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SACRAMENTO MUNlr1 PAL UTluTY DISTRICT O P. O. Box 15830, Sacramento CA 95852-1830.(916) 452-3211AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA
October 2,198J
JEW 86-559
Director of Nuclear Reactor RegulationAttention: Frank J. Miraglia, Jr.Division of PWR Licensing-BU. S. Nuclear Regulatory ConunissionWashington, DC 20555
DOCKET 50-312RANCHO SECO NUCLEAR GENERATING STATIONUNIT NO. I10 CFR 50.71 REVISION OF FSAR
Dear Mr. Miraglia:
Several replacement pages were inadvertantly omitted from Amendment 4 of theUpdated Safety Analysis Report (USAR) for Rancho Seco Unit No. 1. Thisamendment was originally sent to you on July 23, 1986.
Attached are the replacement sheets which should have been included in theJuly 23, 1986 submittal. If you have any questions or comments, please callRobert Little of my staff at (916) 732-6021.
Sincerely,
E. Nard [Deputy General Manager,Nuclear
Attachment
cc: S. Miner, NRC, BethesdaG. Perez, NRC, Rancho Seco
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| DISTRICT HEADOUARTERS O 6201 S Street, Sacramento CA 95817-1899
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TABLE 4.2-1
/''T REACTOR VESSEL DESIGN DATA\ r
Design pressure 2,500 psigDesign temperature 650*FCoolant operating temperature, inlet / outlet 556.5/607.7'FHydrotest pressure 3,125 psigcoolant volume (hot, core and internals in place) 4,058 ft3
,
Reactor coolant flow 369,600 gpmNumber of reactor closure head studs 60,
Diameter of reactor closure head studs 6 1/2 in.Vessel dimensions
overall height of vessel and closure head 40 ft, 8-7/8 in.(*)Shell ID 171 in.Flange ID
_
165 in.'Straight shell minimum thickness 8 7/16 in.Shell cladding minimum thickness 1/8 in.Shell cladding nominal thickness 3/16 in.Insulation thickness 3 in, r
Closure head minimum thickness 6-5/8 in.r
Lower head minimum thickness 5 in.
Vessel nozzles
Function No. ID, in. Material
x,,) Coolant inlet 4 28 Carbon steel, SS CladCoolant outlet 2 36 Carbon steel, SS CladCore flooding - LP 2 14 Sch 140 Carbon steel (b)SS CLADinjection
Control rod drive 61 2.76 Inconel ("}IAxial power shaping 8 2.76 Inconel
rod drive
Instrumentation 8 3/4 Sch 160 Inconel (#)In-core instrumentation 52 3/4 Sch 160' Inconel
.
Dry weight, lb
Vessel 678,800
Closure head 162,400Studs, nuts, and washers 38,400
(* Instrument nozzle to CRD flange.
I ) With stainless steel safe end added after stress relief.(c) With stainless steel flanges.
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All transients are considered as normal operating conditions and are includedwhen determining thermal stresses and the fatigue usage factor. The fatigueanalysis includes a weld strength reduction factor of two in accordance withASME Section III. The weld has been designed to withstand a differentialpressure of 2250 psi which can occur following a core flooding line breakLOCA. A dynamic magnification factor of two was applied to the pressure toaccount for instantaneous application. Based on these assumptions, theaverage shear stress in the weld yields a safety margin of 1.4. Theseassumptions and safety margin are sufficient to ensure the structuralintegrity of the nozzle, restrictor, and weld for all operating and faultedconditions. During the core flooding transient, the maximum P across thenozzle is expected to be approximately 200 psi. This is more than a factor of20 less than the design loading assumptions. During operation of the decayheat system, the P loads on the restrictor are insignificent.
The reactor vessel contains the core support assembly, upper plenum assembly,fuel assemblies, control rod assemblies, axial power shaping rod assemblies,and incore instrumentation. Guide lugs, welded to the inside of the reactorvessel wall, limit reactor internals and core to a vertical drop of one-halfinch or less and prevent rotation of the core and internals about the verticalaxis in the unlikely event of a major core barrel or core support shieldfailure. The reactor vessel internals are designed to direct the coolantflow, support the reactor core, and guide the control rods throughout theirfull stroke. The internals and the core are supported by the reactor vesselflange. The centrol rod drive mechanisms are supported by the nozzles.
4.2.2.2 Steam Generator
OThe steam generator general arrangement is shown in Figurc 4.2-5. Principaldesign data are tabulated in Table 4.2-2.
The once-through steam generator supplies superheated steam and provides abarrier to prevent fission products and activated corrosion products fromeatering the stear ystem.
The steam generator is a vertical, straight tube, tube and shell heatexchanger which produces superheated steam at constant pressure at the turbinethrottle. Reactor coolant flows downward through the tubes and transfers heatto generate steam on the shell side. The high pressure (reactor coolantpressure) parts of the unit are the hemispherical heads, the tube sheets, andthe tubes between the tube sheets. Tube support plates maintain the tubes ina uniform pattern along their length. The unit is supported by a skirtattached to the bottom head.
A method has been developed by which leaks in steam generator tubes can besealed using a ribbed mechanical plug. This plug is held in the tube by amandrel which expands the 3 ribs of the plug body into the tube wall to form aleak tight seal.
The ribbed plug was developed by B&W for 0.625 inch OD steam generator tubes.It was designed to be compatible with the primary and secondary sideenvironment, and is made from heat treated Inconel 600. The mandrel is madefrom heat treated Inconel X-750. Details are provided in B&W Report )51-1148870-00.
Stabilizers are installed in steam generator tubes which are removed from
4.2-4Amendment #4
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service to prevent further tdae degradation resulting from vibration, axialloading, and continued corrosion. These stabilizers consist of short rigid
/'''N segments of Inconnel rod which are screwed or crimped together. Flexible
(_-) stabilizers, each consisting of a ene piece device approximately 107 inches ini length, made of Inconnel 600, have also been qualified for use in some tubes.
Details are provided in B&W Report 51-1147502-00.
i The shell, outside of the tubes, and the tube sheets form the boundaries ofthe steam producing section of the vessel. Within the shell, the tube bundle'
is surrounded by a two-section cylindrical baffle. The upper section of theannulus formed by the baffle plate and the shell is sealed at the lower endand serves as the superheated steam outlet. The lower section of the baffleis the feedwater heating chamber. An cpening between the baffle sections at
! the feedwater inlet nozzle elevation provides a path for steam from the tube,
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4.2-4a: Amendment #4'
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TABLE 4.2-2STEAM GENERATOR DESIGN DATA
(''N ' (Data per Steam Generator),
's.
Steam conditions at full load, outlet nozzlesSteam flow 6.12 x 106 lb/h-Steam temperature 570*FSteam pressure 910 psig
Feedwater temperature 470 FReactor coolant flow 68.94 x 108 lb/hReactor coolant side
Design pressure 2,500 psigDesign temperature 650*F
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Hydrotest pressure 3,125 psig
Coolant volume (Hot) 2,030 ft3'
Full load temperature inlet / outlet 557.5/606.3 F
Secondary side
Design pressure 1,050 psigDesign temperature 600*FHydrotest pressure 1,312.5 psigNet volume 3,412 ft3
Dimensions% Tubes, OD/ min wall 0.625/0.034 in.
Overall height (including skirt) 73-ft-2-1/2 in.Shell OD 151-1/8 in.Shell minimum thickness 4.1875 in.Shell minimum thickness (at tube sheetr.and feedwater connect) 6.625 in.Tube sheet thicknesses 24 in.
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Dry weight 1,140,000 lb1 Tube length (Less length of tube in tubesheet) 52 ft, 1-3/8 in.
Nozzles - reactor coolant side
Function No. ID, in. Material
| Inlet 1 36 Carbon steel - SS Cladoutlet 2 28 Carbon steel - SS CladDrain 1 1 Sch 160 InconelManways 2 18 Carbon steel - SS CladHandholes 2 5 Carbon steel - SS Clad
Nozzles - secondary side
Steam 2 24 Carbon steelVent 1 1 1/2 Sch 80 Carbon steel,,
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TABLE 4.2-2STEAM GENERATOR DESIGN DATA (Continued)
(Data per Steam Generator)
Nozzler - secondary side
Drains 6 1 1/2 Sch 80 Carbon steelDrain 2 1 Sch 80 Carbon steelLevel sensing 8 1 Sch 80 Carbon steelTemperature well 3 3/8 InconelManways 2 16 Carbon steelFeedwater conne: 32 Note 1Auxiliary feedwaterconnect 6 3 Sch 80 Carbon steelHandholes 9 5 Carbon steel
NOTES
1. The original nozzles were made of carbon steel. New spray head assembliesmade of Inconnel 600, which has higher corrosion / erosion resistance thancarbon steel, were installed in 1985.
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4.2-6Amendment #4
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TABLE 4.2-4t'
PRESSURIZER DESIGN DATA_
Design / operating pressure 2,500/2,155 psigDesign / operating temperature 670/648 FSteam volume ~ 700 fta4
Water volume 800 ft3,
Hydrctest pressure 3,125 psigElectric heater capacity 1,638 kw
DimensionsOverall height 44-11-3/4 ft-in.Shell OD 96 3/8 in.*
Shell minimum thickness 6.188 in.Dry weight 304,000 lb
.
NozzlesFunction No. ID, in. Material
Surge line 1 10 Sch 140 Carbon Steel, SS Clad ")I
I}Spray line 1 4 Sch 160 Carbon steel, SS Clad
Carbon steel, SS Clad (c)Safety valve 2 3
Vent 1 1 Sch 160 InconelSample 1 1 Sch 160 InconelTemperature well 1 1 1/2 Inconel''
;
Level sensing 6 1 Sch 160 InconelHeater bundle 3 19 1/8 Carbon steel, SS Clad
Hanway 1 18 Carbon steel, SS Clad
Electromatic relief 1 2 1/2 Carbon steel, SS Clad (c)
I")With stainless steel safe end added after stress relief.(b)With Inconel safe end.(C)With stainless flange terminal.
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During cooldown and after the decay heat system is placed in service, thepressurizer can be cooled by circulating water through a connection from thedischarge of the decay heat removal pump to the pressurizer spray line.
Electroslag welding was utilized in the fabrication of the pressurizer only inthe longitudinal seams of the shell courses as shown in Figure 4.2-6a. Atotal of three individual electroslag welds were made in the fabrication ofthe pressurizer. The electroslag welding process and quality control is thesame as described in Section 4.2.2.2.
4.2.2.4 Reactor Coolant Piping
General arrangement of the reactor coolant piping is shown in Figures 4.2-2and 4.2-3. Principal design data are tabulated in Table 4.2-5.
The major piping components in this system are the 28-inch ID cold leg pipingfrom the steam generator to the reactor vessel and the 36-inch ID hot legpiping from the reactor vessel to the steam generator. (The straight sectionsof the 28-inch ID and 36-inch piping are hollow-forged.) Also included inthis system are the 10-inch surge line and the 2 1/2-inch spray line to thepressurizer. The system piping also incorporates the auxiliary systemconnections necessary for operation. In addition to drains, vents, pressuretaps, injection and temperature element connections, there is a flow metersection in each 36-inch line to the steam generators to determine the flow ineach loop.
The 28-inch and 36-inch piping is carbon steel clad with austenitic stainlesssteel. Short sections of 28-inch stainless steel transition piping areprovided between the pump casing and the 28-inch carbon steel lines.Stainless steel or Inconel safe ends are provided for field welding the nozzle
- connections to smaller piping. The piping safe ends are designed so thatthere will not be any furnace-sensitized stainless steel in the pressureboundary material. This is accomplished either by installing stainless steelsafe ends after stress relief or by using Inconel. Smaller piping, in-''itnqthe pressurizer surge and spray lines, is austenitic stainless steel.piping connections in the reactor coolant system are butt-welded except tuthe flanged connections on the pressurizer for the relief valves.
Thermal sleeves are installed where required to limit the thermal stressesdeveloped because of rapid changes in fluid temperatures. They are providedin the four high-pressure injection nozzles on the reactor inlet pipes, thesurge line nozzle on the pressurizer, and spray line nozzle on the pressurizer.
4.2.2.5 Reactor Coolant Pumps
The reactor coolant pumps are single suction, single stage, vertical, radtallybalanced, constant speed centrifugal pumps. This type of pump employs abalanced stator 3 stage seal assembly to prevent reactor coolant fluid leakageto the Reactor Building atmosphere. A view of the pump is shown in Figure4.2-7, and the principal design parameters are listed in Table 4.2-6. Thereactor coolant pump performance characteristics are shown in Figure 4.2-8.The reactor inlet temperature curve in Figure 4.2-9 defines the reactorcoolant pump operating temperature conditions versus load.
4.2-12Amendment #4
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TABLE 4.2-5
REACTOR COOLANT SYSTEM PIPING DESIGN DATA (Continued)
Function No. ID, In. Material
on pressurizer surge piping
Drain 1 1 Sch 160 Stainless steel
on pressurizer spray piping
Auxiliary spray 1 1 1/2 Sch 160 Stainless steel
Spray valve bypass 2 1/2 Sch 160 stainless steel
(a)With stainless steel safe end added after stress relief.
I }With Inconel safe end.,
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TABLE 4.2-6REACTOR COOLANT PUMP AND MOTOR DESIGN DATA
(Data per pump or motor)
Pump Data
Design pressure / temperature 2,500 psig/650 FHydrotest pressure 3,125 psigRpm at nameplate rating 1,178Developed head 362 ftCapacity 92,400 gpmSeal water injection 9.5 gpmControlled bleedoff 1.5 gpmInjection water temperature 120 F + 10*FCooling water temperature 105 FPump discharge nozzle ID 28 in.Pump suction nozzle ID 28 in.Overall height (pump-motor) 31 ft, 5 .''/16 in.Dry weight without motor 100,000 lbCoolant volume 98 ft3
Motor Data
Type Squirrel cage inductionSingle-speed, water-cooled
Voltage 6,600Phase 3Frequency 60 HzInsulation class FStarting current 4,500 amperes
(full voltage)Power 10,000 HP (nameplate)Rotor moment of inertia 70,000 lb/f ta
Motor weight 77,200 lb
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O4.2-16
Amendrent #4
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A readout of the integrated amount of water added to the reactor coolantsystem, from which boric acid concentration is calculated for the purpose of
. "') achieving cold shutdown, is available in the control room. As a check on this(\s ,/ information, the system boron concentration can also be monitored in the
purification loop as described above. During use of the decay heat removalsystem, reactor coolant samples are taken at the outlet of the decay heatremoval coolers. Whenever the reactor vessel head is off, reactor coolantsamples normally are taken directly from the decay heat system, or can betaken directly from the system.
Reactor coolant can be sampled at system pressure through a point in theletdown system. The sample is allowed to degas into a precalibrated knownvolume and the pressure of the gas above the solution is then measured. Fromthis pressure, the total gas concentration in the sample is calculated.System pressure is then brought to atmospheric pressure and the gas mixture isinjected into the laboratory gas chromatographic system for the identificationof the gas mixture. The RCS gases are sampled weekly. Noncondensible gasesare vented from the pressurizer when pressure of the system makes thisdesirable.
9.3.1.4 Post Accident Sampling System (PASS)
The PASS is designed to collect samples of the reactor coolant, ReactorBuilding sump water, and Reactor Building atmosphere following an accident.In-line sampling is performed to determine conductivity, pH, and radioisotope,boron, chloride, and total gas concentrations. Diluted or undiluted grabsamples may be co)lected for laboratory analysis. Results are used to() determine the extent of core damage and the status of the reactor coolant
s system.
Reactor coolant samples are obtained from coolant pump P-210D suction.Reactor Building sump samples are obtained from decay heat cooler E-260A orE-260B outlet. Reactor Building atmosphere samples are obtained from acontainment atmosphere sample line.
Liquid wastes are disposed of in the reactor coolant drain tank and gaseouswastes are returned to the Reactor Building atmosphere.
9.3.2 SYSTEM DESCRIPTION
The schematic flow diagram of the reactor coolant chemical addition andsampling system is shown in Figure 9.3-1. Design data for its majorcomponents are given in Table 9.3-3.
9.3.2.1 Chemical Addition System
The boric acid addition system consists of a boric acid mix tank, boric acidstorage tank, boric acid pumps and filter and the interconnecting piping andvalves. The mix tank is equipped with a heater and mixer for the preparationof boric acid solution from demineralized water and boric acid at an elevatedtemperature. The concentrated solution is drained by gravity into the storagetank where it is diluted to a concentration of 4 percent. During a reactor
I ' shutdown two centrifugal boric acid pumps inject boric acid solution into the(>)
Amendment #49.3-4
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Docket No. 50-312~ July 1986
Amendment 4
Page or Figure No. Issue
.Section tab 8.4 Original'
8.4-1 Amendment 2Section tab 9 Original
n' 919-11 Amendment 2
"9-1111
9-iv Amendment 3*
9-v Amendment 4"
9-vi9-vii Amendment 39-viii Amendment 49-ix Original
"9-x4
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Section tab 9.19.1-1 Amendment 4
"'9.1-2
Fig. 9.1-1 Original"
Section tab 9.2,
"9.2-1-
9.2-2 Amendment 2,
] 9.2-3 Amendment 49.2-4 Original'
"9.2-5
"9.2-6
"9.2-7
"9.2-89.2-9 Amendment 29.2-10 Original
"9.2-11
Fig. 9.2-1 Amendment 2,
i Section tab 9.3 Original9.3-1 Amendment 49.3-la Amendment 2
j- 9.3-2 Amendment 4
"t 9.3-3
"9.3-49.3-5 Original .
"9.3-6'
9.3-7 Amendment 4"
9.3-8;"'
9.3-8a9.3-9 Amendment 2
"9.3-10;
"Fig. 9.3-1Section tab 9.4 Original
"9.4-1
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Docket No. 50-312July 1986Amendment 4
Page or Figure No. Issue
9.4-5 Original"9.4-6"9.4-7"
9.4-8"9.4-9"
9.4 10"9.4-11"
9.4-12"
9.4-13"
9.4-14"
9.4-15"9.4-16"9.4-17"9.4-18"9.4-19"9.4-20"9.4-21"9.4-22"9.4-23"9.4-24"9.4-25"9.4-26"9.4-27
9.4-28 Amendment 2"9.4-28a
9.4-29 Original"9.4-30"9.4-31"9.4-32
9.4-33 Amendment 49.4-34 Original9.4-35 Amendment 2
"Fig. 9.4-19.4-2 Original
"9.4-3"9.4-4"9.4-5"9.4-6"9.4-7"9.4-8"9.4-9"9.4-10"9.4-11"
9.4-12"9.4-13"9.4-14"9.4-15
9.4-16 Amendment 29.4-17 Original
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approach the elasto-plastic range for the design base earthquake. In no caseare stress levels permitted to approach failure limits.
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In the area of inelastic strain compatibility where plastic deformation ofsupports has been permitted there is no inter-relationship between' support and
- supported element deformations that would invalidate the design. The steamgenerator upper lateral support, for example, was designed to constrain thevessel based on system analysis wherein the spring constants of the supportswere used to evaluate component response under the prescribed loadingconditions. The same approach has been-applied to the reactor coolant pumps
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and motors supports.
- The steam generator upper-lateral support and the reactor-coolant pump.andmotor shock suppressors remain in the elastic range using elastic analysistheory. The steam generator upper lateral support was designed using ASTMA-514 Grade F steel having a minimum yield st,rength of 100 ksi. Using thedesign combination of dead load, pipe rupture and design basis earthquake, theactual stresses for this maximum' loading combination are 90 percent of yield.
The shock suppressors were designed for SAE 4340 heat treated steel having aminimum yield strength of 217 ksi. Using the design combination of dead load,pipe rupture and design basis earthquake, the actual stresses for this maximumloading combination are 86 percent of yield.
4.1.4 CODE CASE INTERPRETATIONS
The following ASME and ANSI-code case interpretations have been applied to
() components and piping within the reactor coolant pressure boundary:
A. INTERPRETATIONS OF ASME. BOILER AND PRESSURE VESSEL CODE
1. Case 1141-1 (Special Ruling) - Foreign produced steel (with thesupplemental provision that.the arsenic level of the-
j steel must be below 0.1%).
2. Case 1332-4 (Special. Ruling) - Requirements for steel forgings.
3. Case 1335-2 (Special Ruling) - Requirements for bolting materials.
i 4. Case 1335-3 (Special Ruling) - Requirements for bolting materials.-
i 5. Case 1336 (Special Ruling) - Requirements fornickel-chromium-iron alloy (all product forms)
i
6. Case 1338-3, Alt. 1 Ultrasonic examination of plates, Section III
f 7. Case 1338-4, Ultrasonic examination of plates, Section III *
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8. Case 1338-4, Alt. 1 Ultrasonic examination of plates, Section III
; 9. Case 1339-3 (Special Ruling) - Requirements for plates.
| 10. Case 1355 (Special Ruling) - Electroslag welding,
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'4.1-21
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11. Case 1355-2 (Special Ruling) - Electroslag welding
12. Case 1359-1 (Special Ruling) - Ultrasonic examination of forgingsSection III
13. Case 1361-1 (Special Ruling) - Socket welds - Section III
14. Case 1388 (Special Ruling) - Requirements for stainless steelprecipitation hardening
15. Case 1401 (Special Ruling) - Welding repairs to a Class ASection III vessel after final post-weld heattreatment
16. Case 1415 Grain size requirements, SA-516 Plate, Section III,
17. Case 1437 (Interpretation) - Effective code, issue for nuclearpower systems.
18. Case 1449 (Special Ruling) - Use of USAS B31.7 1969 Code fornuclear piping.
19. Case 1459 Welding repairs to base metal of Class I, Section IIIcomponent after final post-weld heat treatment.
20. Case N411 Alternative damping values for seismic analysis ofpiping, Section III, Division 1, Class 1, 2, and 3construction.
B. INTERPRETATIONS OF THE CODE FOR PRESSURE PIPING ANSI B31
1. Case 81 9/70 Class I nuclear power piping 2 in, and smaller.
2. Case 83 10/70 Weld reinforcements and undercutting, Classes I, IIand III nuclear power piping.
3. Case 69 1/70 Partial penetration welds in nuclear piping
4. Case 70 1/70 Design criteria for nuclear power piping underabnormal conditions.
5. Case 83R 8/71 Weld reinforcement and undercutting, Classes I, IIand III nuclear power piping.
O4.1-22
Amendment #4
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Docket No. 50-312July 1986Amendment 4'
Page or Figure No. Issue
3.2-19 Original"
Section tab 3.3"- 3. 3-14
3.3-2- Amendment 4"
3.3-2a
f- 3.3-3- Original3.3-4 Amendment 43.3-5 Original.
"3.3-6.
"3.3-7
"3.3-8
"3.3-94
'"Section tab 3.4
f3.4-1 Amendment 4
"3.4-2
"3.4-3
"3.4-4"
3.4-5.
i Section tab 4 Original"
4-1i "4-11
"4-111;
4-iv Amendment 2|'
4-v Amendment 4! 4-vi Original
"i 4-vii
"4-viii
"Section tab 4.1
"| 4.1-1
"4.1-2
"' 4.1-3"
i- 4.1-4"
4.1-5"
4.1-6"
i 4.1-7"
l 4.1-8"4.1-9"
4.1-10<"
4.1-11"
4.1-124
"4.1-13'
"4.1-14"' 4.1-15
, "4.1-164.1-17 Amendment 4
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Docket No. 50-312July 1986
Page or Figure No. Issue
4.1-22 Amendment 4Section tab 4.2 Original
"4.2-1
"4.2-2
"4.2-34.2-4 Amendment 4
"4.2-4a4.2-5 Original4.2-6 Amendment 44.2-7 Amendment 14.2-8 Original
"4.2-9
"4.2-10
"4.2-114.2-12 Amendment 44.2-13 Original
"4.2-14
"4.2-154.2-16 Amendment 4
"4.2-174.2-18 Original
"4.2-19
"4.2-20
"4.2-21
"4.2-22
"4.2-23
"4.2-24
"4.2-25
"4.2-264.2-27 Amendment 1
"4.2-284.2-29 Original4.2-30 Amendment 1
"4.2-31
"4.2-324.2-33 Original
"4.2-344.2-35 Amendment 2
"4.2-364.2-37 Original
"4.2-38
"4.2-394.2-40 Amendment 14.2-41 Amendment 24.2-42 Amendment 4
"4.2-42a4.2-43 Original4.2-44 Amendment 44.2-44a Amendment 24.2-45 Amendment 4
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