containment Isolation FSAR Section 15.2.4 Seismic Design FSAR Section 15.5.1 Containment Structure Technical Specification Section 3.3 Safety Injection System
TS 5.3-1 12-29-71 5.3 REACTOR Applicability Applies to the reactor core, Reactor Coolant System, and Safety Injection System. Objective To define those design features which are essential in providing for safe system operations * .Specifications J A. Reactor Core
- 1. The reactor core contains approximately 176,200 lbs. of uranium J dioxide in the form of slightly enriched uranium dioxide pellets.
The pellets are encapsulated in Zircaloy-4 tubi,ng to form fuel rods. All fuel rods are pressurized with helium during fabrication. The reactor core is made up of 157 fuel assemblies. Each fuel assembly contains 204 fuel rods.
- 2. The average enrichment of the initial core is 2.51 weight per cent of U-235. Three fuel enrichments are used in the initial core. The highest enrichment is 3.12 weight per cent of U-235.
TS 5.3-2 12-29-71
- 3. Reload fuel will be similar in design to the initial core.
The enrichment of reload fuel will not exceed 3.60 weight per cent of U-235.
- 4. Burnable poison rods are incorporated in the initial core.
There are 816 poison rods in the form of 12 rod clusters, which are located in vacant control rod assembly guide thimbles. The burnable poison rods consist of pyrex glass clad with stain-less steel.
- 5. There are 48 full-length control rod assemblies and 5 part-length control rod assemblies in the reactor core. The full~
length control rod assemplies contain a 144 inch-length of silver-indium-cadmium alloy clad with stainless steel. The part-length control rod assemblies contain a 36 inch length of silver-indium-cadmiun alloy with the remainder of the stainless steel sheath filled with Al 2o3 *
- 6. . The initial core and subsequent cores will meet the following performance criteria at all times during the operating lifetime.
- a. Nuclear hot channel factors:
Deslgn Limits Interim Limits FN 2.72 FN = 2.52 q q N N F l'iH = 1.58 F l'iH = 1 *.50
TS 5.3-3 12-29-71 - b. Moderator temperature coefficient n_egative in the power operating range.
- c. Capable of being made subcritical in accordance with Specification 3.12 A.3.C
- 7. Up to 10 grams of enriched fissionable material may be used either in the core or available on the plant site, in the form of fabricated neutron flux detectors for the purposes of monitoring core neutron flux.
B. Reactor Coolant System
- 1. The design o_f the Reactor Coolant System complies with the code requirements specified in Section 4 of the FSAR.
- 2. All piping, components and supporting structures of the Reactor Coolant System are designed to Class 1 seismic requirements, and have been designed to withstand:
.a. Primary operating stresses combined with th,e Operational seismic stresses resulting from a horizonal ground acceleration of 0.07g and a simultaneous vertical ground acceleration of 2/3 the horizonal, with the stresses maintained within code allowable working stresses.
- b. Primary operating stresses when combined with the Design Basis Earthquake seismic stresses resulting from a horizonal ground acceleration of 0.15g and a simultaneous vertical ground
TS 5.3-4 12-29-71 acceleration of 2/3 of the horizonal, with the stresses maintained within code allowable working stresses.
- 3. The total liquid volume of the Reactor Coolant System, at rated operating conditions, is approximately 9300 cubic feet.
TS 5.4-1 12-29-71 5.4 FUEL STORAGE Applicability Applies to the design of the new and spent fuel storage areas. Objective To define those aspects of fuel storage relating to prevention of criticality in fuel storage areas; to prevention of dilution of the borated water in the reactor; and to prevention of inadvertent draining of water from the spent fuel storage area. Specification A. The reinforced concrete structure and steel superstructure of the Fuel Building and spent fuel storage racks are designed to withstand Design Basis Earthquake loadings as Class I structures. The spent fuel pit has a stainless steel liner to ensure against loss of water*. B. The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed loca-tions. Both new and spent fuel is stored vertically in an array with a distance. of 21 inches between assemblies to assure kefr0.90, even if unborated water were used to fill the spent fuel storage pit or the new fuel storage area.
TS 5.4-2 6-30-71 C. Whenever there is spent fuel in the spent fuel storage pit, the pit shall be filled with borated water at a boron concentration not less than 2,000 ppm to match that used in the reactor cavity and refueling canal during refueling operations. D. The only drain which can be connected to the spent fuel storage area is that in the reactor cavity. The strict step-by-step procedures used during refueling ensure that the gate valve on the fuel transfer tube which connects the spent fuel storage area with the reactor cavity is closed before draining of the cavity commences. In addition, the procedures require placing the bolted blank flange on the fuel -- transfer tube as soon as the reactor cavity is drained. References FSAR Section 9.5 Fuel Pit Cooling System FSAR Section 9.12 Fuel Handling System
TS 6.1-1 2-1-72 6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION, SAFETY AND OPERATION REVIEW Specification A. The Station Manager shall be responsible for the safe operation of the facility. The Station Manager shall report to the Superintendent - Production Operations. The relationship between this supervisor and other levels of company management is shown in TS Figure 6.1-1. B. The station organization shall conform to the chart as shown in TS Figure 6.1-2.
- 1. Qualifications with regard to education and experience and the technical specialties of key supervisory personnel will meet the minimum acceptable levels described in ANSI Nl8.1 "Selection and Training of Nuclear Power Plant Personnel," dated March 8, 1971.
The key supervisory personnel are as follows:
~) Manager b) Superintendent-Station Operations c) Operating Supervisor d) Supervisor-Electrical Maintenance e) Supervisor-Mechanical Maintenance f) Supervisor-Engineering Services g) Chemistry and Health Physics Supervisor h) Shift Supervisor
TS 6.1-2 2-1-72
- 2. Retraining and replacement training of station personnel shall be 11 in accordance with ANSI Nl8.l Selection and Training of Nuclear Power Plant Personnel, 11 dated March 8, 1971.
- 3. The following requirements supplement the applicable regulations of 10 CFR 50.54:
Condition Minimum Complement
- 1. One unit operating 1 SOL, 2 LO, 2 AO
- 2. One unit operating and 1 SOL°*l 3 LO, 2 AO one unit shutdown
- 3. Both units fueled and 1 SOL, 1 LO, 1 AO shutdown
- 4. Both units operating 2 SOL, 3 LO, 2 AO Note:
SOL= Senior Licensed Operator as defined by 10 CFR 55.4(e) LO - Licensed Operator as defined by 10 CFR 55.4(d) AO = Auxiliary Operator
- When the shutdown unit is undergoing refueling or startup, 1 additional SOL will be added to this shift complement to ensure supervision of these activities.
C. Organization units to provide a continuing review of the operational and safety aspects of the nuclear facility shall be constituted and have the authority and responsibilities outlined below:
- 1. Station Nuclear Safety and Operating Committee
- a. Membership
- 1. Chairman - Manager
- 2. Vice Chairman - Superintendent-Station Operation
TS 6.1-3 2-1-72
- 3. Operating Supervisor
- 4. Supervisor-Electrical Maintenance
- 5. Supervisor-Mechanical Maintenance 6, Supervisor-Engineering Services
- 7. Chemistry and Health Physics Supervisor b, Qualifications: The qualifications of the regular members of the Station Nuclear Safety and Operating Committee with regard to the combined experience and technical specialties of the individual members shall be maintained at a level at least equal to those described in Section 6.1, B.l. of these Specifications.
- c. Meeting frequency: As called by the Chairman but not less than monthly.
- d. Quorum: Chairman or Vice Chairman, Chemistry and Health Physics Supervisor or his designee, and three others to provide a quorum of five members.
- e. Responsibilities
- 1. Periodically review all proposed normal, abnormal, and emergency operating procedures and all proposed maintenance procedures. Review proposed changes to those procedures, and any other proposed procedures or changes thereto as determined by the Station Manager to affect nuclear safety.
TS 6.1-4 2-1-72
- 2. Review all proposed test and experiment procedures and results thereof when applicable.
- 3. Review proposed changes to Technical Specifications.
- 4. Review all proposed changes or modifications to systems or equipment that would require a change in established procedures, or which would constitute a design change.
5, Periodically review all operations to detect any potential safety hazards.
- 6. Investigate all reported instances of departure from Technical Specification. limits, such investigations to include review, evaluation and recommendations to prevent recurrence, to the Station Manager, Superintendent-Production Operations and to the Chairman of the System Nuclear Safety and Operating Committee.
7, The Station Nuclear Safety and Operating Committee shall make tentative determinations as to whether or not proposals considered by the Committee involve unreviewed safety questions. This determination shall be subject to review and approval by the System Nuclear Safety and Operating Committee.
- 8. Review all abnormal occurrence reports.
TS 6.1-5 2-1-72 9, Perform special reviews and investigations and render reports thereon as requested by the Chairman of the System Nuclear Safety and Operating Committee.
- 10. Initiate periodic drills to test the effectiveness of the emergency procedures.
- f. Authority
- 1. The Station Nuclear Safety and Operating Committee shall advise the Manager on all matters affecting the safe operation of the facility.
- 2. The Station Nuclear Safety and Operating Committee shall recommend to the Station Manager approval or disapproval of proposals under items e(l) through (4) above.
a) In the event of disagreement between the recommendations of the Station Nuclear Safety and Operating Committee and the actions contemplated by the Station Manager, the course determined by the Station Manager will be followed with immediate notification to the Superintendent-Production Operations and the Chairman of the System Safety and Operating Committee.
- g. Records Minutes shall be kept of all meetings of the Station Nuclear
TS 6.1-6 2-1-72 Safety and Operating Committee and copies shall be sent to the Superintendent-Production Operations and to all members of the Station and System Nuclear Safety and Operating Committees.
- h. Procedures Written administrative procedures for committee operation shall be prepared and maintained describing the method of submission, and the content of presentations to the committee, provisions for the use of subcommittees; review and approval by members of written committee evaluations and recommendations; the distri-butions of minutes; and, such other matters as may be appropriate,
- 2. System Safety and Operating Cormnittee
- a. Membership
- 1. Chairman and Vice Chairman appointed by name by the Vice President-Power, which may be an individual listed in Item 2.
- 2. Five members of the Production Department system office staff (refer to TS Fig. 6.1-1) who are experienced in utility operation and procedures:
Director - Production Operations and Maintenance Superintendent - Production Operations Director - Power Station Design
TS 6.1-7 2-1-72 Director - Nuclear Services Supervisor - Nuclear Design
- 3. Manager of each nuclear generating station operating on the Virginia Electric and Power Company system, or his designee. In matters or consideration of proposals pertinent to a particular station, the Manager of this station shall serve as a non-voting member of the committee, In matters pertaining to other stations, Station Managers will serve as voting members.
- 4. At least one qualified non-company affiliated technical consultant. Duly appointed consultant members shall have equal vote with permanent members of the Committee.
- b. Qualifications The minimum qualifications of the Company members of the System Safety and Operating Committee will be:
an engineering graduate or equivalent with combined nuclear and conventional experience in power station design and/or operation of eight years, with at least two years involving the direction of nuclear operations or design activity.
- c. Consultants The committee shall have the authority to call technically
TS 6.1-8 2-1-72 qualified personnel from within the Virginia Electric and.Power Company organization or from any other consultant source.
- d. Quorum: Either the Chairman or Vice Chairman and two thirds of the other members shall constitute a quorum.
- e. Meeting frequency: As required by the Chairman but not less than quarterly.
- f. Responsibilities
- 1. Review proposed changes to the operating license including Technical Specifications.
- 2. Review minutes of meeting of the Station Safety and Operating Committee(s) to determine if matters considered by that committee involve unreviewed or unresolved safety questions.
3, Review matters including proposed changes or modifications to systems or equipment having safety significance referred to it by the Station Nuclear Safety and Operating Committee or by the Station Manager.
- 4. Conduct periodic review of station operations.
- 5. Review all reported instances of departure from Technical Specification limits and report findings and recommendations to prevent recurrence to the Manager-Power Production.
TS 6.1-9 2-1-72
- 6. Perfonn special reviews and investigations and render reports thereon as requested by company management, or as the committee deems necessary.
7, Review proposed tests and experiments and results thereof when applicable.
- 8. Review abnormal performance of plant equipment and anomalies.
9, Review unusual occurrences and incidents which are reportable under the provisions of 10 CFR 20 and 10 CFR 50.
- 10. Review of occurrences if Safety Limits are exceeded.
- g. Authority
- 1. Recommend approval of proposed changes to the operating license, including Technical Specifications, for submission to the A.E.C.
- 2. Recommend approval of proposed changes or modifications to systems or equipment, provided such changes or modifications do not involve unreviewed safety questions.
h, Records Minutes shall be recorded of all meetings of this Committee.
TS 6.1-10 2-1-72 Copies of the minutes shall be forwarded to the Vice President-Power Production, all members of the committee and any others that the Chairman may designate.
- i. Procedures Written administrative procedures for committee operation shall be maintained describing the method of submission and the content of presentations to the committee; provisions for use of subcommittee evaluations and recommendations; distribution of minutes; and, such other matters as may be appropriate.
TS FIG. 6.1-1 2-1-72 POWER PRODUCTION DEPARTMENT ORGANIZATION CHART VIRGINIA ELECTRIC AND POWER COMPANY VICE PRESIDENT POWER I I MANAGER MANAGER SYSTEM NUCLEAR POWE~ PRODUCTION I-- SAFETY & OPERATING FUEL RESOURCES COMMITTEE DIRECTOR DIRECTOR DIRECTOR PRODUCTION POWER STATION OPERATIONS AND NUCLEAR MAINTENANCE SERVICES DESIGN I : SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT SUPERVISOR PRODUCTION PRODUCTION PRODUCTION NUCLEAR MAINTENANCE OPERATIONS ENG !NEERING DESIGN MANAGER SURRY
e VffiGINIA ELECTRIC AND POWER COMPANY* SURRY POWER STATION ORGANIZATION CHART STATION NUCLEAR SAFETY & OPERATING STATION MANAGER COMMITTEE
- SUPERINTENDENT STAT ION E SL OPERATIONS I I I OPERATING SUPERVISOR. SUPERVISOR SUPERVISOR CHEMISTRY AND SUPERVISOR ELECTRICAL MECHANICAL ENGINEERING HEALTH PHYSICS*
MAINTENANCE MAINTENANCE SERVICES SUP ERV ISOR ESL E C I I I I SHIFT INSTRUMENT SUPERVISOR ENGINEER SUPERVISOR SL E I
** CONTROL ROOM OPERATOR OL LEGEND I
SL - SEN IOR LICENSE ASSISTANT OL - OPERATOR'S LICENSE 1-3 CJ) CONTROL ROOM E - GRADUATE ENGINEER >tj OPERATOR C - COLLEGE GRADUATE H OL SENIOR LICENSE WILL BE OBTAINED N c.l I WITHIN 18 MONTHS AFTER INITIAL I
- I-' °'
** CRITICALITY OF UNIT NO. 1 I I-'
-...i I AUXILIARY SHIFT COMPLEMENTS FOR DIFFERENT NN OPERATOR STATION CONDITIONS ARE DETAILED IN SPECIFICATION 6. 1-B.3 I
TS 6.2-1 2-1-72 6.2 ACTION.TO BE TAKEN IN THE EVENT OF AN ABNORMAL OCCURRENCE IN STATION OPERATION Specification A. Any abnormal occurrence shall be reported immediately to and promptly
. reviewed by the Chairman of Station Nuclear Safety and Operating Committee or his designee, Superintendent - Production Operations and the Chairman of the System Nuclear Safety and Operating Committee.
B. The Shift Supervisor on duty and subsequently the Operating Supervisor shall prepare a report for each abnormal occurrence. This report shall include an evaluation of the cause of the occurrence and also recommenda-tions for appropriate action to prevent or reduce the probability of a re-currence. Immediate corrective action shall be taken to correct the anomaly. C. Copies of all such reports shall be submitted to the Superintendent-Station Operations, the Station Manager, who also serves as the Chairman of the Station Nuclear Safety and Operating Committee, Superintendent-Production Operations, and to the Chairman of the System Nuclear Safety and Operating Committee for review and approval of any recommendation. D. The Vice President - Power shall report the circumstances of any abnormal occurrence to the AEC as specified in Section 6.6 of these Specifications.
TS 6.2-2 5-1-71
- Station Operations, the Station Manager, who also serves as the Chairman of the Station Nuclear Safety and Operating Committee, Superintendent - Production Operations, and to the Chairman of the.
System Nuclear Safety and Operating Committee for review and approval of any reconnnendation. D. The Vice President - Power shall report the circumstances of any abnormal occurrence to the AEC as specified in Section 6.6 of these Specifications
- TS 6.3-1 2-1-72 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED Specification A. Should a safety limit (see Section 2.0 of the Technical Specifications) be exceeded, the reactor shall be shutdown and reactor operation shall only be resumed in accordance with the authorization within 10 CFR
- 50. 36 (c) (1) (i).
B. An immediate report of the incident shall be made to the Station Manager, Superintendent - Production Operations and the Chairman of the System Nuclear Safety and Operating Connnittee. C. The Station Manager shall promptly report the circumstances to the AEC as specified in Section 6.6 of these Specifications. D. A complete analysis of the incident together with recommendations to prevent recurrence shall be prepared by the Shift Supervisor and the Operating Supervisor. A preliminary written report shall be submitted to the Superintendent Station Operations, Station Manager who is also the Chairman of the Station Nuclear Safety and Operating Committee, Superintendent - Production Operations, and the Chairman of the System Nuclear Safety and Operating Connnittee within 24 hours following the incident. Appropriate analyses or reports will be submitted to the AEC by the Vice President - Power as specified in Section 6.6 of these Specifications.
TS 6.3-2 5-1-71
- D. A complete analysis of the incident together with reconnnendations to prevent recurrence shall be prepared by the Shift Supervisor and the Operating Supervisor. A preliminary written report shall be submitted to the Superintendent Station Operations, Station Manager who is also the Chairman of the Station Nuclear Safety and Operating Conunitte, Superintendent - Production Operations, and the Chairman of the System Nuclear Safety and Operating Committee within 24 hours following the incident.
Appropriate analyses or reports will be submitted to the AEC by the Vice President - Power as specified in Section 6,6 of these Specifications
- TS 6.4-1 2-1-72 6.4 UNIT OPERATING PROCEDURES Specification A. Detailed written procedures with appropriate check-off lists and instructions shall be provided for the following conditions:
- 1. Normal startup, operation, and shutdown of a unit, and of all systems and components involving nuclear safety of the station.
- 2. Calibration and testing of instruments, components, and systems involving nuclear safety of the station.
- 3. Actions to be taken for specific and foreseen malfunctions of systems or components including alarms, primary system leaks and abnormal reactivity changes.
- 4. Release of radioactive effluents
- 5. Emergency conditions involving potential or actual release of I
radioactivity
- 6. Emergency conditions involving violation of industrial security.
- 7. Preventive or corrective maintenance operations which would have an effect on the safety of the reactor.
- 8. Refueling operations.
B. Radiation control procedures shall be provided and made available to all station personnel. These procedures will show permissible radiati.on exposure. This radiation protection program shall be organized to meet the requirements of 10 CFR 20 and/or the following provisions:
TS 6.4-2 2-1-72
- 1. The intent of 10 CFR 20.203(c)(2)(iii) shall be implemented by satisfying the following conditions:
- a. The entrance to each radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posed.
- b. The entrance to each radiation area in which the intensity of radiation is equal to or greater than 1000 mrem/hr shall be provided with locked barricades to prevent unauthorized entry into these areas. Keys to these locked barricades shall be maintained under the administrative control of the Shift Supervisor.
- c. All such accessible high radiation areas shall be surveyed by Health Physics personnel on a routine schedule, as determined by the Chemistry and Health Physics Supervisor, to assure a safe and practical program.
d, Any individual entering a high radiation area shall have completed the indoctrination course designed to explain the hazards and safety requirements involved, or shall be escorted at all times by a person who has completed the course.
- e. Any individual or group of individuals permitted to enter a high radiation area per l.d. above, shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.
TS 6.4-3 2-1-72 f, Entrance to areas with radiation levels in excess of 1 R/hr shall require the use of the "buddy system, 11 whereby a minimum of two individuals maintain continuous visual and/or verbal communication with each other; or other mechanical and/or electrical means to provide constant communication with the individual in the area shall be provided.
- g. A Radiation Work Permit system shall be used to authorize and control any work performed in high radiation areas.
h, All buildings or structures, in or around which a high radiation area exists, shall be surrounded by a chain-link fence.The entrance gate shall be locked under administrative control, or continuously guarded to preclude unauthorized entry. i, Stringent administrative procedures shall be implemented to assure adherence to the restriction placed on the entrance to a high radiation area and the radiation protection program associated thereto.
- 2. Pursuant to 10 CFR 20.103(c)(l) and (3), allowance can be made for the use of respiratory protective equipment in conjunction with activities authorized by the operating license in determining whether individuals are exposed to concentrations in excess of the limits specified in Appendix B, Table I, Column 1, of 10 CFR 20, subject to the following limitations:
- a. The limits provided in 10 CFR 20.103(a) and (b) are not exceeded,
- b. If the radioactive material is of such form that intake through the skin or other additional route is likely, individual exposures to radioactive material shall be controlled so that the radioactive
TS 6. 4-4 12-29-71 e content of any critical organ from all. routes of intake averaged over seven (7) consecutive days does not exceed that which would result from inhaling such radioactive material for forty (40) hours at the pertinent concentration values provided in Appendix B, Table I, Column 1, of 10 CFR 20.
- c. For radioactive materials designated "Sub" in the "Isotope" column of Appendix B, Table I, Column 1 of 10 CFR 20, the concentration value specified is based upon exposure to the material as an external radiation source. Individual exposures to these materials shall be accounted for as part of the limitation on individual dose in §20.101
- 3. In all operations in which adequate limitation of the inhalation of radioactive material by the use of process or other engineering controls is impracticable, an individual shall be permitted to use respiratory protective equipment to limit the inhalation of airborne radioactive material provided:
a) The limits specified in B.2 above are not exceeded. b) Respiratory protective equipment is selected and used so that the peak concentration of airborne radioactive material inhaled by an individual wearing the equipment does not exceed the pertinent concentration values sp.ecified in Appendix B, Table I, Column 1, of 10 CFR 20. The concentration of radioactive material that is inhaled when respirators are worn can be determined by dividing the ambient airborne concentration by the protection factor specified in this specification. If the intake of radioactivity is later deter-mined by o.ther measurements to have been different than that initially estimated, the later quantity shall be used in
TS 6.4-5 12-29-71 - c) evaluating the exposures. Each respirator user is advised that he may leave the area at any time for relief from respirator use in case of equipment malfunction, physical or psychological discomfort, or any other condition that might cause reduction in the protection afforded the wearer. d) A respiratory protective program adequate to assure the requirements of this specification are met is maintained. Such a program shall include: (1) Air sampling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper selection of respiratory protective equipment. (2) Written procedures to assure proper selection, super-vision, and training of personnel using such protective equipment. (3) Written procedures to assure the adequate fitting of respirators; and the testing of respiratory protective equipment for operability. (4) Written procedures for maintenance.to assure full effectiveness of respiratory protective equipment, including issuance, cleaning and decontamination, inspection, repair and storage. (5) Written procedures for proper use of respiratory pro-tective equipment including provisions for planned limitations on working times as necessitated by operational conditions.
TS 6.4-6 2-1-72 (6) Bioassays and/or whole body counts of individuals, and other surveys, as appropriate, to evaluate individual exposures and to assess protection actually afforded. e) Equipment approved by the U. s. Bureau of Mines is used. Equipment not approved under u. s. Bureau of Mines Approval Schedules may be used only if the equipment has been evaluated and it can be shown on the basis of reliable test information or manufacturer's data, that the material and performance characteristics of the equipment are at. least equal to those afforded by U. S. Bureau of Mines approved equipment of the same type. f) Protection factors in excess of those specified in B.4 of this specification shall not be used.
- 4. The type of respiratory protective equipment shall be selected and used based on the protection factor required to satisfy the limits specified in B.2 above. No protection factor is to be used to reduce the concen-tration of tritium oxide or noble gases. The protection factor to be used on other particulates, vapors, and gases are delineated in.Table TS 6. 4-1.
- 5. Specification B.2, 3. and 4 shall not preclude the use of respiratory protective equipment in emergencies.
- 6. Specification B.2, 3, and 4 shall be superceded by changes to 10 CFR 20.103.
C. All procedures described in A and B above, and changes thereto, shall be ~ reviewed by the Station Safety and Operating Committee and approved by the Station Manager prior to implementation.
TS 6. 4-7 2-1-72 e D, All procedures described in A and B above shall be followed. E. Temporary changes to procedures which do not change the intent of the original procedure may be made, provided such changes are approved by the Operating Supervisor or Superintendent Station Operation. Such changes will be documented and subsequently reviewed by the Station Nuclear Safety and Operating Committee and approved by the Station Manager. F. Practice of site evacuation exercises shall be conducted annually, following emergency procedures and including a check of communications with off-site report groups. An annual review of the emergency plan will be performed. G. The industrial security program which has been established for the station e shall be implemented, and appropraite investigation and/or corrective action shall be taken if the provisions of the program are violated. An annual review of the program shall be performed.
TABLE TS 6.4-1 PROTECTION FACTORS FOR RESPIRATORS PROTECTION FACTORS 1/ PARTICULATES AND VAPORS AND GASES EXCEPT DESCRIPTION MODES.1f TRITIUM OXIDE~/ I. AIR-PURIFYING RESPIRATORS Facepiece, half-mask 4/ 6/ NP 5 Facepiece, full 6/ - - NP 100 II. ATMOSPHERE-SUPPLYING RESPIRATOR
- 1. Air Line respirator Facepiece, half-mask CF 100 Facepiece, full CF 1,000 Facepiece, full 6/ D 500 Facepiece, Full - PD 1,000 Hood CF 5/
Suit CF ii
- 2. Self-contained breathing apparatus (SCBA)
Facepiece, full 2./ D 500 Facepiece, full PD 1,000 Facepiece, full R 1,000 H III. COMBINATION RESPIRATOR N I C/l Any combination of air- Protection factor for I-' I *"
....., .i::--
purifying and atmosphere- type and mode of operation NI supplying respirator as listed above. CX>
TS 6.4-9 2-1-72 FOOTNOTES TO TABLE TS 6,4-1 ];/ See the following symbols: CF: continuous flow D : demand NP: negative pressure (i.e., negative phase during inhalation) PD: pressure demand (i.e., always positive pressure) R: recirculating (closed circuit) J:._/ (a) For purposed of this specification the protection factor is a measure of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radio-active material outside the respiratory protective equipment to that inside the equipment (usually inside the facepiece) under conditions of use. It is applied to the ambient airborne concentration to estimate the concentration inhaled by the wearer according to the following formula: Ambient Airborne Concentration Concentration Inhaled= Protection Factor (b) The protection factors apply: (i) only for trained individuals wearing properly fitted respirators used and maintained under supervision in a well-planned respiratory protective program. (ii) for air-purifying respirators only when high efficiency [above 99.5% removal efficiency by U. S, Bureau of Mines type dioctyl phthalate (DPO) test] particulate filters and/or sorbents appropriate to the hazard are used in atmospheres not deficient in oxygen. (iii) for atmosphere-supplying respirators only when supplied with adequate respirable air. ]./ Excluding radioactive contaminants that present an absorption or submersion hazard. For tritium oxide approximately half of the intake occurs by absorption through the skin so that an overall protection factor of not more than approximately 2 is appropriate when atmosphere-supplying respirators are used to protect against tritium oxide. Air-purifying respirators are not recommended for use against tritium oxide. See also footnote J_/, below, concerning supplied-air suits and hoods.
TS 6.4-10 2-1-72
!!_I Under chin type only. Not recommended for use where it might be possible for the ambient airborne concentration to reach instantaneous values greater than 50 times the pertinent values in Appendix B, Table I, Column 1 of 10 CFR, Part 20.
J./ Appropriate protection factors must be determined taking account of the design of the suit or hood and its permeability to the contaminant under conditions of use. No protection factor greater than 1,000 shall be used except as authorized by the Commission.
!}../ Only for shaven faces.
NOTE 1: Protection factors for respirators, as may be approved by the U. S. Bureau of Mines according to approval schedules for respirators to protect against airborne radionuclides, may be used to the extent that they do not exceed the protection factors listed in this Table. The pro-tection factors in this Table may not be appropriate to circumstances where chemical or other respiratory hazards exist in addition to radioactive hazards. The selection and use of respirators for such circumstances should take into account approvals of the U. S. Bureau of Mines in accordance with its applicable schedules. NOTE 2: Radioactive contaminants for which the concentration values in Appendix B, Table 1 of this part are based on internal dose due to inhalation may, in addition, present external exposure hazards at higher e concentrations. Under such circumstances, limitations on occupancy may have to be governed by external dose limits.
TS 6.5-1 2-1-72 - 6.5 STATION OPERATING RECORDS Specification A. Records and logs relative to the following items shall be retained for 5 years,. unless a longer period is required by applicable.regulations.
- 1. Records of normal plant operation, including power levels and periods of operation at each power level.
- 2. Records of principle maintenance activities, including inspection, repair, substitution or replacement of principle items of equipment pertaining to nuclear safety.
- 3. Record of abnormal occurrences.
- 4. Record of periodic checks, inspections and calibrations performed to verify that surveillance requirements are being met.
- 5. Records of any special reactor test or experiments pursuant to 10 CFR 50.59.
- 6. Records of changes made in the Operating Procedures pursuant to 10 CFR 50.59.
- 7. Records of shipment of radioactive material.
TS 6.5-2 2-1-72 B. Records relative to the following items shall be retained for the life of the plant.
- 1. Records of changes made to the plant and plant drawings as described in the FSAR pursuant to 10 CFR 50.59.
- 2. Records of new and spent fuel inventory and assembly histories.
3, Records of plant radiation and contamination surveys.
- 4. Records of off-site environmental monitoring surveys.
5, Records of radiation exposure of all plant personnel, and others as required by 10 CFR 20.
- 6. Records of radioactivity in liquid and gaseous wastes released to the environment.
TS 6.6-1 2-1-72 6.6 STATION REPORTING REQUIREMENTS Specification A. Routine Operating Reports - A routine operating report shall be submitted in writing to the Director, Division of Reactor Licensing, U.S. Atomic Energy Commission, Washington, D. C. 20545 at the end of each six month period or fraction thereof terminating on June 30 and December 31. Such reports are due within 60 days after the end of each reporting period. The following information, summarized on a semi-annual basis, except as noted,shall be provided:
- 1. Operations Summary A summary of operating experience occurring during the reporting period that relates to the safe operation of the facility, including a summary of:
- a. changes in facility design,
- b. performance characteristics (e.g., equipment and fuel performance),
- c. changes in procedures which were necessitated by (a) and (b) or which otherwise were required to improve the safety of operations,
- d. results of surveillance tests and inspections required by the licensee's technical specifications,
- e. the results of any periodic containment leak rate tests performed during the reporting period,
- f. a brief summary of those changes, tests and experients
.requiring authorization from the Commission pursuant to 10 CFR 50.59(a), and
TS 6.6-2 2-1-72
- g. any changes in the plant operating organization which involve positions for which minimum qualifications are specified in the technical specifications.
- 2. Power Generation A summary of power generated during the reporting period and the cumulative total outputs since initial criticality, including:
- a. gross thermal power generated (in MWH)
- b. gross electrical power generated (in MWH)
- c. net electrical power generated (in MWH)
- d. number of hours the reactor was critical
- e. number of hours the generator was on-line
- f. histogram of thermal power vs. time
- 3. Shutdowns Descriptive material covering all outages occurring during the reporting period. For each outage, information shall be provided on:
- a. the cause of the outage,
- b. the method of shutting down the reactor; e.g., scram, automatic rundown, or manually controlled deliberate shutdown,
- c. duration of the outage,
- d. plant status during the outage; e.g., cold shutdown or hot standby,
- e. corrective action taken to prevent repetition, if appropriate.
J
TS 6.6-3 2-1-72
- 4. Maintenance A discussion of safety-related maintenance (excluding pre-ventive maintenance) performed during the reporting period on systems and components that are designed to prevent or mitigate the consequences of postulated accidents or to prevent the release of significant amounts of radioactive material. Included in this category are systems and components which are part of the reactor coolant pressure boundary defined in 10 CFR ~ 50.2(v), part of the engineered safety features, or associated service and control systems that are required for the normal operation of engineered safety features, part of any reactor protection or shutdown system, or part of any radioactive waste treatment handling and disposal system or other system which may contain significant amounts of radioactive material. For any malfunction for which corrective maintenance was required, information shall be provided on:
- a. The system or component involved,
- b. the cause of the malfunction,
- c. The results and effect on safe operation,
- d. corrective action taken to prevent repetition,
- e. precautions taken to provide for reactor safety during repair
- f. time required for performing maintenance
TS 6.6-4 2-1-72
- 5. Changes, Tests and Experiments A summary of all changes in the facility design and procedures that relate to the safe operation of the facility shall be included in the Operations Summary section of the semi-annual report. Changes, tests and experiments performed during the reporting period that require authorization from the Commission pursuant to 10 CFR 50.59(a) and those changes, tests and experiments that do not require Commission authorization pursuant to§ 50.59(a) should be addressed. The report shall include a brief description and a summary of the safety evaluation for those changes, tests, and experiments, carried out without prior Commission approval, pursuant to the requirements of e § 50.59(b) of the Commission's regulations, that "The licensee shall furnish to the Commission, annually or at such shorter intervals as may be specified in the license, a report containing a brief description of such changes, tests and experiments, including a summary of the safety evaluation of each."
- 6. Radioactive Effluent Releases Data shall be reported in the form given in Appendix A of U.S.A.E.C. Safety Guide No. 21 for water cooled nuclear power plants, entitled "Measuring and Reporting of Effluents from Nuclear Power Plants," dated December 29, 1971, or in equivalent form. Effluent data ~hall be summarized on a monthly basis except that when the majority of the activity is released as batches and there are less than 3 batches per
TS 6.6-5 2-1-72 month, each batch shall be reported. Estimates of the error associated with each six month to*tal shall be reported. Specifically, the following data shall be reported.
- a. Gaseous Releases (1) total radioactivity (in curies) releases of noble and activation gases.
(2) maximum noble gas release rate during any one-hour period. (3) total radioactivity (in curies) releases, by nuclide, based on representative isotopic analyses performed. (4) percent of technical specification limit.
- b. Iodine Releases (1) total (I-131, I-133, I-135) radioactivity (in curies) released.
(2) total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed. (3) percent of technical specification limit.
- c. Particulate Releases (1) gross radioactivity (8,i) released (in curies) excluding background radioactivity.
(2) gross alpha radioactivity released (in curies) excluding background radioactivity. (3) total radioactivity released (in curies) of nuclides with half-lives greater than eight days. (4) percent of technical specification limit.
TS 6.6-6 2-1-72
- e. d. Liquid Releases (1) gross radioactivity (S,i) released (in curies) and average concentration released to the un-restricted area.
(2) total tritium and alpha radioactivity (in curies) released and average concentration released to the unrestricted area. (3) total dissolved gas radioactivity (in curies) and average concentration released to the unrestricted area. (4) total volume (in liters) of liquid waste released. (5) total volume (in liters) of dilution water used prior to release from the restricted area. (6) the maximum concentration of gross radioactivity (t!,J) released to the unrestricted area (averaged over the period of release). (7) total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed. (8) percent of technical specification limit for total activity released.
- 7. Solid Radioactive Waste ( Summarized Monthly)
- a. Total amount of solid waste packaged (in cubic feet).
- b. Estimated total radioactivity (in curies) involved.
- c. Dates of shipment and disposition (if shipped off-site)
- 8. Fuel Shipments Information relative to each shipment of new and spent fuel shall be provided, including the following:
TS 6.6-7 2-1-72
- a. Date of shipments
- b. Number of elements shipped
- c. Identification number and enrichment of elements shipped.
- d. Activity level at surface of each shipping cask con-taining spent fuel.
9, Environmental Monitoring a) Descriptive material covering the off-site environmental surveys performed during the reporting period including information on: (1) The number and types of samples taken; e.g., air, soil, fish, etc. (2) The number and types of measurements made; e.g. dosimetry. (3) Location of the sample points and monitoring stations. (4) The frequency of the surveys (5) A summary of survey results, including: (a) number of locations at which activity levels are found to be significantly greater then local backgrounds. (b) highest, lowest, and the annual average con-centrations or levels of radiation for the sampling point with the highest average and description of that point with respect to the site.
TS 6.6-8 2-1-72 b) If levels of station contributed radioactive materials in environmental media indicate the likelihood of public intakes in excess of 3% of those that could result from continuous exposure to the concentration values listed in Appendix B, Table II of 10 CFR 20, estimates of the likely resultant exposure to individuals and to population groups, and assumptions upon which estimates are based shall be provided. c) If a particular sample or measurements indicate statistically significant levels of radioactivity above established or concurrent backgrounds, the following information shall be provided: (1) The type of analysis performed; e.g.,alpha, beta, gama and/or isotopic (2) The minimum sensitivity of the monitoring system. (3) The measured radiation level or sample concentration. (4) The specific times when samples were taken and measurements were made. (5) An estimate of the likely resultant exposure to the public if it exceeds 10 mrem. B. Non-Routine Reports
- 1. Abnormal Occurrence Reports~ A notification shall be made within 24 hours by telephone or telegraph to the Director, Region II Compliance Office, followed by a written report within 10 days to the Director, Division of Reactor Licensing (carbon copy to the Director, Region II Compliance Office) in the event of an abnormal occurrence. Abnormal occurrences are defined in Section 1 of these
TS 6.6-9 2-1-72 Technical Specifications. The written report on abnormal occurrences, and to the extent possible, the preliminary telephone or telegraph notification, shall:
- a. describe, analyze and evaluate safety implications,
- b. outline the measures taken to assure that the cause of the condition is determined, and
- c. indicate the corrective action(including any changes made to the procedures and to the quality assurance program) taken to prevent repetition of the occurrence and of similar occurrences involving similar components or systems.
- d. r~late any failures or degraded performance of systems and components for the incident to similar equipment failures that may have previously occurred at the facility. The evaluation of the safety implications of the incident should consider the cumulative experience obtained from the record of previous failures and mal-functions *of the affected systems and components or of similar equipment.
- 2. Reporting of Unusual Safety Related Events A written report shall be forwarded within 30 days to the Director, Division of Reactor Licensing and to the Director, Region II Compliance Office in the event of an unusual safety related event. Unusual safety related events are defined in Section 1 of these Technical Specifications.
TS 6.6-10 12-29-71 e 3. Special Reports
- a. Startup Report - A summary report of unit startup and power escalation testing and the evaluation of the results from these test programs shall be submitted when a unit is initially placed in service and when the unit has been modified to an extent that the nuclear, thermal, or hydraulic perform~nce of the unit may be significantly altered. The test results shall be compared with design predictions and specifications. Startup reports shall be submitted within 60 days following commencement of commercial power operation (i.e., following synchronization of the turbo-generator to produce commercial power).
- b. First Year Operation Report - A report shall be submitted within 60 days after completion of the first year of operation (the first year begins with the synchronization of the turbogenerator to produce commercial power). This report may be incorporated into the semiannual operating report and shall cover the following:
(1) an evaluation of plant performance to date in comparison with design predictions and specifications (2) a reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analyses. (3) an assessment of the performance of structures,systems and co~ponents important to safety (4) a progress and status report on any items identified as requiring additional information during the operating
TS 6. 6-11 12-29-71 license review or during the startup of the facility, including items on which additional information was required as conditions of the license and items identi-fied in the licensee's startup report
- c. Containment Leak Rate Test - Each containment integrated leak rate test shall be the subject of a summarv technical report. Upon completion of the initial containment leak rate test specified by proposed Appendix J to 10 CFR 50, a special report shall, if that Appendix is adopted as an effective rule, be submitted to the Director, Division of Reactor Licensing, USAEC, Washington, D.C. 20545,. and other containment leak rate tests specified by Appendix G that fail to meet the acceptance criteria of the appendix, shall be the subject of special summary technical reports pursuant to Section V.B of Appendix J, The requirements of Section V.B are as follows:
"B. Report of Test Results The initial Type A test shall be subject of a summary technical report submitted to the Commission approximately 3 months after the conduct of the test.
This report shall include a schem~tic arrangement of the leakage rate measurement system, the instrumenta-tion used, the supplemental test method, and the test program selected as applicable to the initial test, and all subsequent periodic tests. The report shall contain an analysis and interpretation of the leakage rate test data to the extent necessary to demonstrate the acceptability of the containment's leakage rate in
TS 6,6-12 12-29-71 meeting the acceptance criteria."
"For periodic tests, leakage rate results of Type A, B, and C tests that meet the acceptance criteria of Sections III.A.7, III.B.3, and III.C.3 respectively shall be reported in the licensee's periodic operating report. Leakage test results of Type A, B, and C tests that fail to meet the acceptance criteria of Sections III.A.7, III.B.3, and III.C.3 respectively shall be reported in a separate summary report.that includes an analysis and interpretation of the test data, the least-squares fit analysis of the test data, the instru-ment error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria.
Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate-test measurements shall also be included."
- d. In-service Inspection Evaluation - A special summary technical report shall be submitted to the Director, Division of Reactor Licensing, USAEC, Washington, D.C. 20545 afte'r 5 years of operation.
This report shall contain an evaluation of the results of the in-service inspection program and will be reviewed in light of the technology available at that time.
- e. Initial Containment Structura+ Test - A special summary technical report shall be submitted to the Director, Division of Reactor Licensing, USAEC, Washington, D.C., 20545 within 3 months after e completion of the test. This report will include a summary of the
/
I I TS 6.6-13 ..I 12-29-71 measurements of deflections, strains, cr~ck width, crack patterns observed, as well as comparisons with predicted values of acceptance criteria.
- 4. Other Reporting Requirements In addition to the report requirements listed in these Technical Specifications, there are specific reporting requirements included in 10 CFR 20, 40, 50, 70 and 73. For convenience, these specific reporting requirements are tabulated in Table TS 6.6-1. This table is presented only as a reporting sununary, and since it can be expected that the regulations may change with time, the applicable regulation of date as presented in the Code of Federal Regulations shall supercede those presented in Table TS 6.6-1.
- APPENDIX B SEISMIC DESIGN FOR THE NUCLEAR STEA}! SUPPLY SYSTEM B.l GENERAL SEISMIC DESIGN CRITERIA FOR THE NUCLEAR STEAM SUPPLY SYSTEM B.2 SEISMIC DESIGN CRITERIA FOR PIPING, VESSELS, SUPPORTS, AND REACTOR VESSEL INTEP~ALS B.2.1 PIPING, VESSELS, AND SUPPORTS B.2.2 REACTOR VESSEL INTERNALS B.2.2.1 Design Criteria for Normal Operation B.2.2.2 Design Criteria for Abnormal Operation B.3 GENERAL ANALYTICAL PROCEDURE FOR SEISMIC DESIGN B.4 MOVEMENT OF REACTOR COOLANT SYSTEM COMPONENT B.5 TESTS TO DEMONSTRATE THE CONSERVATISM OF THE LIMIT CURVES
B.1-1 2-13-70
- APPENDIX B SEISMIC DESIGN FOR THE NUCLEAR STEAM SUPPLY SYSTEM B.l GENERAL SEISMIC DESIGN CRITERIA FOR THE NUCLEAR STEAM SUPPLY SYSTEM All Class I components of the Nuclear Steam Supply System are designed in accordance with the following criteria:
- 1. Primary operating stresses, when combined with the Operational Basis Earthquake seismic stresses resulting from a dynamic analysis using a response spectrum normalized to a maximum horizontal ground acceleration of 0.07g and a simultaneous vertical ground acceleration of 2/3 the horizontal are maintained within the allowable stress limits in Table B.2-1 of this appendix.
- 2. Primary operating stresses when combined with the Design Basis Earthquake seismic stresses resulting from.a dynamic analysis using a response spectrum normalized to a maximum horizontal ground acceleration of 0.15g and a simultaneous vertical ground acceleration of 2/3 of the horizontal are limited so that the function of the component or system shall not be impaired as to prevent a safe and orderly shutdown of the unit. Further, the primary operating stresses are maintained within the allowed stress limits in Table B.2-1.
No loss of. function requires that rotating equipment will not
- seize, pressure vessels will not rupture, supports will not
B.1-2 2-13-70 collapse or deform to such a degree as to cause failure of the supported equipment. In addition, systems required to be leak tight will remain leak tight and engineered safeguards will perform intended functions.
B.2-1 2-13-70
- B.2 SEISMIC DESIGN CRITERIA FOR PIPING, VESSELS, SUPPORTS AND REACTOR VESSEL INTERNALS Reference is made in this section of the Appendix to WCAP-5890, Rev. 1, "Ultimate Strength Criteria to Ensure No Loss of Function of Piping and Vessels under Earthquake Loading," by R. A. Wiesemann, R. E. Tome and R. Salvatori. Following discussions with the AEC DRL Staff during the Construction Permit Application Review for Diablo Canyon I, the criteria presented in WCAP-5890, Rev, 1, for the generation of limit curves were modified. Details of the manner in which this modification were developed are given in Note 1 of Section B.5.
The loading conditions and stress limits which are employed in the design of Class I piping, vessels, supports, and other pertinent components are shown in Table B.2-1. This Table also indicates the allowable stress limits which are used in the design of the components for the various load-ing combinations. To be able to perform their function, i.e., allow core shutdown and cooling, the reactor vessel internals must satisfy deformation limits. For this reason the reactor vessel internals are treated separately in Section B.2.2. B.2.1 PIPING, VESSELS AND SUPPORTS The reasons for selection of the above mentioned loading conditions and
- allowable stress limits are as follows:
B.2-2 2-13-70
- 1. When subjected to the Operational Basis Earthquake, the Nuclear Steam Supply System is designed to be capable of continued safe operation.
Equipment and supports needed for this purpose are required to operate within normal design limits as shown in Line 2 of Table B.2-1. The load combination corresponding to the upset loading condition in line 2 is the normal plus Operational Basis Earthquake load.
- 2. In the case of the Design Basis Earthquake 9 it is necessary to ensure that components required to shut the unit down and maintain it in a safe shutdown condition do not lose their capability to perform their safety function. This capability is ensured by maintaining the stress limits as shown in Line 3 of Table B.2-1.
- 3. For the highly unlikely but postulated case of reactor coolant pipe rupture, stresses in the unbroken leg of the affected loop and components of the unaffected loops of the reactor coolant system will be as noted in Line 4 of Table B.2-1. The load combination corresponding to the faulted loading condition in line 4 is the Design Basis Earthquake and/or Design Basis Accident load.
- 4. For the extremely remote event of simultaneous occurrence of a Design Basis Earthquake and reactor coolant system pipe rupture,
~
B.2-3 2-13-70
- the Class I piping and-components, excluding the broken pipe, are checked for no loss of function, i.e., the capability to contain fluid and allow fluid flow and perform vital engineered safeguards functions. This is assured by limiting the various stress combina-tions within the limits shown in Line 4 of Table B.2-1.
The minimum margin of safety between the design limit stress and the expected collapse condition is for the case of pure tension and is defined as s u 1 t1mate
. - sd es1gn sd es1gn Under more realistic operating conditions, piping and vessels will always experience some combination of tension and bending. For these combined load cases the margin of safety is greater than that for pure tension, as shown by the limit curves contained in WCAP-5890, Rev. 1 and shown in Figures B.5-1 and B.5-2. Therefore, it is conservative to base the margin of safety on pure tension. Table B.2-2 illustrates the margin to safety between the stress limits for various load conditions and the expected failure or collapse condition for typical materials.
Plastic or limit analysis conducted within the limits of the faulted condition will be performed considering plastic material behavior, including as required modifications of material stiffness characteristics, formation of plastic hinges and other non-linear effects as determined in detailed structural analysis and will be provided in standard stress reports
- B.2-4 12-13-70 TABLE B.2-1 LOADING CONDITIONS AND STRESS LIMITS DEFINITIONS*
- 1. "Normal Conditions. Any condition in the.course of system startup, operation in the design power range, and system shutdown, in the absence of Upset, Emergency or Faulted Conditions.
- 2. "Upset Conditions. Any deviations from Normal Conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The Upset Conditions include those transients which result from any single operator error or control malfunction, transients caused by a fault in a system component requiring its isolation from the system, transients due to loss of load or power, and any system upset not resulting in a forced outage. The estimated duration of an Upset Condition shall be included in the Design Specifications--The Upset Conditions include the effect of the Operational Basis Earthquake for which the system must remain operational or must regain its operational status.
'I .J. "Emergency Conditions. Any deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system.. The conditions have a low probability of occurrence
- Summer 1968 Addenda to the ASME B&PV Code-Nuclear Vessels, Section III.
B.2-5 2-13-70
- but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any damage developed in the system.
- 4. "Faulted Conditions. Those combinations of conditions associated with extremely low probability postulated events whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent where considerations of public health and safety are involved, Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities.
Among the Faulted Conditions may be a specified Design Basis Earthquake for which safe shutdown is required."
B.2-6 2-13-70 TABLE B.2-1 (continued)
- LOADING CONDITIONS AND STRESS LIMITS: PRESSURE VESSELS Loading Conditions Stress Limits Note
- 1. Normal Conditions (a) p <S m- m (b) P (or P )+PB<l.5Sm 1 m 1 (c) P (or P )+PB+Q.::_3.0Sm 2 m 1
- 2. Upset Conditions (a) p <S m- m (b) P (or P )+P B.::_1. 5Sm 1 m 1
- 3. Emergency Conditions (c) P (or P )+PB+Q.::_3.0Sm m
(a) P <l. 2S m-1 m' or 2 p <S
?1 y whichever is larger (b) P (or P )+PB.::_1.5(1.2Sm) or 3 m 1 p (or P )+P B.::_1. 5 (Sy) 3 m 1 whichever is larger
- 4. Faulted Conditions Design Limit Curves of 4 WCAP-5890, Rev. 1 as Modified by Note 1 of This Appendix
B.2-7 2-13-70 where: P m
= primary general membrane stress intensity P = primary local membrane st~ess intensity 1
PB= primary bending stress intensity Q = secondary stress intensity S m
= stress intensity value from ASME B&PV Code, Section. III, Nuclear Vessels.
S y
= minimum specified material yield
_j_ . _ __i.__
B.2-8 9..;.15-71 TABLE B.2-1 (continued) LOADING CONDITIONS AND STRESS LIMITS: PRESSURE PIPING Loading Conditions Stress Limits
- 1. Normal Conditions p < s m-
- 2. Upset Conditions P < l.2S nr-
- 3. Emergency Conditions Pm -.
< l.8S
- 4. Faulted Conditions Design Limit Curves of Note 4 WCAP-5890, Rev. 1 as Modified by Note 1 of
. this Appendix where:
p principal stress m s = allowable stress from USAS B31.l, Code for Power Piping
B.2-9 2-13-70 TABLE B.2-1 (continued) LOADING CONDITIONS AND STRESS LIMITS: EQUIPMENT SUPPORTS Loading Conditions Stress Limits
- 1. Normal Conditions Within working limits
- 2. Upset Conditions Within working limits
- 3. Emergency Conditions Within material yield strength after load redistribution(l)
- 4. Faulted Conditions Within material yield strength after load redistribution(l)
Note (1) Higher stress values can be adopted if a valid limit or plastic instability analysis of the support and supported component/system is performed
- B.2-10 2-B-10' Note 1:
NOTES FOR TABLE B.2-1 The limits on local membrane stress intensity (PL< 1.5S) and primary membrane plus primary bending stress intensity m (PM (or PL)+ PB.::_ 1.5Sm) need not be satisfied at a specific location if it can b~ shown by means of limit analysis or by tests that the specified loadings do not exceed 2/3 of the lower bound collapse load as per paragraph N- 417.6(b) of the ASME B&PV Code, Section III, Nuclear Vessels. Note 2: In lieu of satisfying the specific requirements for the local membrane (PL.::_ l.5Sm) or the primary plus secondary stress intensity (PL+ PB+ Q .::_ 3Sm) at a specific location, the structural action may be calculated on a plastic basis and the design will be considered to be acceptable if shakedown occurs, as opposed to continuing deformation, and if the deformations which occur prior to shakedown do not exceed specified limits, as per paragraph N- 417.6(a)(2) of the ASME B&PV Code, Section III, Nuclear Vessels. Note 3: The limits on local membrane stress intensity (PL< l.5S) and
- m primary membrane plus primary bending stress intensity (Pm (or PL)+ PB.::_ l.5Sm) need not be satisfied at a specific location if it can be shown by means of limit analysis or by test that the specified loadings do not exceed 120 per cent of 2/3 of the lower bound collapse load as per paragraph N- 417.lO(c) of the ASME B&PV Code, Section III, Nuclear Vessels.
B.2-11 2-13-70
- Note 4: As an alternate to the design limit curves which represent a pseudo plastic instability analysis, a plastic instability analysis may be performed in some specific cases considering the actual strain-hardening characteristics of the material, but with the yield strength adjusted to correspond to the tabulated value at the appropriate temperature in Table N- 424 or N- 425, as per paragraph N- 417.llc of the ASME B&PV Code, Section III, Nuclear Vessels. These specific cases will be justified on an individual basis
- B.2-12 4-13-70 TABLE B.2-2 MINIMUM MARGINS OF SAFETY LOADING CONDITIONS Upset Emergency Faulted Material Conditions Conditions Conditions*
SA302 Gr. B 200% 150% 27% Inconel 600 316 SST A212 Gr. B 228% 222% 346% 172% 169% 272% 43% 60% 55%
- Based upon the limit curves computed using Note 1 of Section B.5.
B.2-13 2'."'"13-70
- B.2.2 B.2.2.1 REACTOR VESSEL INTERNALS Design Criteria for Normal Operation The internals and core are designed for normal operat:ing conditions and subjected to loads of mechanical, hydraulic, and thermal origin. The response of the structure under the uperational Basis Earthquake is included in this category as well as operational transients (upset conditions).
The stress criteria established in Section III of the ASME Boiler and Pressure Vessel Code, Article 4, has been adopted as a guide for the design of the internals and core with exception of those fabrication
- techniques and materials which are not covered by the Code, such as the fuel rod cladding. Seismic stresses are combined in the most conservative way and are considered primary stresses.
The members are designed under the*basic principles of: (1) maintaining deflections within acceptable limits, (2) keeping the stress levels within acceptable limits, and (3) prevention of fatigue failures. B.2.2.2 Design Criteria for Abnormal Operation The abnormal design condition assumes blowdown effects due to a pipe break combined in the most unfavorable manner with the effects associated with the Desi.gn Basis Earthquake .
B.2-14 2-13-70 For this condition the criteria for acceptability are that the reactor be capable of safe shutdown and that the engineered safety features are able to operate as designe,d. Consequently, the limitations established on the internals for these types of loads are concerned principally with the maximum allowable deflections. Additional stress criteria for critical structures under normal operation, plus the Design Basis Earthquake and blowdown excitation, assures that the structural integrity of the components is maintained.
B.3-1 2-13-70
- B.2 GENERAL ANALYTICAL PROCEDURE FOR SEISMIC DESIGN The design and analysis of Class I components of the Nuclear Steam Supply System utilizes the "response spectrum" approach.
The dynamic analysis is performed using the normal mode methods. The inertial properties of the models are characterized by the mass, and mass moment of inertia of each mass point. The stiffness properties are characterized by the moment of inertia, area, shear shape factor, torsion constant, Young's modulus, and shear modulus. Table 15.2.4-1 gives the damping ratios to be used in the dynamic analysis of components .
B.4-1 2.,..13-70
- B.4 MOVEMENT OF REACTOR COOLANT SYSTEM COMPONENTS The criterion for movement of the reactor pressure vessel, under the worst combination of loads, i.e., normal plus the Design Basis Earthquake plus reactor coolant pipe rupture loads, is that the movement of the reactor vessel will not exceed the cleargnce between a reactor coolant pipe and the surrounding concrete.
The relative motions between Reactor Coolant System components will be controlled by the structures which are used to support the reactor pressure vessel, steam generators, pressurizer and reactor coolant pumps .
B.5-1 2-13-70
- B.5 TESTS TO DEMONSTRATE THE CONSERVATISM OF THE LIMIT CURVES Tests performed at Westinghouse Material Testing Laboratory in Pittsburgh demonstrate the conservatism of the limit curves presented in WCAP 5890, Rev. 1. Carbon steel and* stainless steel pipes have been tested under various combinations of axial and transverse loads to determine failure loads. Specimens, about 1.5 foot long have been cut from 1.5 inch nominal diameter Schedule 160 pipes. The materials employed were SA 106B carbon steel and Type 304 stainless steel. These specimens were kept internally pressurized to 3000 psia for the entire duration of the tests. Tables B.5-1 and B.5-2 summarize the tests that have undergone evaluation and the results of this evaluation .
TABLE B.5-1 TESTS AND TEST RESULTS ON SA 106B CARBON STEEL PIPE SPECIMENS (internal pressurization= 3000 psia) Pseudo-elastic Pseudo-elastic Pseudo-elastic Axial Stress Bending Stress Stress Intensity Strain Normalized to the Normalized to the Normalized to the % Yield Stress Yield Stress Yield Stress (gage length) Pure Tension 1-. 736 0.0 1. 770 22.61% (12") (no weld) 1.840 0.0 1.847 22.32% (12") Pure Bending* 0.10 >2.348 >2.382 >35.4% (1 ") Tension+ >1.375 >1.030 >2.440 >7.75% (6") Bending* >l.585 >0.565 >2.180 (no weld) >1. 845 >0.266 >2.145 Compression+ Bending* >-1.130 1.08 2.410 (no weld) Pure Tension (circumf. weld) 1. 852 0.0 1.886 20.05% (12") Pure Bending* (circumf. weld) 0.10 >2.580 *>2 .. 614 >30 .19% (1. 5")
~ td Pure Bending** f--1.
w Ul (rejected 0.10 2.460 2.494 14. 51% (2 ") I
-..J N I
circum. weld) 0
- ** See following page .
- TABLE B.5-1 (Continued)
- The limit capability of the test apparatus has been reached before failure of these specimens was approached.
** This test was performed on a welded pipe specimen which has been rejected by the inspector prior to the test for gross lack of penetration in weld. This was the only test in which the weld failed. The specimen exhibited substantial ductility prior to failure .
TABLE B.5-2 TESTS AND TEST RESULTS ON 304 STAINLESS STEEL SPECIMENS (internal pressurization= 3000 psia) Pseudo-elastic Pseudo-elastic Pseudo-elastic Axial Stress Bending Stress Stress-Intensity Normalized to the Normalized to the Normalized to the Yield Stress Yield Stress Yield Stress Strain Pure Tension 2.495 0.0 2.53 52.1% (no weld) (12" gage length) Pure Bending* +0.10 >* 2.91 >2.945 >30.0% (no weld) (l" gage leng.th) Tension+ >+2.09 >* 0. 42 >2.55 Bending* (no weld) Pure Bending +0.10 > 3.27 >3.30 >25.0% (with circumf. weld) (1.5" gage length) Pure Tension (with circ.umf. +2.46 0.0 2.49 44.6% weld) (12" gage length) t;" ~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-t;, ~
!.r ~
0
- The limit capability of the test apparatus has been reached before failure of these specimens was approached .
B.5-5 2-13-70
- Standard ASTM tensile specimens have been modeled from pieces of the test pipes and stress-strain curves determined. These curves have been conservatively approximated with trapezoidal stress-strain curves as indicated in WCAP-5890, Rev. 1. The limit curves for both SA 106B carbon steel and Type 304 stainless steel for the test conditions, have been calculated and are reported in Figures B.5-1 and B.5-2, respectively. The experimental points, i.e., stress intensities versus axial stress as listed in Table B.5-1 and B.5-2 are shown in Figures B.5-1 and B.5-2. Also shown in these figures are the limit curves as calculated by the use of the trapezoidal stress-strain curves up to the ultimate stress. Comparisons between the experimental points and the design limit curves show the conservatism of the latter .
- Note 1 WCAP-5890, Rev. 1 will be replaced in the near future by WCAP-7287. The revisions will affect limits for the combination of normal loads plus Design Basis Earthquake loads plus pipe rupture loads associated with a loss of coolant accident. The changes will reflect agreement with the AEC-DPL Staff on the stress limits for the above mentioned load combinations.
Details of the manner in which the revisions will be developed are as follows:
- 1) Use material data to develop stress-strain curves.
- Typical stress-strain curves of Type 304 stainless steel (Figure B.5-3),
Inconel 600 (Figure B.5-4) and SA 302B low alloy steel (Figure B.5-5) at
B.5-6 2-13-70 600°F have been generated from tests using graphs of applied load versus cross-head displacement as automatically plotted by the recorder of the tensile test apparatus. The scale and sensitivity of the test apparatus recorder assure accurate measurement of the uniform strain. For materials other than these three, stress-strain curves developed by conservative use of pertinent available material data (i.e., lowest values of uniform strain and initial strain hardening). Where the available data are not sufficient to develop a reliable stress-strain curve, three standard ASTM tensile tests of the material in question are performed at design temperature. These data are conservatively applied in developing a stress-strain curve as described above. *
- 2) Normalize the ordinate (stress) of the stress-strain curves to the measured yield strength.
- 3) Use 20% of uniform strain as defined on the curve developed under Item 1 as the allowed membrane strain.
- 4) Establish the'normalized stress ratio at 20% of uniform strain on the normalized stress-strain curves developed under Item 2.
- 5) Establish the value of the membrane stress limit.
Multiply the normalized stress ratio in Item 4 by the applicable code
- B.5-7 2-13-70
- yield strength at the design temperature to get the membrane stress limit. As an alternate, the actual physical properties as determined from standard ASTM tensile tests on specimens from the same heats may be used to determine the membrane stress limit. If such an approach is adopted, sufficient documentation will be provided to support the actual material properties used.
- 6) Develop limit curves for the combination of local membrane and bending stresses.
The limit curves will be developed by using the analytical approach presented in WCAP-5890, Rev. 1 and the stress-strain curve up to the
- membrane stress limit as developed under Item 5. It is anticipated that these limit curves will be within the limit curves discussed with the AEC DRL Staff during the meetings of November 30 and December 1, 1967 for the same materials .
MATERIAL: SA 106B Carbo:1 Steel CROSS-SECTION: Hollow-circular ~ Experimental point with weld t:, trl 3.5 HOOP STRESS: 0.2 S EB Experimental point witl10ut \,eld 3.5 C/l y H (K> Experir:1ental point _with rejected weld C") z t"' H
- 3. '3. 0 H -j t-l ROml TE:lPER,\TLRE t,.
C/l U) N() S!IEAR U) LL: 0::: r
- 2. 5 j '**0 5 U) ,,,--
I.
~ limit curve / L w
,_. 2.0 (WCAP 5890 theory)/ ')
- 0 i ._
/
*r-'
H I U) z LI..;
- 1. 5--
'- faulted ~1.5
?:
,_. condition ~
U) U) r'
~-*--rr-7-r:-~~~~
I.,_;
- 1. 0 .. , /
- ,:; f-1. 0 r- emergency ~~--,_~';'
er. ~ , condition 1 0.5 upset 'e O. ..,~
~
-1 condition Cl. o-+*i ,-~--- ---,. ..........- ~ ~ - - - * - - " [ /
, - , ~ - 1 r*-, -- i-- *r -+-
1.
~-r--,.--,---..--..:..,ll.-4--.-,- 1 ----** *.
- r-J_-,-.1 -r-r -.,--,-JO, 0
-2. 5 -2.0 01.5 -1. 0 -0.5 0.0 0.5 1 .0 1.5 2.0 2.5 STRESS FRO:! AXIAL FORCE/YIELD STF.E~;s ( i.1dudin!'-; pressure end force)
L
~ Experimental point with weld
!-iATERIAL: 304 Stainless Steel EB Experimental point without weld CROSS SECTION: Hollow-circular ti HOOP STRESS: 0.2 s I:'!
y en 3.5 3.5 H
~
C') z r-' H ROOM TEMPERATURE NO SHEAR /
/ "-
"" / -- '-
"' l /
limit curve based on X
"<lt~
/ J . ()
H
,-,J 3.0 /
tn
/
/
V / n ~ 0
~
(/) (/) w 2.5 /
/
/
/ ~ ',--l _ -I 2.5
/
( / ~ c:r:: V I:'! E-< \ ti (/)
~ l i m i t curve
,-,J Q
....., (WCAP5890 0
w / theory) H 2.0 2.0 ~ I:'!
- ,ii E-<
H H (/) X r:-, zw 1. 5 !faulted l. 5 ~ ,-,J E-< z Icondition [!:: H I 0 (/) (/) I H w 1.0 Iemergency l. 0 z ,-,J c:r:: E-< en (/) !condition I I 0.5 0.5 lupset lcondition I 0.0 0.0 N lzj IH
-2.5 -2.0 -1. 5 f--1 (j)
-1. 0 -0.5 0.0 0.5 1.0 1. 5 2.0 2.5 w C:
!.i ~
STRESS FROM AXIAL FORCE/YIELD STRESS (including pressure end force) 0 l)j IJI I N
TYPICAL STRESS STRA!tJ CL/RVE. 5 TA ;\JD ARD A 5- T M TE- lv .5 I L £ TE s* T
!v!ATERIAL: 304 STA/1\/LEs*s* STEE-L C) ~3.0 TE !v1PE-RA TURE~ 600 °F Ct u,
V') 2.5 Lu Q f--_ V)zo I N't:rj I H I I-' (j") i/"
*-+,-*-*----**- ---*r ---
~~ U1 0 10 40 l
TYPICAL STRESS STRAIN CURVE 5 T/1.tv'[JARO ASTA!f TENSILE TEST MATERIAL! INCONEL 600 TEMP,ERA TURE: 600 °F 3.0 2.5 () f--..
~
Q:: 2.0 I V') I V) I Lu 1 Cr. ~ 1. f-- 15-~ I I
~ ~
~~
ID~
-j I
I I Q: II (_)
---::, 0.5
<._ I I I
1 ().2 x Eu I ru 1;J Q0*-+--------+- I/,,,;
,-----r-------,--**...L*-~ - .. --, . ..
0 /0 20 30 40 SC STPAI/\, _%
TYPICAL STRESS STRAIN CURVE STANDARD ASTM TENSILE TEST MATERIAL: SA 302 GRADE B 1.5* TEMPERATURE: 600°F () f--
-~--
"-T ct u, J) Lu w (
- )::
f--- V) a Lu
~ ._J --c0.5 ~
C(
~
l;2xfu 'N>:cj IH t-' c;-:i (..,.) c:: O'
.v - - - - , - - - - ---*-*-r- ~:
0 5 10 IS 20 25 - 30 °b:i 0 Ln 5 TRAIN, /o I Ln
C.1-i 2-1-72 APPENDIX C EMERGENCY CORE COOLING SYSTEM PERFORMANCE BASED ON AEC INTERIM CRITERIA
C,1-ii 2-1-72 APPENDIX C C.l EMERGENCY CORE COOLING PERFORMANCE C.2 HEAT TRANSFER COEFFICIENTS USED IN THE LOCTA-R2 CORE THERMAL ANALYSIS C.3 RESIDUAL DECAY HEAT C.4 EMERGENCY CORE COOLING SYSTEM C.5 PEAKING FACTOR C.6 EMERGENCY CORE COOLING PERFORMANCE FOR LOOP OUT OF SERVICE OPERATION
C.1-1 9-15-71 e* APPENDIX C C.1 EMERGENCY CORE COOLING PERFORMANCE This report is in response to information requested by Dr. Peter A. Morris, Director of the Division of Reactor Licensing, in his letter dated July 9, 1971 and covers the analytical techniques described in the topical report "Westinghouse PWR Core Behavior Following a Loss-of-Coolant Accident, "WCAP-7422-L, January 1970 (Proprietary); the supplementary proprietary Westinghouse report "Emergency Core Cooling Performance", dated June 1, 1971; and includes the additional assumptions identified in Appendix A of the AEC Interim Policy Statement, "Criteria for Emergency Core Cooling Systems for Light Water Power Reactors", published in the Federal Register June 29, 1971, Vol. 36, p. 12247. The information requested by the Division of Reactor Licensing is presented on pages C.1-16 and C.1-17. The results of the loss of coolant accident analysis are shown on page C.l-19 and show compliance with the Interim Policy Statement. The additional assumptions identified in Appendix A, Part 3 of Interim Policy Statement, are incorporated in the text describing the cal-culational procedure. These assumptions are repeated below and the pages noted where each is discussed.
- 1. The break discharge coefficient, (CD), used with the Moody discharge flow model should be equal to 1.0 for all break sizes. Page C.l-7
- 2. The decay heat curve described in the proposed ANS Standard, with a 20% allowance for uncertainty, should be used. The fraction of decay heat generated in the hot rod may be considered to be 95% of this value. Section C.3
- 3. For large breaks in the range 0.6 to 1.0 times the total area of the double-ended break of the largest cold-leg pipe, two break models should be used. The first model should be the double-ended severance
("Guillotine"), which assumes that there is break flow from both
C.1-2 9-15-71 ends of the broken pipe, but no communication between the broken ends. The second model should assume discharge from a single node ("split"). Page C.1-15 and C.1-16
- 4. The time after the break for the onset of departure nucleate boiling at the hot spot should be equal to 0.1 second. Page C.1-3
- 5. For cold-leg breaks, all of the water injected by the accumulators prior to end-of-blowdown shall be assumed to be lost. In this context the end-of-blowdown shall be specified as the time at which zero break flow is first computed. The containment back pressure assumed for the blowdown analysis should not be higher than the initial pre-break pressure plus 90% of the increase in pressure calculated for the acci-dent under consideration. Page C.1-9
- 6. The pump resistance, K, used for analysis should be fully justified.
The effect of pump speed upon K should be considered. The more con-servative of two assumptions (locked or running) should be used for the pump during the blowdown calculation. Page C.1-15
- 7. A calculation for the reflooding heat transfer should be performed.
The containment back pressure assumed for the analysis should not be higher than the initial pre-break pressure plus 80% of the increase in pressure calculated for the accident under consideration. Page C.1-12 The following items should be constraints on the calculation:
- a. No steam flow should be permitted in intact loops during the time period that accumulators are injecting. Page C.1-10
- b. Core exit quality should be calculated from entering mass flow rate and nominal FLECHT heat transfer. Page C.1-11
- c. Pump resistance should be calculated on the basis of a locked rotor. Page C.1-10
.e
- d. The effects of the nitrogen gas in the accumulator, which is discharged following accumulator water discharge, should be
C.l-3 9-15-71 taken into account in calculating steam flow as a function of time. Page C.1-13
- e. The pressure drop in the steam generator should be calculated with the existing fluid conditions and associated loss coefficient.
Page C.1-10
- f. All effects of cold injection water, in either a hot or cold leg, on steam flow (and ~P) should be included in the calculation.
Page C.1-13
- g. The heat transfer coefficient during reflood should be derived from FLECHT data. Page C.2-6 PHENOMENOLOGICAL DESCRIPTION OF LOSS OF COOLANT ACCIDENT A LOCA would result from a rupture of the Reactor Coolant System (RCS) or of any line connected to that system up to the first closed valve. The charging pumps have the capability to make up for leakages resulting from ruptures of a small cross section, thus permitting an orderly shutdown. A small quantity of the coolant containing fission products present in the coolant would be released to the containment.
For larger breaks, subcooled fluid is expelled from the break rapidly reducing the pressure to saturation, For a postulated large break, reactor trip is initiated when the pressurizer low pressure set point is reached while the Safety Injection System (SIS) signal is actuated by coincident pressurizer low pressure and low level. The reactor trip and SIS actuation are also initiated by a higher containment pressure signal. The consequences of the accident are limited in two ways:
- 1. Reactor trip and borated water injection supplement void formation in causing rapid reduction of the nuclear power to a residual level cor-responding to the delayed fission and fission product decay.
- 2. Injection of borated water ensures sufficient flooding of the core to prevent excessive temperatures.
C.1-4 9-15-71 At the beginning of the blowdown phase, the entire reactor coolant system contains subcooled liquid which transfers heat from the core by forced con-vection with some fully developed nucleate boiling. One tenth of a second after the break the AEC requires the conservative assumption that departure from nucleate boiling occurs. Thereafter the core heat transfer is unstable with both nucleate boiling and film boiling occurring. As the core becomes uncovered both turbulent and laminar forced convection to steam are considered as the only core heat transfer media. When the RCS pressure falls below 600 psia the accumulators begin to inject borated water. The conservative assumption is maintained that accumulator injection bypasses the core and goes out the break until the termination of the blowdown phase. This conservatism is consistent with the AEC Interim Criteria for ECCS. WESTINGHOUSE PERFORMANCE CRITERIA FOR ECCS The reactor is designed to withstand thermal effects caused by a LOCA including the double ended severance of the largest RCS pipe. The reactor core and internals together with the ECCS are designed so that the reactor can be safely shutdown and the essential heat transfer geometry of the core preserved following the accident. The ECCS, even when operating at partial effectiveness due to loss of one vital power bus (limiting case without off-site power), is designed to limit the cladding temperature to well below the melting temperature of Zircaloy-4 (3375°F) and below the temperature at which gross core geometry distortion, including clad fragmentation, can be expected. In addition, the core metal-water reaction is limited to less than 1% of the available Zircaloy. In order to assure effective cooling of the core, limits on peak clad tempera-ture and local metal-water reaction have been established as defined in Figure C.1-1 in box labeled Safe Operation. It has been demonstrated in the single rod burst test phase of the Westinghouse Rod Burst Program(l) and in the fuel rod quench tests performed by the Argonne National Laboratory (13) that for conditions within the area of safe operation, as shown in Figure C.1-1, fuel rod integrity is maintained. Results of the multi-rod burst test phase of the Westinghouse Rod Burst Program(Z) show that peak clad temperature
C.1-5 9-15-71 calculated during the accident increases less than 100°F due to geometry distortion; thus peak clad temperatures determined on the basis of no geometry distortion should be limited to 100°F below the limits presented in Figure C.1-1. However, the peak clad temperature calculated without geometry dis-tortion is limited to 2300°F consistent with AEC interim criteria thereby providing an additional 300°F safety margin compared to test data. Consequences of the loss of coolant accident are well within the limits of 10CFR100 guidelines. METHOD OF ANALYSIS This analysis is performed by considering the following aspects of the accident: blowdown hydraulics including reactor kinetics, and core thermal transient. a) Blowdown Hydraulics This calculation provides a description of the hydraulic response to a rupture of the RCS through completion of depressurization and the operation of the safety injection systems. The basic information concerning the dynamic behavior of the reactor core environment is thus provided for use in reactor kinetics and core cooling analyses. The code used in this part of the analysis is SATAN-V. The core power transient is forced by the blowdown dynamics which in turn affects the blowdown. The kinetics calculation determines the energy added to the core which forms an essential input to the core thermal transient analysis. The SATAN-V code includes reactor kinetics calculation. b) Core Thermal Transient Based on the above information, a detailed analysis of reactor core cooling is performed. The code used in this part of the analysis is LOCTA-R2 for fuel pellet temperature, cladding temperature, and metal-water reaction evaluations.
C.l-6 9-15-71 e DESCRIPTION OF BLOWDOWN MODEL (SATAN-V) The blowdown hydraulic analysis for large size ruptures is required to provide the basic information concerning the dynamic behavior of the reactor core environment for the reactor kinetics and core cooling analysis. This requires the ability to predict the flow, quality, and pressure of the fluid passing through the core region. The SATAN-V code was developed with a capa-bility to provide this information. 4 The SATAN-V( ) (E_Ystem ~ccident and !_ransient Analysis of !!uclear !'.._lants) computer code performs a comprehensive space-time dependent analysis of a LOCA and is designed to treat all phases of the accident. The three major phases are: a) A subcooled phase where the rapidly changing pressure gradients in the subcooled fluid exert an influence upon the RCS internals and support structures. b) A two-phase depressurization phase. c) A refill-phase where the emergency core cooling system floods the core. The code employs a one-dimensional analysis in which the entire reactor coolant loop system is divided into control volumes. The fluid properties are considered uniform and thermodynamic equilibrium is assumed in each element. Pump characteristics, pump coast-down and cavitation, core and steam generator heat transfer including the W-3 DNB correlation in addition to the reactor kinetics are incorporated in the code. Reactor Kinetics The reactor kinetics considered in the code is a point kinetics model 9 identical to the CHIC-KIN( ) model. The reactor kinetics equations are the point kinetics equation with appropriately weighted feedback effects. The major reactivity feedback mechanisms incorporated in the kinetics equations
C.1-7 9-15-71 are the moderator density change, moderator temperature change, Doppler broadening as well as the control rod motion. These reactivity forcing functions are inputs in tabular form as functions of time. A block diagram of this continuous feedback model is shown in Figure C.1-2. Fundamental Equations The fundamental equations of conservation of mass, energy, and momentum are applied for single and two phase system in order to calculate the main de-pendent variables: density, internal energy and mass flow rate. A set of auxiliary dependent variables is defined by the system governing algebraic equations. Critical Flow A critical flow calculation for subcooled, two phase, or superheated break flow is incorporated in the analyses. The subcooled break flow is calculated using a modified version of the Zaloudek correlation. ( 5 ) The results are consistent with test results and do. not produce a discontinuity between the subcooled and two-phase break flows. The two-phase break flow is calculated by using the Moody(G) two-phase correlation. Comparison of these correlations with experiments indicate that the break area must be reduced by a discharge coefficient to match t~e experimental break flow and depressurization rates. However, in all the design calculations, a break discharge coefficient of 1.0 is used. Geometry The SATAN-V Code is equipped to analyze a maximum of two-loops plus "special" elements such as the reactor core, pressurizer and accumulators. For the three loop Surrey plant, both the broken loop and the unbroken loops were analyzed. - In the analysis the RCS is divided into a finite number of control volumes. Separate groups of control volumes are assigned to simulate the croken and unbroken loops and two control volumes are used to describe the coolant in
C.1-8 9-15-71 a-'"*.. the core region. This analysis replaces the FLASH-R analysis which had a limitation in that only three control volumes were availableo The current multi-control volume analysis, particularly with two control volumes in the core, permits execution of a detailed parametric survey of the important phenomena affecting the blowdown process. Method of Solution The RCS is described in the model via the following information: a) Description of the way in which the elements are interconnected. Each element can have as many as two flow paths out and up to five flow paths in, plus a break flow out. b) Initial thermodynamic data for each element: enthalpy, pressure, mass flow rates and heat transfer characteristics. c) ,Phrsical data for each element: dimensions and flow area, heat -/ transfer area, length, volumes and friction characteristics. Starting from a set of initial conditions the program evaluates the updated values of the dependent variables for all system elements using an integra-tion routine based on the Taylor series. The computer updated values are then substituted as the initial conditions for the next integration cycle and the procedure repeated until the final problem time is reached. The adequacy of the SATAN code to predict the hydraulic behavior during blowdown of the reactor coolant system during a loss-of-coolant accident has been verified by comparing the SATAN results of loss-of-coolant accidents of various sizes and locations with experiments performed at the Idaho and Battelle Northwest facilities(lZ). In selecting the control volumes for the SATAN code, truncations errors are minimized by selecting control volumes of equal length. In order to provide detail of the core region two core control volumes were selected. This enables the flow at the top, middle, and bottom of the core to be calculated
C.1-9 9-15-71 simultaneously. A total of 63 control volumes overall were selected for the large break analysis. Studies were performed to determine the effect of using more and less control volumes. The studies for less control volumes (15-20) showed that in addition to not having available the detail necessary to assess the flow regimes in the reactor vessel and core regions, significant changes in the answers resulted from small modeling changes. On the other hand, a study was performed using 10 rather .than 2 control volumes in the core re-sulted in only an*insignificant change in the results. The current control volume selection is therefore felt to be adequate. Later in this section, a discussion is presented which shows the methods used to establish a conservative use of all the main assumptions used in the analysis. Tqis study was performed to establish the conservative upper limit of the "confidence band" and forms the basii; for this analysis. DESCRIPTION OF THE CORE REFLOODING MODEL The SATAN calculations are performed until the completion of blowdown. In this context the end of blowdown is defined as the time at which zero break flow is first computed. The containment back pressure assumed for the blowdown analyses was 50 psia. At this time, the normal blowdown transient calculations are terminated and the reflooding calculations are performed. The reflooding model consists of three reference volumes which represent the downcomer region, the lower plenum region, and the active core region. The core and the downcomer volumes both communicate with the lower plenum volume via non-resistive flow paths. An input containment backpressure is assumed to act directly on the top of the downcomer volume, and any st_eam generated in the core region is vented to the containment via a flow path whose resistance simulates the flow path to the break. The model is shown in Figure C.1-3. Provisions for heat transfer from.vessel walls and reactor internals to injection water are also included in this model. At the end of blowdown, all of the water injected by the accumulators prior to this time is discarded. The refill calculations are then continued by adding the accumulator and safety injection flow to the water remaining in the downcomer region below the nozzle and in the lower plenum. No credit is
C.1-10 9-15-71 taken for water remaining in any other control volume. The conservative assumption of losing all of the water injected by the accumulators during blowdown delays the reflooding, extends the period of core recovery and results in a longer adiabatic core heat-up. -Accumulator flow and cold-legs safety injection flow are continuously calculated as they were during the blowdown transient and are added to the downcomer region. The injection system characteristics are as described in Section C.4 "Emergency Core Cooling Sys tern . " Until the water reaches the bottom of the core, the water levels in the down-comer and in the region below the core are equal. When the bottom of the core is reflooded by the accumulator water, steam is generated by the hot fuel rods, causing a pressure build-up in the core region. This retards the core reflooding process. The steam generated must be vented from the system through the break, and the flooding rate is limited by the resistance of the loop to the steam and water flow. There are two paths available for the steam and entrained water to flow to the break. The first path is directly to the break through the broken loop. The other path is through the intact loops, back into the inlet annulus, and.
., \ finally to the break through the inlet nozzle in the broken leg. These flow paths, as depicted, in Figures C.1-4 and C.1-5 show the path the steam mu~t follow for the cold-leg break. (Note that Figure C.1-4 and C.1-5 show one intact loop and one broken loop for the purposes of\illustrati~n.) The pressure drops along these two paths are calculated with the extsting fluid conditions and associated loss coefficients.
The presstke drop across the pump is maximized by assuming that the rotor is lo[Red. In addition, it is postulated that the accumulator water injected into the unbroken cold-leg pipes completely fills the cross-section, thus forming a plug that prevents venting of the steam generated in the core during the_accumu1ator injection phase of reflooding. These assumptions tend to reduce the core flooding rate and the fuel rod heat-transfer, thus resulting ih increased peak clad temperature. When the water level in the downcomer region reaches the bottom of the inlet nozzle, water from the injection flow is assumed to spill through the break into the containment sump.
C.1-11 9-15-71 The amount of mass vaporized and entrained as a function of core flooding rate 7 and time after reflooding is obtained from an analysis of the FLECHT( )(B) results. These results indicate that several flow regimes are present in the rod bundle during reflooding. For the first few seconds of the reflooding transient, until the core floods to approximately 20 inches, most of the heat transferred from the rod to the coolant goes to increase the liquid enthalpy. During this period almost no steam generation takes place and the core flood-ing rate equals the cold flooding rate. Following this initial period the steam velocity increases above the value required for entrainment and a dis-persed flow regime begins. This flow pattern is characterized by a continuous vapor phase with dispersed droplets and by a fast increase in rod heat transfer coefficient. It is during this phase of the reflooding transient that the flooding rate into the core is reduced by the resistance of the flow paths from the core to the break. The core flooding rate transient during this period is a function of the core and loop resistance, the fraction of coolant vaporized and en-trained, and the difference in water level between the downcomer and the core. The fraction of coolant vaporized, entrained and leaving the core is not constant during the transient, but increases from zero at the beginning to 70 to 80 percent of the entering coolant several seconds after initiation of reflooding depending on the core flooding rate. This is supported by FLECHT data. At the same time, due to the reduction in core flow, the water level in the downcomer region increases at a faster rate thus providing the water head required to discharge the increased exit core flow to the break. The following assumptions are made in the reflooding calculations. Each of these assumptions is conservative because they result in increased venting path pressure drop and therefore lower flooding rate.
- 1. The reactor coolant pump rotor is assumed locked.
- 2. The cold-leg pipe is plugged during accumulator injection.
C.1-12 9-15-71
- 3. The fraction of coolant vaporized and leaving the core is assumed to be equal to 0.8 for flooding rates up to 2 in/sec and increases to 0.85 for flooding rates higher than 4 in/sec.
- 4. No transient effects are considered in the transition between high to low flooding rates, but, even during this period of time, the core flooding rate is calculated by assuming that the fraction of core inlet flow that has to be vented to the break is equal to the equilibrium values specified in No. 3 above.
- 5. No water-steam separation is assumed to occur in the upper plenum and the quality of the mixture entering the loops in calculated from the core mass flow rate and nominal FLECHT heat transfer.
- 6. All the coolant is assumed to be superheated to 500°F in the steam generators. The superheating of the steam is conservatively assumed to take place instantaneously at the steam generator inlet.
- 7. The containment back pressure is equal to 40 psia.
- 8. For the steam velocities occurring in the cold-leg pipe and mass flow rate of the safety injection system, the acceleration pressure loss of accelerating all the water to the steam velocity is balanced by a small reduction in the steam temperature. Therefore, the injection of SIS water in the cold-leg does not increase the loop pressure losses during reflooding.
- 9. The effects of nitrogen discharging after the end of blowdown are not taken into account during the reflooding time period. Since the nitrogen discharges into the cold leg, the downcomer will be pressurized a very short time period before the hot leg. This will result in an additional amount of downcomer water entering the core that is not taken credit for in the initial reflood analysis.
C.1-13 9-15-71 DESCRIPTION OF THE THERMAL MODEL (LOCTA-R2) The LOCTA-R2 transient digital computer program was developed for evaluation of fuel pellet and cladding temperature during a LOCA. It also determines the extent of the Zircaloy-steam reaction and magnitude of the resulting energy release in Zircaloy clad cores. The transient heat conduction equation is solved by means of finite differences considering only heat flow in the radial direction. A lumped parameter method is used; the fuel containing three radial nodes and the cladding one node with a specified resistance simulating the pellet to clad gap. Internal heat generation can be specified as a function of time. The decay heat is based on the heat generated from: a) fission products; b) capture products; and c) residual fissions. It is assumed that the core has been irradiated for an infinite period of time. The method used to calculate the total decay heat is presented in Section C.3. In addition to decay heat the code calculates the heat generated due to the Zircaloy-steam reaction. The Zr-H o reaction is governed by the parabolic 2 rate equation(l))(l 4 ) unless there is an insufficient supply of steam. In this case, the Zircaloy oxidation is limited by the reduced steam avail-ability. In the design calculations the parabolic rate equation is assumed for the full transient even when the fuel rod is assumed adiabatic. The build-up of the Zircaloy-oxide film is calculated as a function of time, and its effect on heat transfer is considered. An isothermal clad melt is con-sidered based on the heat of fusion of Zircaloy. The code has been developed to stack axial sections and thereby describe the behavior of a full length region as a function of time. A mass and energy
C.1-14 9-15-71 balance is used to evaluate a temperature rise in the steam as it flows through the core. Each radial region is considered independently and a number of axial sections which may be analyzed is essentially unlimited. The initial conditions for the fuel rod are specified as a function of power. The following core conditions are also introduced as a function of time: a) mass flow rate through the core; b) coolant quality; c) pressure. To assure the conservatism of the hot channel temperature calculations the following procedure is used for the hot spot temperature calculation:
- 1. The higher of the two SATAN core control volume quantities is used in.hot spot temperature calculations.
e 2. The flow used in the hot channel temperature calculations is the core midplane SATAN flow reduced by 20%. Blowdown and reflooding heat transfer processes in the core are evaluated by means of using correlations summarized in Section C.2 which have been validated with several Westinghouse Research and Development Programs. From the end of blowdown until core reflood the core is assumed adiabatic even though steam cooling is expected because of the local depressurization caused by the accumulators and by circulation promoted by the core and other heat sources. In summary, .information generated by LOCTA-R2 as a function of time includes: a) fuel temperature; b) clad temperature; c) steam temperature; d) amount of metal-water reaction, if any; e) volume of core melt, if any; f) total heat released to coolant. A more detailed explanation of the models and correlations used in LOCTA-R2 is presented in WCAP-7437-L(lO).
C.1-15 9-15-71 RESULTS The detailed description of the Reactor Coolant System that can be obtained in the multi-control volumes SATAN-V code has been used to analyze the following phenomena affecting the blowdown process, such as:
- 1. Heat transfer from core to the coolant during blowdown
- 2. Reactor Coolant Pump Characteristics
- 3. Steam Generator Heat Transfer Characteristics
- 4. Loop Resistances and Break Location
- 5. Accumulator Performance A parametric survey study was performed for the Indian Point Unit No. 2 Final Safety Analysis Report, (ll) Supplements 12 and 13 with the purpose of determining the most conservative combination of the above assumptions as input to the SATAN-V code. Additional parametric studies performed for Turkey Point Unit No. 3 and Indian Point Unit No. 2 are presented in the 6/1/71 AEC submittal "Emergency Core Cooling Performance 11 ( 3 ). These studies demonstrated that conservative clad temperature calculations result from the core thermal analysis when the heat transfer from the core to the coolant during blowdown as calculated by SATAN is equal to or greater than that calculated by the LOCTA-R2 code. The present analysis satisfies the above requirement. The phenomena listed in Item 3 above were not significant.
The significant phenomena, with respect to input assumptions for the SATAN-V code, during blowdown are the reactor coolant pump assumptions (Item 2) and the break characteristics (Item 4). These results indicate that the most conservative pump assumption is to assume that the pumps trip at the time of the break and continue to coast down until cavitation conditions were reached. At this time and for the remainder of blowdown the pump was assumed to not lock but continued to develop a conservative head. The pump speed is con-tinually calculated as a function of prevailing conditions and the pump characteristics. An additional study investigated the sensitivity of the results to the mode of cold leg pipe failure. The double ended break was analyzed for two forms
C.1-16 9-15-71 of break characteristic: guillotine and split. This break size yielded worst answers for the guillotine failure; the less severe break mode was more than 200°F lower in peak clad temperature. The analysis of the LOCA was performed at 102% of the core maximum calculated power of 2546 MWt and at a peak linear power of 16.4 kw/ft. This value of peak linear power includes a 5% allowance for nuclear uncertainties. The heat flux distribution in the hot channel, consistent with this power, was a 1.61 peak to average cosine. Table C.1-1 presents the peak clad temperatures and the hot spot clad metal reaction for a range of break sizes. Figures C.1-6 through C.1-15 present the transients for significant parameters for the break sizes analyzed. The following items should be noted: Figures C.1-6 The system pressure curve shown is the calculated pressure to C.1-10 in the lower core region. The quality curve presents the values for the hot spot; this is the larger of the two values calculated in the core region. The core flow shown is the actual value for the core midplane. The average mass velocity 2 can be obtained by dividing the flow rate shown in 41.7 ft . The hot spot mass velocity is taken to be 80 percent of the average mass velocity. The heat transfer coefficient shown is for the hot spot. Figures C.1-11 These figures show the hot spot clad temperature transient. to C.1-15 The remainder of the figures show additional details for the worst break (Double Ended Guillotine break in the cold_leg). Figure C.1-16 The core pressure drop shown is from the lower plenum, near the core, to the upper plenum at the core outlet. Figure C.1-17 This is the hot spot fluid temperature.
C.1-17 9-15-71 Figure C.1-18 The flow rate out of the break plotted is the sum of both ends for this case. The lower plenum flow is at the bottom of the core; the upper plenum flow is at the top of the core. Figure C.1-19 The hot leg flows are shown for a position near the end of the nozzles. The cold leg flows are at the vessel-nozzle con-nection. The unbroken loop flows shown are the sum for the two intact loops. Figure C.1-20 This figure shows the core reflood transient. Figure C.1-21 The accumulator flow shown is the sum for the two intact loops. 2 The peak clad temperature occurred for the Double Ended (8.24 ft ) guillotine cold-leg break and was 2300°F which meets the AEC interim criterion. The hot spot metal water reaction is 8.6 which is well below the embrittlement limit of 16% described in Figure C.1-1. In addition, the total core metal water reaction is less than 0.1% for the highest temperature break. This compares with the AEC interim criterion of 1%-. The hot-leg break case has not been presented here because experience has shown that clad temperatures are low for the following reasons:
- 1) There is no flow reversal during hot leg break blowdown; the flow just decays smoothly, providing good heat transfer in the core.
- 2) In a cold-leg analysis, the accumulator and high head injection flow for the broken loop is assumed to spill to the containment. This assumption is obviously not applicable for a hot-leg break; thus there is more accumulator water mass and more injection flow available for cooling.*
- 3) During the reflood phase of the accident there is no problem with steam binding since the steam generated in the core is vented directly to the containment via the broken hot-leg; heat transfer during re-flood for the hot-leg break will be better than for the cold-leg break. The peak clad temperature will be less than 2000°F.
C,l-17a 2-1-72 e Evaluation of the containment backpressure of 40 PSIA used in the reflood analysis shows that the value used is conservative with respect to peak clad temperature. To illustrate this, two additional calculations were performed for the worst-case break (double-ended guillotone cold leg rupture) as shown on Figure C.1-30. Case A is the reflood pressure transient used in the original calculation - the containment backpressure of 40 PSIA was held constant during the reflood analysis. Case A gave a peak clad temperature of 2300 °F. An additional study, Case B, was made using ttie complete pressure transient as calculated by the Stone and Webster "LOCTIC" code for the Interim Criteria analysis, Since there is a higher backpressure during reflood for this case, peak clad temperatures were below the 2300 °F criterion. An additional analysis, Case C, was performed assuming a containment backpressure constant at 40 PSIA for 110 seconds and then following the "LOCTIC" pressure decay transient. Case C resulted in a peak clad temperature of 2300 °F.
C.1-18 9-15~71 Description of Small Break Analysis (SLAP Code) The SLAP code is a digital computer code employed to calculate the transient depressurization of the RCS as well as to describe the mass and enthalpy of flow through a break. The code considers three regions: a) the Reactor Coolant System; b) the pressurizer; c) the shell side of the steam generators. The three regions are represented as shown in Figure C.1-22. The initial system conditions are designated by: a) volume of each region; b) RCS pressure and average enthalpy; c) pressurizer water level; d) steam generator shell side pressure and water level. The code uses the equations of state, continuity, and energy conservation to define the condition in each volume as a function of time. The model assumes complete separation of water and steam in each volume. Fluid can flow between the pressurizer and the RCS; the flow is defined by the momentum equation. Under the initial steady state operating conditions~ fluid flow between the pressurizer and the RCS is zero. Heat can be transferred between the RCS and the secondary system; the initial heat transfer rate in the steam generators is equal to the heat generated in the reactor coolant system. e
C.1-19 9-15-71 The break is considered in a cold leg between the reactor coolant pump outlet and inlet nozzle. This depicts a worst break location. The initial sub-cooled break flow is treated by the correlation of Fauske(lS), who concluded that sharp-edged orifice test data can be accurately correlated using the imcompressible flow equations for a nozzle. Subsequent saturated water flow through the break is treated by the correlation of Moody( 6 ). Two-phase fluid blowdown is allowed to continue until the quiet water volume, resulting from complete separation of steam and water, drops below the break volume (Figure C.1-22). A water filled loop seal was considered in the analyses of the small break transient. That is, the water volume in the RCS takes into account the water volume necessary to maintain a full downcomer and loop seal until the loop seal is uncovered. Following loop seal uncovery the downcomer water volume above the covered portion of core flows into the plenum and equalizes the water level across the reactor vessel. After the water level in core reaches the loop seal e level, steam is assumed to be discharged from the break. Reactor trip is caused by the pressurizer pressure reaching the low pressure set point. Following the trip, decay heat, heat from hot internals, and the vessel, is transferred to the RCS. This continues unless th.e core is uncovered. If the core is uncovered, the core stored and decay heat added to the RCS fluid is directly proportional to the amount of the core that is covered by the two phase mixture of steam and water. For this purpose the two phase level is the calculated quiet water level plus a fraction to account for froth. The froth level is ca1 cu 1 ate d using t h e W"1 l son corre l ation
. (l6 ) . Heat trans f er toteh secondary system is assumed to occur only in the tube region below the shell side water level. The code accounts for heat transfer in either direction across the tubes. Secondary steam flow is specified as a function of valve flow area to simulate the operation of the safety valves or relief valves. Auxiliary feedwater flow is provided as a function of time after the trip. Auxiliary feedwater pump startup delay times and line purging times are taken into account.
c.1 ....20 9....15 ... 71 Safety injection flow to the RCS is input in a tabular form, i.e., injection flow rate as a function of system pressure. Safety injection is actuated when the specified pressurizer low pressure and low level set points are reached. A safety injection pump start-up delay time is included in the model. The water in the accumulators is automatically discharged when the RCS pressure drops below the accumulator pressure. The reduction in reactor coolant flow due to reactor coolant pump coastdowri and cavitation following pump trip is simulated by the reduction of the steam generator heat transfer coefficient with time after reactor trip. Peak clad temperature analyses use RCS pressure, fluid flow ,rate through core, void fraction, and fluid volume in the reactor vessel, These parameters are calculated with the SLAP code. Results~ Small.Breaks e Analyses results discussed in the previous section indicated the adequacy of the accumulators to reflood the core and limit the temperature rise of the core for large area ruptures. In this section, the analyses deals with the break of up to a 6 inch equivalent diameter pipe, where the safety injection pumps play an increasing role in the initial reflooding because of the slower depressurization of the Reactor Coolant System. Analyses results presented in this section will show that the high head portion of the ECCS, together with the accumulators; provide sufficient core flooding so that the design criteria are met. During the earlier part of the small break transient, the effect of the break flow is not strong eno_ugh to overcome the flow maintained by the Reactor Coolant pumps through the core as they are coasting down following the assumed trip. Therefore, upward flow through core is maintained. The temperature transients were calculated for the range of break sizes wi.t~ the arbitrary assumption that DNB occurs at 0.5 seconds after the ac.~ideti:t
- .~*f*"
initiation. The heat transfer coefficient after DNB was
C.1-21 9-15 .. 71 7 calculated using the transition boiling correlation described in WCAP-9005(! ). For the initial part of the transient the peak clad temperatures are less than 1000°F. During the recovery period of the accident the thermal analysis was performed with a conservative axial flux distribution skewed to the top of the core. The full range of break sizes was analyzed by considering the minimum safeguards ECCS capability and operability. The safety injection system is described in Section C.4. The case analyzed considered a single failure of one diesel-generator unit of the emergency power supply system. This results in one SIP operating. One cold leg line is considered to be injecting near the break and the flow from this line is assumed to spill. These analyses were performed with S.I. system performance depicted in Figure C.1-23. These curves are derated for analyses purposes by reducing pump discharge pressure by 5 percent. Peak clad temperatures were calculated to be less than 1500°F for all break sizes for the case analyzed. These temperatures are acceptable and therefore indicate that adequate protection is provided in the event of a small break LOCA. CONCLUSIONS For breaks up to and including th.e double-ended severance of a reactor coolant pipe, the ECCS with partial effectiveness will limit the clad temperature to 2300°F and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved. There were no deviations from Appendix A, part 2 of the Interim Policy Statement. The ECCS design therefore meets the AEC Interim Core Cooling Criteria
- C.1-22 9-15-71 e TABLE C.1-1 Peak Clad Temperature and Hot Spot Metal-Water Reaction Break Peak TemEerature Percent Zirc Reaction Double Ended (Guill) 2300°F 8.6 0.8 Double Ended (Guill) 2290°F 8.5 2
4.5 Ft (Split) 2240°F 7.4 2 3.0 Ft (Split) 2014°F 3.2 2 0.5 Ft (Split) 1860°F 1.5
c.1-23 9-15-71 REFERENCES (1) "Topical Report - Performance of Zircaloy Clad Fuel Rods During a Simulated Loss-of-Coolant Accident - Single Rod Tests, "WCAP-7379-L, Volume I (Proprietary) and Volume II (Non-proprietary), September, 1969. (2) "Topical Report - Performance of Zircaloy Clad Fuel Rods During a Simulated Loss-of-Coolant Accident - Multi-Rod Tests, "WCAP-7495-L, Proprietary, Volume I - Test Setup and Results; Volume II - Analyses of Results, July, 1969. (3) "Emergency Core Cooling Performance" (W Proprietary) 6/1/71 AEC submitted. (4) F. M, Bordelon, "A Comprehensive Space-Time Analysis of Loss-of-Coolant (SATAN Digital Code), WCAP-7263 (W Proprietary), March, 1969. (5) F, R. Zaloudek "Steam Water Critical Flow from High Pressure System," Hanford Laboratories, HW-80535, January, 1964. (6) F. H. Moody, "Maximum Flow Rate of Single Component, Two-Phase Mixture", Paper No. 64-HT-35, and ASME Publication. (7) J. O. Cermak, et. al., "Full Length Emergency Cooling Heat Transfer (FLECHT) Group I Tests," WCAP-7435, January, 1970. (8) F. F. Cadek, et. al., "Full Length Emergency Cooling Heat Transfer (FLECHT) Group II Test Report," WCAP-7544, September, 1970. (9) J. A. Redfield, "CHIC-KIN. *
- A Fortran Program for Intermediate and Fast Transients in a Water Moderated Reactor, WAPD-TM-479 (January, 1965).
(10) W. A. Bezella, et. al., "LOCTA-R2 Loss of Coolant Transient. Analysis," WCAP-7437-L, January, 1970. (11) Supplement 12 to the !PP No. 2 FSAR, (Docket No. 's 50-247 and 50-286). (12) "Topical Report - Westinghouse PWR Core Behavior Following a Loss of Coolant Accident", WCAP-7422-L (W Proprietary), January, 1970. (13) James C. Hesson, et. al., "Laboratory Simulations of Cladding - Steam Reactions Following Loss of Coolant Accidents in Water-Cooled Power Reactors" ANL-7609. (14) L. Baker, Jr. and Just, J.C., "Studies of Metal Water Reactions at High Temperature," ANL-6548, 1962.
FIGURE C, i'-1
!l-1 !'i-71 50
- F-11
- F-10
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UJ u 35 0F LEGEND: I I" LONG PREVIOUSLY BURST TUB I NG c::. LU OQ 4" LONG NEW TUBING __.
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DIRECT HEAT FRACTION DETAILED HEAT MOMENTUM KINETICS HEAT CONDUCT ION SECTIONALIZED CALCUUTI ONS WITH DELAYED SOLUTION IN WATER CHANNEL ON REFLECTOR, NEUTRONS FUEL ELEMENT CNAHEL AND Uf'PEI f'LEXUM &K j1 TOTAL FLUID MOTION liK (DOPPLER EFFECT)
~ liK (FUEL ELEMENT EXPANSION)
~ 6K (TEMPERATURE CO EFF IC l;f NT)
_ 6K ( DENS ITV COEFFICIENT) liK (FORCING FUNCTIONS) I A Block Diagram of the Continuous Feedback Model SA.TAN-V Code ' *"' UI
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'I 50 40 3'
e"' g i 20 CONSTANT REFLOOD PRESSURE= 40 PSIA PEAK CLAD TEMPERATURE= 2300°F 0 CASE B - S&W CALCULATED PRESSURE TRANSIENT CURVE PEAK CLAD TEMPERATURE< 2300°F 10 CASE C - REFLOOD PRESSURE = 40 PSIA TO 110 SECOND, S&W CALCULATED PRESSURE AFTER 110 SECONDS, PEAK CLAD TEMPERATURE = 2300 .., H G') c::: 0 "'....I 1::l 1 10 100 1000 ...., :' I
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TIME AFTER BREAK ( SECONDS)
C.2-1 9-.15-71
- C.2 HEAT TRANSFER COEFFICIENTS USED IN THE LOCTA-R2 CORE THERMAL ANALYSIS The heat tranfer correlations used in the LOCA analyses are presented below in order of application during the accident.
Time of Break Until Occurrence of DNB The time of the break until occurrence of DNB is taken conservatively to be 0.1 sec. The heat transfer regime during this period of the.accident is forced convection turbulent heat transfer or fully developed nucleate boiling. The correlation for nuclear boiling is that by Jens and Lottes. (l) In the nucleate boiling regime the wall temperature is a function of the heat flux and pressure, not coolant velocity. The Jens and Lottes correlation e is independent of geometry, i.e., valid for tubes, plates or rod bundles. It is also used for both local and bulk boiling. The correlation has been compared to subcooled water data obtained from single heated t~bes having internal diameters from 0.143" to 0.288", length from 3" to 24.6" J and pressure ranging from 500 to 2000 psia. The Dittus-Boelter correlation is used for forced convection turbulent heat transfer. From DNB Until Time of Uncovering (Steam Cooling Period) In the large break design calculations DNB is assumed to occur at 0.1 sec. After DNB occurs, the mode of heat transfer is unstable with both nucleate boiling and film boiling existing from times of short duration. The heat transfer correlation used in the transition to film boiling period 2 is described in WCAP-9005( ). The verification of the use of this correlation for post DNB heat transfer during a transient blowdown is also presented in WCAP-9005.
C.2-2 9-15-71 The initial operating conditions prior to blowdown are:
- 1. Pressure 2250 psia 2
- 2. Mass Velocity 1,000,000 to 2,500,000 lb/hr ft
- 3. Inlet Temperature 480 to 560°F 2
- 4. Heat Flux 635,000 to 1,100,000 Btu/hr ft Blowdown conditions were:
- 1. Initial blowdown rate 200 to 10,000 psi/sec 2
- 2. Average flow decay rate 250,000 to 1,150,000 lb/hr ft sec The test section consisted of a 1/2 inch inside diameter circular tube 3 feet long. The 3 foot length is sufficient to establish fully developed flow at the exit of the test section (L/D = 72).
A total of 50 transient blowdown runs were performed. To determine the effect of flow decay 20 runs were performed during which the flow rate was maintained as close as possible to the initial value.* Thirty runs were performed in which the flow to the test-section was allowed to decay in addition to the depressurization~ This latter condition more nearly resembles the predicted conditions in a PWR core during the large break LOCA. For this reason the majority of the runs were performed with flow decay. The comparison of predicted heat transfer coefficient with the measured data is shown in Figure C.2-1 for all data points. It is readily apparent that the correlation is conservative with respect to the results of this test since the measured value is greater than predicted for 95% of the data. The degree of conservatism contained in the correlation increases with increasing values of the heat transfer coefficient.
C.2-3 9-15-71 - During Uncovering (Steam Cooling Period) This period of the loss-of-coolant accident considers either turbulent or laminar forced convection to steam combined with radiation from the fuel rod surface to the steam. Radiation between fuel rods is not considered. a) For turbulent forced convection to steam a Dittus-Boelter type equation, modified by McEligot( 3 )( 4 ) to account for the variation in fluid properties near the wall due to a large temperature gradient, is used. The ( TTwb) term i s id n ependent o f geometry. The Di ttus- Boe1 ter type equat i on was developed from flow inside tubes with values of C = 0 .023.
- Weisman( 5 )
has shown that C is higher for rod bundle data (~0.023). A lower value of C as shown in the above correlation is presently used in the loss-of-coolant analyses. The McEligot correlation was compared to data with hydraulic equivalent diameter values of 0.125" and 0.25" and L/D greater than 150. Additional coolant conditions are described below. Coolant Reynolds Number Maximum Wall Temperature, OF Tw/Tb Air 1450-15000 1520 2.17 Helium 7570-13400 1050 1.56 Nitrogen 18200-45000 1620 4.78 The Prandtl number of steam is similar to those obtained with the above coolants. b) For laminar forced convection to steam, the heat transfer correlation used is based on theoretical calculations of laminar flow in tubes made by Hausen CG) and Kays( 7).
i C.2-4 9-15-71 h/h = (Tw)-0.25 iso Tb These calculations indicate that the local Nusselt nwnber is highest near the inlet and drops until it reaches a limit corresponding to fully-developed thermal conditions. For the case of constant wall temperature, the limiting Nusselt number is 3.66. For. constant heat flux at the wall, the asymptote is 4.36. Furthermore, these calculations indicate that- the asymptotic values are reached for all practical purposes when 1/D/RePr > 0.05. For the Reynolds numbers ranging from 100 to 1000 and a Pr= 1.0 for steam, the developing length is from 2.5" to 25". The correlation was compared to data from laminar air flow in circular tubes where the 1/D ranged from 42 to 80 and Re< 3000. T 0 25 The (Tw)-
- term is to account for variations in fluid properties b
near the wall due to large temperature gradients. c) Radiation to steam is evaluated employing the emperical method of Hottel(S). T
. Tw 4 H20 4 h = 0.1713 XE [(
100 ) ( lOO) ]
---------T - T w b where TH 0 E
1 H 0 (-2-)0.45 2 T w
C.2-5 9-15-71 - The present value of the correction factor, pressure than O or 1 atmosphere is 2.0. C, for Verification of Correlations Used During Steam Cooling Period E
. ~o at higher The use of the above correlations during the steam cooling period was verified by the work performed at the University of Michigan under Westinghouse funding and direction. This was part of the Flashing Heat Transfer Program.
The results of this phase of the program have been documented in WCAP-7396-1( 9 ). The primary objective of this test was to determine the behavior of radiation heat transfer to steam at elevated pressures (up to 5 atm.). The heat transfer test facility consisted of an open heat transfer loop. Steam was delivered to the test section and discharged to the atmosphere through the necessary piping and control apparatus. The test sections consisted of 1/2 inch ID pipes, with an active length e of three feet. The walls of the test section were heated by electrical resistance heating. Test sections having both uniform and non-uniform axial heat generation were employed. The range of variables were representative of that in a PWR when the core is uncovered and is as follows: 3 4 2 Mass Velocity G =4 X 10 to 4 X 10 lb/hr-ft Temperature T. = 300 to 100°F 1n Wall Temperature 400 to 1800°F Pressure P. = 25 to 75 psia 1n Inlet Reynolds Number 1900 to 35000 The results of the low pressure heat transfer test yield the following conclusions:
- 1. The McEligot et. al. correlation realistically predicts the convective heat transfer coefficients in turbulent flow.
J
C.2-6 9-15-71 - 2. In turbulent flow the radiant heat transfer contribution to the total heat transfer coefficient is adequately predicted by Hottel's technique.
- 3. The total heat transfer coefficient to steam in turbulent flow may be calculated by adding the convective term determined by McEligot's correlation and radiative term determined by Hottel's technique.
A comparison of predicted versus measured total turbulent heat transfer coefficient is shown in Figure C.2-2 and excellent agreement can be seen.
- 4. In laminar flow the total heat transfer coefficient is conservatively predicted by using*the correlation of Hausen and Hays for the convective contribution and the method of Hottel for the radiant contribution.
- 5. The prediction of laminar heat transfer coefficient can be improved by evaluating the steam properties at film conditions instead of e bulk conditions.
- 6. The effect of a non-uniform heat flux on the heat transfer coefficient is negligible for the conditions which exist during a LOCA.
Recovery Phase of the Accident After entrainment has been initiated, heat transfer coefficients obtained from the FLECHT Program(!O) are used. Detailed results of the Group I and Group II test series can be found in Reference 11 and 12, respectively. In summary, the current test results verified the ability of a bottom flooding SIS design to terminate the temperature increase during a LOCA. In particular, it has been shown that the effects of a variable flooding rate can be predicted, using constant flooding rate data and that complete blockage of as many as sixteen adjacent channels will not impair bottom flooding core cooling effectiveness.
C.2-7 9-15-71 NOMENCLATURE e SYMBOLS D - equivalent diameter e G - mass velocity L - length of heat source N - Nusselt number u P - system pressure Pr - Prandtl number Re - Reynolds number T - temperature g - gravity constant h - heat transfer coefficient k - thermal conductivity q" - heat flux
£ - effective emissivity µ - dynamic viscosity e p - density SUBSCRIPTS b - quantities evaluated at bulk fluid temperature iso - evaluation of the parameter when the temperature difference (Tw - Tb) is small sat - refers to saturated condition v - saturated vapor L - refers to liquid w - wall
C.2-8 9-15-71 REFERENCES APPENDIX A
- 1. W. H. Jens and P.A. Lattes, "Anaylses of Heat Transfer, Burnout, Pressure Drop, and Density Data for High Pressure Water," USAEC Report ANL-4627, 1951. Pleted. Plant S
- 2. R. F. Farman and J. O. Cermak, "Post DNB Heat Transfer During Blowdown,"
WCAP-9005, October 1968, Proprietary.
- 3. D. M. McEligot, P. M. Magee, and G. Leppert, "Effect of Large Tempera-ture Gradients on Convective Heat Transfer: The Downstream Region,"
J. of Heat Transfer, Vol. 87, 1965, pp 67-76.
- 4. D. M. McEligot, L. W. Ormand, and H. C. Perkins, "Internal Low Reynolds-Number Turbulent and Transitional Gas Flow with Heat Transfer",
J. of Heat Transfer, Vol. 88, 1966, pp. 239-245.
- 5. J. Weisman, "Heat Transfer to Water Flowing Parallel to Tube Bundles,"
Nuclear Science and Engineering 6, 79, 1969.
- 6. H. Hausen, "Darstellung des Warmeuberganges in Rohren durch verall gemeinerte Potenzbezienhungen," VDI Zeit., No. 4, p. 91, 1943.
- 7. W. M. Kays, "Numerical Solutions for Laminar-Flow Heat Transfer in Circular Tubes, 11 Trans. ASME, Vol. 77, 1955, pp. 1265-237 4.
- 8. H. C. Hottel, "Radiation Heat Transmission," Ch. 4 of Heat Transmission by W. H. McAdams, McGraw-Hill, 1954.
- 9. R. M. Hunt (Editor) "Safety Related Research and Development for Westinghouse Pressurized Water Reactors - A Program Outline Fall 1969,"
WCAP-7396-L, November 1969, Proprietary.
- 10. R. M. Hunt (Editor), "Safety Related Research and Development for Westinghouse Pressurized Water Reactors - A Program Outline Spring 1969," WCAP-7304-L, April 1969, Proprietary.
- 11. J. O. Cermak, et al., "Full Length Emergency Cooling Heat Transfer (FLECHT) Group I Tests," WCAP-7435, January 1970.
- 12. F. F. Cadek, et al., "Full Length Emergency Cooling Heat Transfer (FLECHT) Group II Test Report, 11 WCAP-7544, September, 1970.
FIGURE C.2-1 9-15-71 13 C 0 0 t:, 0 12 0 *o-0 II 0 0 0 10 9
~
0 Oo C 0
/
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/
( LL 0 0 0 N 0 I-LL
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- 0 2 3 5 6 7 8 9 10 Heat Transfer Coefficient Comparison
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I-C) 0 l&.._ _ _........_ _ _ _.,___ _ _-.L_ _ _ _....___ _ _ _ __ _ , 0- 20 ltO 60 10 100 TOTAL NEAT TRANSFER COEFFICIENT PREDICTED (h,) (ITU/HR - FT2 Of) University of Michigan Heat Transfer Test Turbulent Flow Data Total Heat Transfer Coefficient Low Pressure Heat Transfer Test
C.3-1 9-15-71 RESIDUAL DECAY HEAT e C.3 In the analysis of loss of coolant transients, the residual decay heat during the period immediately after shutdown is important in determining peak clad temperatures. The decay heat data used as input to these analyses is described in this section. Residual heat in a subcritical core consists of (a) fission product decay energy, (b) decay of neutron capture products, and (c) residual fissions due to the effect of delayed neutrons. These constituents are discussed separately in the following paragraphs. Fission Product Decax_ 3 For short times (<10 seconds) after shutdown, data on yields of short half-life isotopes in sparse. Very little experimental data is available for they-ray contributions and even less for the S-ray contribution. Several authors have compiled the available data into a conservative estimate of fission product decay energy for short times after shutdown, notably Shure(!), Dudziak( 2 ), and Teage( 3 ). Of these three selections, Shure's curve is the highest, and it is based on the data of Stehn and Clancy( 4 ) and Obenshain and Foderaro(S). The fission product contribution to decay heat which has been assumed for the loss of coolant analyses in this application is the curve of Shure increased by 20% for conservatism. This curve with the 20% factor included is shown in Figure C.3-1. The curve of Shure coincides with the recommendation in the proposed ANS standard on fission produ~t decay heat (unpublished). The proposed standard also recommends that 20% be added for conservative design.
C.3-2 9-15-71 Decay of U-238 Capture Products Betas and gammas from the decay of U-239 (23.5 min. half-life) and Np-239 (2.35 day half-life) contribute significantly to the heat generation after shutdown. The cross-section for production of these isotopes and their decay schemes are relatively well known. For long irradiation times their contribution can be written as:
= yl + ESl 200 Mev Ey2 + ES2
= 200 Mev c(l+a) watts/watt where: P /P is the energy from U-239 decay 1 0 P /P is the energy from Np-239 decay 2 0 tis the time after shutdown (secs) c(l+a) is the ratio of U-238 captures to total fissions = .6(1+.2)
-4 secs -1 "1 = the decay constant of U-239 = 4.91 x 10 e "2 = the decay constant of Np-239 = 3.41 x 10
-6 secs -1 Eyl = total y-ray energy from U-239 decay= .074 MeV Ey2 = total y-ray energy from Np-239 decay= .30 MeV ESl = total S-ray energy from U-239 decay= 1/3 x 1.18 MeV ES2 = total 8-ray energy from Np-239 decay= 1/3 x .43 MeV (Two-thirds of the potential 8-ray energy is assumed to escape by the accompanying neutrinos).
This expression with a margin of 10% is shown in Figure C.3-1. The 10% margin, compared to 20% for fission product decay, is justified by the availability of the basic data required for this analysis. The constants given above are in agreement with the coefficients of the proposed ANS standard. The decay of other isotopes, produced by_neutron reactions other than fission, is neglected.
C.3-3 9-15-71
- Residual Fissions The time dependence of residual fission power after shutdown depends on core properties throughout the transient. Spatially dependent kinetics calculations have not been performed for these transients. It is assl.Dlled that core average conditions are more conservative for the calculation of reactivity and power level than actual local conditions as they would exist in hot areas of the core. Thus, static power shapes are asswned (with the exception of the effect discussed in the next section) and these are factored by the time behavior of core average fission power calculated by a point model kinetics calculation with six delayed neutron groups.
For the purpose of illustration, only a one delayed neutron group calculation, with a constant reactivity of -4%~p, is shown in Figure C.3-1. Distribution of Decay Heat e During a loss of coolant accident the core is rapidly shut down by void formation or control rods, or both, and a large fraction of the heat generation to be considered comes from fission product decay gamma rays. This heat is not distributed in the same manner as steady state fission power. Local peaking effects which are important for the neutron dependent part of the heat generation do not apply to the gamma ray contribution. The steady state factor of 97.4% which represents the fraction of heat generated within the fuel, drops_ to 95% for the hot rod in a loss *of coolant accident. For example, consider the transient resulting from the double ended break of the largest primary circuit pipe; 1/2 second after the rupture about 30% of the heat generated in the fuel rods *is from gamma-ray absorption. The gamma power shape is less peaked than the steady state fission power shape, reducing the energy deposited in the hot rod at the expense of adjacent colder rods. A conservative estimate of this effect is a reduction of 10% of the gamma-ray contribution of 3% of the total. Since the water
C.3-4 9-15-71 density is considerably reduced at this time, an average of 98% of the e available heat is deposited in the fuel rods, the remaining 2% being absorbed by water, thimbles, sleeves and grids. The net effect is a factor of .95 rather than .974, to be applied to the heat production in the hot rod. e
C.3-5 9-15-71 REFERENCES
- 1. K. Shure, "Fission Product Decay Energy," Bettis Technical Review, WAPD-BT-24, December, 1961.
- 2. K. Shure and D. J. Dudziak, "Calculating Energy Release by Fission Products," Trans. Am. Nucl. Soc. 4 (1) 30, 1961.
- 3. U.K.A.E.A. Decay Heat Standard (Reference unavailable).
- 4. J. R. Stehn and E. F. Clancy, "Fission-Product Radioactivity and Heat Generation," "Proceedings of the Second United Nations Inter.:..
national Conference on the Peaceful Uses of Atomic Energy, Geneva, 1958," Vol. 13, pp. 49-54, United Nations, Geneva, 1958.
- 5. F. E. Obenshain and A.H. Foderaro, "Energy from Fission Product Decay" WAPD-P-652, 1955.
10° 8 .. 6 IJ 2
\ '-..
TOTAL RESIDUAL HEAT (WI TH ij% SHUTDOWN) 10- 1 ac: U.I 8 ]I:: 0 Q. 6 I- PRODUCT DECAY ,c IJ I-I- ,c ]I: ( I) 2 RESIDUAL FISSIONS i:: C FOR~% SHUTDOMI ]I:: 10- 2 8 6 U-238 CAPTURE DECAY IJ 2
,o-3 10- 1 2 2 1J 6 8 10 I 2 1J 6 8 10 2 2 IJ 6 8103 TIME AFTER SHUTDOWN (SECONDS) -.,
Ci) c:: co ;ID I rr, Residual Decay Heat - 11'1
- 0 I CA>
'-..! I
C.4-1 9-15-71 e C.4 EMERGENCY CORE COOLING SYSTEM The following is a description of the revised low head and high head Safety Injection Systems. These systems have been revised to meet the following requirements.
- 1. Provide injection of sufficient borated water into the cold legs during the first hour following an LOCA.
- 2. Retain the capability for long-term recirculation flow to the hot legs from both the low head and high head Safety Injection System (SIS).
To meet these requirements the following paths for the low head and high head safety injection are provided. Low Head Safety Injection Pumps
- 1. Injection flow to the three cold legs.
- 2. Recirculation flow to three hot legs or three cold legs.
Centrifugal Charging Pumps
- 1. Injection flow to three cold legs.
- 2. Recirculation flow to three hot legs or three cold legs.
A flow diagram illustrating the revised Safety Injection System is shown in Figure C. 4-1. The low head SI pumps are started on receipt of the safeguards initiation signal and provide injection flow to the three cold legs. During normal plant operation, the three motor operated valves (MOVs) 864A, 864B and 890C would be in a normally ~. open position to ensure the availability of low head cold leg injection from both pumps in the event of a loss-of-coolant accident.
C.4-2 9-15-71 The low head SI pumps also provide recirculation flow to the three hot legs. During normal plant operation, MOV's (890A and 890B) would be normally closed to prevent flow to the hot legs during the initial reflooding phase of emergency core coolant following a LOCA. These MOV's (890A and 890B) would be opened and 890C would be closed to provide low head hot leg recirculation flow for long term cooling. The high head centrifugal charging pumps provide injection flow thru the Boron Injection Tanks to the three cold legs via the low head cold leg injection lines. The parallel MOV's 867A&B and 867C&D located upstream and downstream of the Boron Injection Tank would be in a normally closed position. These four valves open and the charging pumps start automatically upon receipt of the safeguards initiation signal. Manual valves are located in each of the 2 inch high head cold leg branch lines inside containment for balancing the individual lines and for adjusting the pump runout during pre-operational tests. A bypass line around the Boron Injection Tank through MOV 842 is also available to provide flow from the high head charging pumps to the three cold legs. This valve, MOV 842, is normally closed and this flow path would only be used should a leak (passive failure) occur in the BIT line outside containment during the recirculation phase. The high head centrifugal charging pumps also provide recirculation flow capability to the three hot legs. Two flow paths are available from the charging pumps to a header inside the containment which distributes the flow to the three hot legs via the low head hot leg recirculation paths. During normal operation the two MOV's 869A and B located in these two flow paths would be normally closed to
C.4-3 9-15-71 prevent flow to the hot legs during the initial reflooding phase of ECC following a LOCA. These MOV's (869 A&B) would be opened and the MOV's 867A,B,C and Din the BIT line would be closed to provide high head hot leg recirculation flow for long term cooling. Manual valves are located in each of the 2 inch high head hot leg branch lines inside containment for adjusting the pump runout during pre-operational tests. For recirculation purposes there is no need to balance these branch lines to the hot legs.
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EMERGENCY CORE COOLING SYSTEM FLOW DIAGRAM
- e
C.5-1 9-15-71 c.s PEAKING FACTOR A thermal criterion in the reactor core design specifies that "no fuel melting during any anticipated normal operating condition" should occur. To meet the above criterion during a thermal overpower of 112% with additional margin for design uncertainties, a steady state maximum linear power is selected. This then is an upper linear power limit determined by the maximum central temperature of the hot pellet. The peaking factor is a ratio taken between the maximum allowed linear power density in the reactor to the average value over the whole reactor. It is of course the average value that determines the operating power level. The peaking factor becomes a constraint upon the designer, who must meet the stated value
- ~ as a limit to assure that the peak linear power density does not exceed the maximum allowed value. The operating data from a number of PWR's presented in Table C.5-1 show that there is considerable margin or conservatism in the value of peaking factor chosen for reactor design. This means that the actual peak power density is lower than that used in the core analyses, The LOCA analysis for the Surry Units No. 1 and 2 plants using the assumptions specified in the USAEC Interim Policy Criteria indicate that at 100% core power the peak linear power should be limited to 16.1 kw/ft to meet the temperature limit of 2300°F. The resulting peaking factor at turbine maximum calculated and maximum guarantee power for these plants are listed in Table C.5-2.
C.5-la 2-1-72 Further information is provided by Figure C.5-1 which shows a typical estimate of the peak clad temperature for the worst case as a function of core burnup. It is estimated that the peak clad temperature would be around 1600-1700 °Fat the end of the first cycle.
C.5-2 9-15-71 TABLE C.5-1 COMPARISON OF MEASURED AND ANALYSIS VALUES OF FN Q (1) (2) FN AS FN AS Q Q POWER AVG USED IN MEASURED PLANT (MWT) KW/FT ANALYSIS (BOL) Zorita 510 5. 04 3.14 2.19 Bernau 1130 5.08 3.14 2.21 Ginna 1300 4.88 3.28 2.01 Point Beach 1 1518.5 5. 71 2. 72 2.30 Mibama 1 1031 4.64 3.14 2.07 Robinson 2 2192 5.54 3.14 2.17 (l)An engineering factor of 3% has been excluded from the values shown. (2)A measurement uncertainty of 4.6% has been included in the value shown. The magnitude of this uncertainty is based on a 95% confidence level.
C.5-3 9-15-71 TABLE C.5-2 PEAKING FACTORS 2440 MWT 2546 MWT AVERAGE KW PER FT. 6.18 6.45 MAXD1UM KW PER FT. 16.1 16.1 PEAKING FACTOR FT 2.60 2.49 Q N 2.42 F 2.52 I I
- q N
F~H 1.50 1.50
2300 PEAK CLAD TEMPERATURE VS BURNUP DOUBLE ENDED BREAK SURRY POWER STATION 2200 2100 b.. 0 w a:: 2000 t-a:: w Q.
- E LIi.i 1900 t-
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C.6-1 2-1-72 C.6 EMERGENCY CORE COOLING SYSTEM PERFORMANCE FOR OPERATION WITH A LOOP OUT OF SERVICE This section presents an emergency core cooling analysis for a loss-of-coolant accident during operation with a loop out of service and presents only those assumptions and methods which differ from those reported in the previous sections of this appendix. The case which has been analyzed for operation with a loop out of service is the double ended serverance of the cold leg piping between the loop isolation valve and the reactor vessel of the isolated loop. This break location is the most severe because there is no means avail-able to vent steam through the hot leg of the broken loop. All ECCS injection paths remain functional during loop-out-of-service operation, since the injection lines connect to the cold legs between the isolation valves and the reactor vessel. BLOWDOWN MODEL (SATAN-V) The SATAN-V computer code model of the reactor coolant system was modified in two ways to accomodate the loop out of service loss-of-coolant analysis. First, the two operating loops were modeled individually in the SATAN analysis. Second, two control volumes were added to the model to account for the piping of the isolated loop between the reactor vessel and the loop isolation valves; that is, one control volume each for the hot and cold leg portions of the isolated loop. CORE REFLOOD MODEL The SATAN blowdown calculation is performed until the time at which zero break flow is first computed. The containment back pressure calculated for the blowdown analysis was 41.7 psia. Subsequent to blowdown, the core reflooding calculations are performed.
c.6-2 2-1-72 I The core reflood model was modified for the loop out of service loss-of-coolant analysis to account for the change in the venting path of the steam generated in the core during reflood. With one loop isolated and with the break occurring in the cold leg of the isolated loop, the steam generated in the core must be vented through the two intact loops into the downcomer region and finally out the break, The containment back pressure used for the reflood calculation was 23.1 psia. As can be seen from Figures C.6-1 and C.6-2, the reflood pressure used is that calculated at the time the peak clad temperature reaches 1640°F. The containment pressure transient used for the loop-out-of-service analysis was computed by the Stone & Webster LOCTIC code, incorporating the same conservative assumptions utilized in the LOCA analysis for three loop operation previously presented in this appendix. THERMAL MODEL (LOCTA-R2) Since this model evaluates the thermal transient for the hot channel region of the reactor core, no modifications were necessary to account for the loss-of-coolant accident during loop out of service operation. The conditions which are input to the LOCTA-R2 code, however, were based on the loop out of service situation. RESULTS The analysis of the loop out of service LOCA was performed at 69% of the core maximum calculated power of 2546 MWt and at a peak linear power of 11.1 kw/ft. The Technical Specifications allow operation at 65% of rated power with two loops. The value of peak linear power includes a 5% allowance for nuclear uncertaintied. The heat flux distribution in the hot channel, consistent with this power, was a 1.61 peak to average cosine. The peak clad temperature for this accident was calculated to be 1640°F, which is below the threshold temperature for
C.6-3 2-1-72 any appreciable clad metal-water reaction. Figures C.6-1 through 6 present the transients for significant parameters of the analysis. The following items should be noted: Figure C.6-1 The system pressure curve shown is the calculated pressure in the lower core region. The quality curve presents the values for the hot spot; this is the larger of the two values calculated in the core region. The core flow shown is the actual value for the core midplane. The average mass velocity can be obtained by dividing the flow rate 2 shown by 41.7 ft , The hot spot mass velocity is taken to be 80 percent of the average mass velocity. The heat transfer coefficient shown is for the hot spot. Figure C.6-2 This figure shows the hot spot clad temperature transient. Figure C.6-3 This figure gives the break flow rate transient. The lower plenum flow is at the bottom of the core; the upper plenum flow is at the top of the core. Figure C.6-4 This figure shows the core reflood transient. Figure C.6-5 The accumulator flow shown is the sum for the two intact loops. Figure C.6-6 Containment pressure transient. CONCLUSIONS For a loss-of-coolant accident during operation with a loop out of service, the ECCS with minimum safeguards will limit the clad temperature to much less than 2300°F and assure that the core will remain in place
C.6-4 2-1-72 and substantially intact with its essential heat transfer geometry preserved. There was no deviation from Appendix A, part 2 of the Interim Policy Statement. The ECCS design meets the AEC Interim Core Cooling Criteria.
e REACTOR PARAMETERS FOR TWO LOOP OPERATION WITH A DOUBLE ENDED COLD LEG BREAK IN THE ISOLATED LOOP SURRY POWER STATION 2000 4000 ,---, 1.0 1000 I \ I \ I \ I I I \
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- !.2-l'.)-70
- I *.
SURRY NUCLEAR STATION EMERGENCY PLAN INTRODUCTION II. GENERAL INFORMATION A. Over-All Plan B. Emergency Committee
- c. Shift Personnel D. Type of Emergencies and Phases E. Re-Evaluation and Revisions F. Accountability and Evacuation
.G. Personnel Decontamination III. RESPONSIBILITIES, DUTIES, and AUTHORITIES A. Emergency Director
- B.
c. D. Emergency Committee Individual Responsibilities Radiation Monitoring Teams E. Fire Fighting Team F. Information Center Personnel IV. TRAINING A. Emergency Plan B. Drills c~ First Aid D. Fire Fighting Team E. Health Physics
ii
- ,J.2-15-70
- v. EQUIPMENT, FACILITIES, and SERVICE AVAILABLE A. .Area Radiation Monitors (Ion Chambers)
B. Radioactive Gas and Particulate Monitors (Beta-Scintillation) C. Process Monitors D. Eberline Radiation Monitor Model RM-14 and RM-15 E. Portable Monitoring Equipment F. Emergency Alarm G. Fire Protection Systems H. First Aid Room I. Transportation J. Coordination Center K. Medical L. Communications VI. M. N. Station Facilities and Supplies Off-Site Servicas NOTIFICATION AND REPORT REQUIREMENTS A. AEC Notification B. . AEC Reports
- c. Commonwealth of Virginia Reports and Notification D. Authorization VII. DOCUMENTATION VIII. PUBLIC NEWS MEDIA
\,'
)
I-1 12-l5e-70
- Introduction SURRY NUCLEAR STATION EMERGENCY PLAN The emergency plan provides the necessary guidelines for station personnel to follow during any emergency situation. These guidelines, when coupled with the responsible judgement of station personnel, offer a flexible sequence for developing a plan of action adaptable.to existing conditions.
The station's' organization (see Figures 1 and 2) insures that a minimum of three individuals (excluding the on duty licensed reactor operators) will be available at all times to cope solely with the emergency. The guide-lines reflect the following order of action; directives will be carried out to deal with the emergency situation in accordance with this order of action:
- 1.
2.
.Protect the health and safety of the general public and station personne+.
Obtain control of the emergency situation, evaluate and minimize the detrimental effects on the station and public, J stop fires, prevent or minimize further }adiation exposures, prevent th~ spread of contamination, notify all organizations and individuals necessary, and save equipment and materials if possible.
- 3. Plan for recovery and restoration operations in an efficient and rational manner which will reduce the long-term accident effects most expeditiously.
The main objective of the plan is in accordance with management's phi-losophy to insure health and safety of the general public and station
I-2
- 12-15-70 personnel. Since time can be of the essence during a radiological emer-gency, it may become necessary to deviate from normal company policy regarding the chain of command. Appropriate notification of station personnel and department management through the levels of responsibiµty will be in a manner that is most consistent with the protection of the general public, station personnel, and plant equipment.
To insure i:naximum effectiveness, it is manadatory that each employee working at the Surry Station know the Emergency Plan and be prepared to fulfill the obligations and responsibilities so assigned him. Periodic drills :will be held and evaluated in order to satisfy these conditions.
VIRGINIA ELECTRIC AND POWER COMPANY ORGANIZATION CHART SURRY POWER STATION 1-UNIT OPERATION SYSTEM NUCLEAR SAFETY STATION MANAGER
& OPERATING COMMITTEE
.E I SUPERINTENDENT STATION' E SL OPERATIONS H SECRETARY-I
-i CLERK TYPIST I
I I I SUPERVISOR CHEMIST ANO I I ENGINEERING HEAL TH PHYSICIST OPERATING SUP ERV I SOR SUP ERV I SOR E SERVICES c SUPERVISOR I SUPERVISOR ELECTRICAL HE CHAN I CAL I E SL 1 HA I NTENAN.CE MAINTENANCE I I I CHEMICAL AND HEAL TH PHYSICS TECHNICIANS INSTRUMENT SHIFT SUPERVISOR ENGINEER SUPERVISOR SL 1 E ELECTRICAL MAINTENANCE I COHROL ROOol FOREMAN FOREMAN I ASSOCIATE I INSTRUMENT OPERATOR' ENGINEER TECHNICIANS CL I I ELECTRICIAN I E I ASSISTANT COITROL ROOII OPERA TOR MECHANIC
--- WELDER ASSISTANT ENGINEER CL I
I TRAINEE I I E tzj I-' H TRAIN'EE H NI AUXILIARY 0 I w OPERATOR
,...._ MACHINIST ~ I-'
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V, 2 I LEGEND -..J I-' 0 SL - SENIOR LICENSE ( CL - OPERATOR'S LICENSE
- LABORER E - GRADUATE ENGINEER OR EQUIVALENT C - COLLEGE GRADUATE
VIRGINIA ELECTRIC AND PDWER COMPANY ORGANIZATION CHART SURRY POWER STATION 2-UNIT OPERATION SYSTEM NUCLEAR SAFETY { & OPERATING COMMITIEE E STATION MANAGER SUPERINTENDENT I STATION E SL OPERATIONS I H SECRETARY I H CLERK TYPIST I
- I I I SUPERVISOR ENGINEERING I OIEMIST AND HEAL TH PHYSICIST SUP ERV ISOR SERVICES c SUPERVISOR OPERATING SUPERVISOR E SUPERVISOR MECHANICAL E SL , ELECTRICAL MAINTENANCE MAINTENANCE I I I I I I CHEMICAL AND HEAL TH PHYSICS TEOINIOANS SHI FT SUPERVISOR INSTRUMENT ENGINEER SUPERVISOR SL I .E ELECTRICAL. MAINTENANCE I
CONTROL ROa.l FOREMAN FOREMAN I I ASSOCIATE INSTRUMENT OPERATOR* ENGINEER TEOINICIANS OL I 2 ELECTRICIAN I E I ASSISTANT CONTROL ROOM OPERA TOR MECHANIC - WELDER ASSISTANT ENGINEER OL I 2 TRAINEE I E}}