ML20215C995
| ML20215C995 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/02/1986 |
| From: | Julie Ward SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| References | |
| JEW-86-559, NUDOCS 8610100566 | |
| Download: ML20215C995 (17) | |
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SACRAMENTO MUNlr1 PAL UTluTY DISTRICT O P. O. Box 15830, Sacramento CA 95852-1830.(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA October 2,198J JEW 86-559 Director of Nuclear Reactor Regulation Attention:
Frank J. Miraglia, Jr.
Division of PWR Licensing-B U. S. Nuclear Regulatory Conunission Washington, DC 20555 DOCKET 50-312 RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. I 10 CFR 50.71 REVISION OF FSAR
Dear Mr. Miraglia:
Several replacement pages were inadvertantly omitted from Amendment 4 of the Updated Safety Analysis Report (USAR) for Rancho Seco Unit No. 1.
This amendment was originally sent to you on July 23, 1986.
Attached are the replacement sheets which should have been included in the July 23, 1986 submittal.
If you have any questions or comments, please call Robert Little of my staff at (916) 732-6021.
Sincerely, E. Nard
[
Deputy General Manager, Nuclear Attachment cc:
S. Miner, NRC, Bethesda G. Perez, NRC, Rancho Seco l
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DISTRICT HEADOUARTERS O 6201 S Street, Sacramento CA 95817-1899
TABLE 4.2-1
/T REACTOR VESSEL DESIGN DATA
\\
r Design pressure 2,500 psig Design temperature 650*F Coolant operating temperature, inlet / outlet 556.5/607.7'F Hydrotest pressure 3,125 psig coolant volume (hot, core and internals in place) 4,058 ft3 Reactor coolant flow 369,600 gpm Number 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 Clad Coolant outlet 2
36 Carbon steel, SS Clad Core flooding - LP 2
14 Sch 140 Carbon steel (b)SS CLAD injection Control rod drive 61 2.76 Inconel ("}
I Axial 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 162,400 Studs, 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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l 4.2-3
All transients are considered as normal operating conditions and are included when determining thermal stresses and the fatigue usage factor. The fatigue analysis includes a weld strength reduction factor of two in accordance with ASME Section III.
The weld has been designed to withstand a differential pressure of 2250 psi which can occur following a core flooding line break LOCA. A dynamic magnification factor of two was applied to the pressure to account for instantaneous application. Based on these assumptions, the average shear stress in the weld yields a safety margin of 1.4.
These assumptions and safety margin are sufficient to ensure the structural integrity of the nozzle, restrictor, and weld for all operating and faulted conditions.
During the core flooding transient, the maximum P across the nozzle is expected to be approximately 200 psi.
This is more than a factor of 20 less than the design loading assumptions. During operation of the decay heat 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 reactor vessel wall, limit reactor internals and core to a vertical drop of one-half inch or less and prevent rotation of the core and internals about the vertical axis in the unlikely event of a major core barrel or core support shield failure.
The reactor vessel internals are designed to direct the coolant flow, support the reactor core, and guide the control rods throughout their full stroke.
The internals and the core are supported by the reactor vessel flange.
The centrol rod drive mechanisms are supported by the nozzles.
4.2.2.2 Steam Generator O
The steam generator general arrangement is shown in Figurc 4.2-5.
Principal design data are tabulated in Table 4.2-2.
The once-through steam generator supplies superheated steam and provides a barrier to prevent fission products and activated corrosion products from eatering the stear ystem.
The steam generator is a vertical, straight tube, tube and shell heat exchanger which produces superheated steam at constant pressure at the turbine throttle.
Reactor coolant flows downward through the tubes and transfers heat to generate steam on the shell side. The high pressure (reactor coolant pressure) parts of the unit are the hemispherical heads, the tube sheets, and the tubes between the tube sheets.
Tube support plates maintain the tubes in a uniform pattern along their length. The unit is supported by a skirt attached to the bottom head.
A method has been developed by which leaks in steam generator tubes can be sealed using a ribbed mechanical plug. This plug is held in the tube by a mandrel which expands the 3 ribs of the plug body into the tube wall to form a leak 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 side environment, and is made from heat treated Inconel 600.
The mandrel is made from 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 4.2-4 Amendment #4
1 service to prevent further tdae degradation resulting from vibration, axial loading, 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 in i
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 of the steam producing section of the vessel. Within the shell, the tube bundle is surrounded by a two-section cylindrical baffle.
The upper section of the annulus formed by the baffle plate and the shell is sealed at the lower end and serves as the superheated steam outlet.
The lower section of the baffle is the feedwater heating chamber. An cpening between the baffle sections at the feedwater inlet nozzle elevation provides a path for steam from the tube, f
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4.2-4a Amendment #4 l
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TABLE 4.2-2 STEAM GENERATOR DESIGN DATA
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(Data per Steam Generator)
's.
Steam conditions at full load, outlet nozzles Steam flow 6.12 x 106 lb/h-Steam temperature 570*F Steam pressure 910 psig Feedwater temperature 470 F Reactor coolant flow 68.94 x 108 lb/h Reactor coolant side Design pressure 2,500 psig Design temperature 650*F 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 psig Design temperature 600*F Hydrotest pressure 1,312.5 psig Net 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.
Dry weight 1,140,000 lb 1
Tube length (Less length of tube in tubesheet) 52 ft, 1-3/8 in.
Nozzles - reactor coolant side Function No.
ID, in.
Material l
Inlet 1
36 Carbon steel - SS Clad outlet 2
28 Carbon steel - SS Clad Drain 1
1 Sch 160 Inconel Manways 2
18 Carbon steel - SS Clad Handholes 2
5 Carbon steel - SS Clad Nozzles - secondary side Steam 2
24 Carbon steel Vent 1
1 1/2 Sch 80 Carbon steel (s
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-4.2-5
a TABLE 4.2-2 STEAM GENERATOR DESIGN DATA (Continued)
(Data per Steam Generator)
Nozzler - secondary side Drains 6
1 1/2 Sch 80 Carbon steel Drain 2
1 Sch 80 Carbon steel Level sensing 8
1 Sch 80 Carbon steel Temperature well 3
3/8 Inconel Manways 2
16 Carbon steel Feedwater conne:
32 Note 1 Auxiliary feedwater connect 6
3 Sch 80 Carbon steel Handholes 9
5 Carbon steel NOTES 1.
The original nozzles were made of carbon steel. New spray head assemblies made of Inconnel 600, which has higher corrosion / erosion resistance than carbon steel, were installed in 1985.
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4.2-6 Amendment #4
TABLE 4.2-4 t
PRESSURIZER DESIGN DATA Design / operating pressure 2,500/2,155 psig Design / operating temperature 670/648 F Steam volume ~
700 fta 4
Water volume 800 ft3 Hydrctest pressure 3,125 psig Electric heater capacity 1,638 kw Dimensions Overall height 44-11-3/4 ft-in.
Shell OD 96 3/8 in.
Shell minimum thickness 6.188 in.
Dry weight 304,000 lb Nozzles Function No.
ID, in.
Material Surge line 1
10 Sch 140 Carbon Steel, SS Clad ")
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Spray line 1
4 Sch 160 Carbon steel, SS Clad Carbon steel, SS Clad (c)
Safety valve 2
3 Vent 1
1 Sch 160 Inconel Sample 1
1 Sch 160 Inconel Temperature well 1
1 1/2 Inconel Level sensing 6
1 Sch 160 Inconel Heater 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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4.2-11
During cooldown and after the decay heat system is placed in service, the pressurizer can be cooled by circulating water through a connection from the discharge of the decay heat removal pump to the pressurizer spray line.
Electroslag welding was utilized in the fabrication of the pressurizer only in the longitudinal seams of the shell courses as shown in Figure 4.2-6a.
A total of three individual electroslag welds were made in the fabrication of the pressurizer.
The electroslag welding process and quality control is the same 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 4.2-2 and 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 piping from the steam generator to the reactor vessel and the 36-inch ID hot leg piping from the reactor vessel to the steam generator.
(The straight sections of the 28-inch ID and 36-inch piping are hollow-forged.) Also included in this system are the 10-inch surge line and the 2 1/2-inch spray line to the pressurizer. The system piping also incorporates the auxiliary system connections necessary for operation.
In addition to drains, vents, pressure taps, injection and temperature element connections, there is a flow meter section in each 36-inch line to the steam generators to determine the flow in each loop.
The 28-inch and 36-inch piping is carbon steel clad with austenitic stainless steel.
Short sections of 28-inch stainless steel transition piping are provided 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 that there will not be any furnace-sensitized stainless steel in the pressure boundary material. This is accomplished either by installing stainless steel safe ends after stress relief or by using Inconel.
Smaller piping, in-itnq the pressurizer surge and spray lines, is austenitic stainless steel.
piping connections in the reactor coolant system are butt-welded except tu the flanged connections on the pressurizer for the relief valves.
Thermal sleeves are installed where required to limit the thermal stresses developed because of rapid changes in fluid temperatures. They are provided in the four high-pressure injection nozzles on the reactor inlet pipes, the surge 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, radtally balanced, constant speed centrifugal pumps.
This type of pump employs a balanced stator 3 stage seal assembly to prevent reactor coolant fluid leakage to the Reactor Building atmosphere.
A view of the pump is shown in Figure 4.2-7, and the principal design parameters are listed in Table 4.2-6.
The reactor coolant pump performance characteristics are shown in Figure 4.2-8.
The reactor inlet temperature curve in Figure 4.2-9 defines the reactor coolant pump operating temperature conditions versus load.
4.2-12 Amendment #4
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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4.2-15
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TABLE 4.2-6 REACTOR COOLANT PUMP AND MOTOR DESIGN DATA (Data per pump or motor)
Pump Data Design pressure / temperature 2,500 psig/650 F Hydrotest pressure 3,125 psig Rpm at nameplate rating 1,178 Developed head 362 ft Capacity 92,400 gpm Seal water injection 9.5 gpm Controlled bleedoff 1.5 gpm Injection water temperature 120 F + 10*F Cooling water temperature 105 F Pump 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 lb Coolant volume 98 ft3 Motor Data Type Squirrel cage induction Single-speed, water-cooled Voltage 6,600 Phase 3
Frequency 60 Hz Insulation class F
Starting 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 O
4.2-16 Amendrent #4
A readout of the integrated amount of water added to the reactor coolant system, from which boric acid concentration is calculated for the purpose of
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achieving cold shutdown, is available in the control room.
As a check on this
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information, the system boron concentration can also be monitored in the purification loop as described above.
During use of the decay heat removal system, reactor coolant samples are taken at the outlet of the decay heat removal coolers. Whenever the reactor vessel head is off, reactor coolant samples normally are taken directly from the decay heat system, or can be taken directly from the system.
Reactor coolant can be sampled at system pressure through a point in the letdown system. The sample is allowed to degas into a precalibrated known volume and the pressure of the gas above the solution is then measured.
From this pressure, the total gas concentration in the sample is calculated.
System pressure is then brought to atmospheric pressure and the gas mixture is injected into the laboratory gas chromatographic system for the identification of the gas mixture.
The RCS gases are sampled weekly. Noncondensible gases are vented from the pressurizer when pressure of the system makes this desirable.
9.3.1.4 Post Accident Sampling System (PASS)
The PASS is designed to collect samples of the reactor coolant, Reactor Building 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 grab samples 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 or E-260B outlet. Reactor Building atmosphere samples are obtained from a containment atmosphere sample line.
Liquid wastes are disposed of in the reactor coolant drain tank and gaseous wastes are returned to the Reactor Building atmosphere.
9.3.2 SYSTEM DESCRIPTION The schematic flow diagram of the reactor coolant chemical addition and sampling system is shown in Figure 9.3-1.
Design data for its major components 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 acid storage tank, boric acid pumps and filter and the interconnecting piping and valves. The mix tank is equipped with a heater and mixer for the preparation of boric acid solution from demineralized water and boric acid at an elevated temperature.
The concentrated solution is drained by gravity into the storage tank 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
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Amendment #4 9.3-4
. - - _ -. -~
4 Docket No. 50-312
~
July 1986 Amendment 4 Page or Figure No.
Issue
.Section tab 8.4 Original 8.4-1 Amendment 2 Section tab 9 Original n
91 9-11 Amendment 2 9-111 1
9-iv Amendment 3 9-v Amendment 4 9-vi 9-vii Amendment 3 9-viii Amendment 4 9-ix Original 9-x 4
Section tab 9.1 9.1-1 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 4 9.2-4 Original 9.2-5 9.2-6 9.2-7 9.2-8 9.2-9 Amendment 2 9.2-10 Original 9.2-11 Fig. 9.2-1 Amendment 2 i
Section tab 9.3 Original 9.3-1 Amendment 4 9.3-la Amendment 2 j
9.3-2 Amendment 4 t
9.3-3 9.3-4 9.3-5 Original 9.3-6 9.3-7 Amendment 4 9.3-8 9.3-8a 9.3-9 Amendment 2 9.3-10 Fig. 9.3-1 Section tab 9.4 Original 9.4-1 9.4-2 9.4-3 9.4-4 Amendment 2 1
23 h
Docket No. 50-312 July 1986 Amendment 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 4 9.4-34 Original 9.4-35 Amendment 2 Fig. 9.4-1 9.4-2 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 2 9.4-17 Original 24
V approach the elasto-plastic range for the design base earthquake.
In no case are stress levels permitted to approach failure limits.
In the area of inelastic strain compatibility where plastic deformation of supports has been permitted there is no inter-relationship between' support and
- supported element deformations that would invalidate the design. The steam generator upper lateral support, for example, was designed to constrain the vessel based on system analysis wherein the spring constants of the supports were used to evaluate component response under the prescribed loading conditions.
The same approach has been-applied to the reactor coolant pumps and motors supports.
- The steam generator upper-lateral support and the reactor-coolant pump.and motor shock suppressors remain in the elastic range using elastic analysis theory. The steam generator upper lateral support was designed using ASTM A-514 Grade F steel having a minimum yield st,rength of 100 ksi.
Using the design combination of dead load, pipe rupture and design basis earthquake, the actual stresses for this maximum' loading combination are 90 percent of yield.
The shock suppressors were designed for SAE 4340 heat treated steel having a minimum yield strength of 217 ksi. Using the design combination of dead load, pipe rupture and design basis earthquake, the actual stresses for this maximum loading combination are 86 percent of yield.
4.1.4 CODE CASE INTERPRETATIONS The following ASME and ANSI-code case interpretations have been applied to
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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 the supplemental 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 for nickel-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 l
8.
Case 1338-4, Alt. 1 Ultrasonic examination of plates,Section III 9.
Case 1339-3 (Special Ruling) - Requirements for plates l
- 10. Case 1355 (Special Ruling) - Electroslag welding 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 steel precipitation hardening
- 15. Case 1401 (Special Ruling) - Welding repairs to a Class A Section III vessel after final post-weld heat treatment
- 16. Case 1415 Grain size requirements, SA-516 Plate,Section III
- 17. Case 1437 (Interpretation) - Effective code, issue for nuclear power systems.
- 18. Case 1449 (Special Ruling) - Use of USAS B31.7 1969 Code for nuclear piping.
- 19. Case 1459 Welding repairs to base metal of Class I,Section III component after final post-weld heat treatment.
- 20. Case N411 Alternative damping values for seismic analysis of piping,Section III, Division 1, Class 1, 2, and 3 construction.
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, II and 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 under abnormal conditions.
5.
Case 83R 8/71 Weld reinforcement and undercutting, Classes I, II and III nuclear power piping.
O 4.1-22 Amendment #4
Docket No. 50-312 July 1986 Amendment 4 Page or Figure No.
Issue 3.2-19 Original Section tab 3.3
- 3. 3-1 4
3.3 Amendment 4 3.3-2a f-3.3 Original 3.3-4 Amendment 4 3.3-5 Original.
3.3-6.
3.3-7 3.3-8 3.3-9 4
Section tab 3.4 f
3.4-1 Amendment 4 3.4-2 3.4-3 3.4-4 3.4-5 i
Section tab 4 Original 4-1 i
4-11 4-111 4-iv Amendment 2 l
4-v Amendment 4 4-vi Original i
4-vii 4-viii Section tab 4.1 l
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 4.1-12 4
4.1-13 4.1-14 4.1-15 4.1-16 4.1-17 Amendment 4 4.1-18 l
4.1-19 Original 2
-4.1-20 4.1-21 f
9 l
1 3
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Docket No. 50-312 July 1986 Page or Figure No.
Issue 4.1-22 Amendment 4 Section tab 4.2 Original 4.2-1 4.2-2 4.2-3 4.2-4 Amendment 4 4.2-4a 4.2-5 Original 4.2-6 Amendment 4 4.2-7 Amendment 1 4.2-8 Original 4.2-9 4.2-10 4.2-11 4.2-12 Amendment 4 4.2-13 Original 4.2-14 4.2-15 4.2-16 Amendment 4 4.2-17 4.2-18 Original 4.2-19 4.2-20 4.2-21 4.2-22 4.2-23 4.2-24 4.2-25 4.2-26 4.2-27 Amendment 1 4.2-28 4.2-29 Original 4.2-30 Amendment 1 4.2-31 4.2-32 4.2-33 Original 4.2-34 4.2-35 Amendment 2 4.2-36 4.2-37 Original 4.2-38 4.2-39 4.2-40 Amendment 1 4.2-41 Amendment 2 4.2-42 Amendment 4 4.2-42a 4.2-43 Original 4.2-44 Amendment 4 4.2-44a Amendment 2 4.2-45 Amendment 4 10