ML19322A787

Chapter 1 of Oconee 1,2 & 3 PSAR, Introduction & Summary. Includes Revisions 1-6

ML19322A787
Person / Time
Site:	Oconee
Issue date:	12/01/1966
From:	DUKE POWER CO.
To:
References
NUDOCS 7911250005
Download: ML19322A787 (58)

Text

i O

~

l ll 1

l O 5 991125 0 00R g 4 00 00037

TABLE OF CONTENTS Section Pm 1 INTRODUCTION AND

SUMMARY

1-1

1.1 INTRODUCTION

1-1 1.2 DESIGN HIGHLIGHTS 1-2 1.2.1 SITE CHARACTERTISTICS 1-2 1.2.2 POWER LEVEL l-2 1.2.3 PEAK SPECIFIC POWER LEVEL 1-2 1.2.4 FUEL SHARING 1-2 1.2.5 REACTOR BUILDING SYSTEM 1-2 1.2.6 ENGINEERED SAFEGUARDS SYSTEM l-3 1.2.7 ELECTRICAL SYSTEMS AND EMERGENCY POWER l-4 1.2.8 ONCE-THRO'IGH STEAM GENERATORS 1-4 1.3 TABULAR CHARACTERISTICS 1-5 1.4 PRINCIPAL DESIGN CRITERIA 1-15 1.4.1 CRITERION 1 1-15 1.4.2 CRITERION 2 1-17 1.4.3 CRITERION 3 1-17 1.4.4 CRITERION 4 1-18 1.4.5 CRITERION 5 1-18 1.4.6 CRITERION 6 1-19 1.4.7 CRITERION 7 1-19 l

1.4.8 CRITERION 8 1-20 1.4.9 CRITERION 9 1-21 1.4.10 CRITERION 10 1-22 1.4.11 CRITERION 11 1-22 1.4.12 CRITERION 12 1-23 -

00 00038 1-1

Section 7 age 1.4.13 CRITERION 13 1-24 1.4.14 CRITERION 14 1-25 1.4.15 CRITERION 15 1-25 1.4.16 CRITERION 16 1-26 1.4.17 CRITERION 17 1-27 j 1.4.18 CRITERION 18 1-28 1.4.19 CRITERION 19 1-29 1.4.20 CRITERION 20 1-30 1.4.21 CRITERICN 21 1-30 1.4.22 CRITERION 22 1-30 1.4.23 CRITERION 23 1-31 1.4.24 CRITERION 24 1-32 O 1.4.25 CRITERION 25 1-32 l

1.4.26 CRITERION 26 1-33 1.4.27 CRITERION 27 1-33 1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS 1-34 1.5.1 ONCE-THROUGH STEAM GENERATOR TEST 1-34 1.5.2 CONTROL ROD DRIVE LINE TEST 1-34 1.5.3 SELF-POWERED DETECTOR TESTS 1-34 1.5.4 THERMAL AND HYDRAULIC PROGRAMS 1-35 1.6 IDENTIFICATION OF CONTRACTORS 1-35

1.7 CONCLUSION

S 1-36 00 00139 H 1-11 '

~

i

O LIST OF TABLES i

Table Pg Enc.geseev~co{ Safegua'Y I S ".9 'Y'"'"y 1-1 Ch: "agin:; red r-f:;.a d. i.i.e. 1-4  !

1-2 Comparison of Design Parameters 1-8 LIST OF FIGURES Figure 1-1 Service Area Map 1-2 Station General Arrangement, Floor Plan Elevation 775+0 1-3 Station General Arrangement, Floor Plan Elevation 785+9 l-4 Station General Arrangement, Floor Plan Elevation 796%

3 1-5 Station General Arrangement, Floor Plan Elevation 809+3 1-6 Station General Arrangement, Floor Plan Elevation 822+O 1-7 Station General Arrangement, Floor Plan Elevation 837 M and Elevation 841%

i 1-8 Station General Arrangement, Floor Plan Elevation 857+0

~

1-9 Station General Arrangement, Cross Section 1-10 Reactor Building, Basement Floor, Elevation 777%

l-11 Reactor Building, Ground Floor, Elevation 797%

l-12 Reacter Building, Upper Canal Floor, Elevation 816%

l-13 Reactor Building, Operating Floor, Elevation 841%

(and Above) l 14 Reactor Building, Sectional View Looking North 1-15 Reactor Building, Sectional View Looking East O "

00 00040 1-111 (Revised 4-18-67)

1 INTRODUCTION AND

1.1 INTRODUCTION

This Preliminary Safety Analysis Report is submitted in support of Duke Power Company's application for a construction permit and facility license for the two-unit Oconee Nuclear Station to be located on the shore of future Lake Keowee in Oconee County, South Carolina. The station location is shown on Duke's Service Area Map, Figure 1-1.

Each generating unit will operate initially at core power levels up to 2452 w t which corresponds to a net electrical output of about 839 me. All phy-sics and core thermal hydraulics information in this report is based upon the reference core design of 2452 mwt. It is expected that each unit will be capable of an ultimate output of 2584 mwt (including 16 mut contribution from reactor coolant pumps), corresponding to a net electrical capability of about 874 mwe. Site parameters, principal structures, engineered safeguards and certain hypothetical accidents are evaluated for the expected ultimate core output of 2568 mwt.

The nuclear steam supply system is a pressurized water reactor type sir.ilar to systems operating or under construction. It uses chemical shim and ;ontrol rods for reactivity control and generates steam with a small amount of super-heat in once-through steam generators. The nuclear steam supply system and two fuel cores for each of the two units will be supplied by The Babcock &

C Wilcox Company.

Construction is scheduled for completion in time for loading fuel into Unit 1 in December 1970 and for its commercial operation by May 1971, with commercial operation of Unit 2 scheduled by May 1972. To meet this schedule, construc-tion of the Reactor Building is to begin by September 1,1967. The general arrangement of major equipment and structures, including the Reactor, Auxiliary and Turbine Buildings, is shown on Figures 1-2 through 1-14. l l

The organization of this report follows as closely as possible the AEC's l

" Guide" announced in the Federal Register on August 16, 1966. Every attempt has been made in this report to be completely responsive to that guide, to the proposed AEC design criteria, and to all known pertinent questions asked of other applicants up until the time of this writing.

As the station design progresses from conceptual design to final detailed design, the station description and analyses will be subject to change and refinement. This report presents descriptive material and analyses of a

" reference desien." Any significant changes to the criteria or designs which affect safety will be promptly brought to the attention of the AEC by suitable supplements.

Duke is fully responsible for the complete safety and adequacy of the station, and, consistent with long-standing practice, Company personnel will design, construct, test, start and operate the units. Assistance in performing these p functions will be rendered principally by B & W and by Duke's general consult-Q ant, Bechtel Corporation, along with such other consultants and suppliers as may be required. The technical qualifications of' Duke, B & W and Bechtel are "

outlined in Appendix 1A.

00 00041 1-1

1.2 DESIGN HIGHLIGHTS 1.2.1 SITE CHARACTERISTICS The site is characterized by a one-mile exclusion radius; remoteness from population centers; sound, hard rock foundation for structures; freedom from flooding; an abundant supply of cooling water; an on-site hydroelectric sta-tion capable of supplying ample emergency power; and favorable conditions of hydrology, geology, seismology and meteorology. The proximity of the hydro tailrace offers the unusual capability of providing emergency, powerless water flow by gravity through the Oconee condensers. This reliable heat sink is available for rejection of decay heat conveyed by natural circulation in the reactor coolant system and steam-driven pumps in the secondary system.

1.2.2 PORER LEVEL Initially licensed power for each reactor core is proposed at 2452 mut, and core performance analyses in this report are based upon this initial power level. Operating confirmation of reactor core parameters is expected to support an ultimate core power level of 2568 mwt, and all steam and power conversion equipment is designed to operate at this output. The analyses of accidents which could release fission products to the environment have been evaluated on the basis of 2568 mwt. An additional 16 mwt will be available to the cycle from the contribution of the reactor coolant pumps, resulting in a net electrical output of about 839 mw at initially licensed power and 874 mw ultimately.

1.2.3 PEAK SPECIFIC PORER LEVEL ./

The peak specific power level in the fuel for initial operation at 2452 mwt results in a maximum thermal output of 17.5 kw per ft of fuel rod. This value is comparable with other reactors of this size and therefore does not repre-sent an extrapolation of technology. This comparison may be seen in the information presented in Table 1-2 in 1.3.

1.2.4 FUEL SHARING Irradiated fuel removed from Unit 1 at the end of its first cycle will consti-tute about one-half the initial core of Unit 2, with attendant economies in fabricating and reprocessing costs. Fuel sharing has no adverse effect upon public health and safety. Core-performance and reactivity-control data for the fuel sharing plan are given in this report.

1.2.5 REACTOR BUILDING SYSTEM J l

The system required to contain the Maximum Hypothetical Accident consists of the Reactor Building envelope and the engineered safeguards.

The prestressed, post-tensioned concrete Reactor Building is of essentially l the same design as the containment buildings for the Turkey Point Plant (

(Docket Nos. 50-250 and 251), the Palisades Plant (Docket No. 50-255) and the l Point Beach Plant (Docket No. 50-266). Several of the engineered safeguards j are similar to those plants and Oconee presents neither uncommon solutions _j to engineering problems nor significant extrapolations in technology.

• II Ch hC 1-2 Ct/[

l

)

1.2.6 ENGINEERED SAFEGUARDS SYSTEMS Engineered saferuards systems are employed to reduce the potential radiation dose to the general public from the Maximum Hypothetical Accident to less than 1 the guideline values of 10 CFR 100. This is acccmplished by automatic isola- l tien of all Reactor Building fluid penetrations that are not required for l limiting the consequences of the accident, thus eliminating potential leakage  !

paths. Long term potential releases following the accident are reduced by l rapidly decreasing the Reactor Building pressure to near atmospheric within l 2h hrs, thereby reducing the driving potential for fission product escape. l 1

In addition, the engineered safeguards systems will prevent core meltdown upon the worst postulated loss-of-coolant accident. This is accomplished by large capacity, injection core flooding systems. These systems, coupled with the thermal, hydraulic and blowdown characteristics of these reactors, will reliably i prevent metal-water reactions, i i

Each reactor unit will have the following engineered safeguards equipment, I with the normal operating mode of each as indicated:

(a) High pressure injection system - a portion normally operates.

(b) Lcw pressure injection system - operates for shutdown cooling.

(c) Core flooding tanks - normally ready for operation.

O d

(d) Reactor Building spray system - normally shutdown. 1 l

(e) Reactor Building emergency ecolers - normally shutdown.

(f) Penetration room ventilation system - normal intermittent operation.

l (F) Reactor Building isolation system - operates on test or accident signal.

The engineered safeguards systems are independent for each unit. The following Table 1-1 lists the major equipment in each system. ,

l l

1 l

l O

00 00043 1 1

1-3 (Revised 4-1-67)

Table 1-1 s Engineered Safeguards Equipment Total Equipment System Ins talled/ Unit High Pressure 3 pumps Injection System 1 storage tank Low Pressure 3 pumps Injection System 2 heat exchangers Core Flooding Tanks 2 tanks -

Reactor Building 2 pumps Spray System 2 spray headers Reactor Building 3 coolers Coolers 1.2.7 ELECTRICAL SYSTEMS AND EMERGENCY POWER Each of the two units at Oconee will have five sources of electric power, each source having several degrees of redundancy:

(a) Its own generator which will continue to supply its auxiliary loads upon a trip separating the generator from the transmission system.

(b) Six 230 kv transmission lines emanating from Oconee in three directions.

(c) The other unit's generator (available after completion of Unit 2).

(d) One 100 kv transmission line.

(e) The on-site Keowee Hydroelectric Station having two, quick-starting 70,000 kw generating units connected to Oconee by one overhead 230 kv line and one underground 13.8 kv cable.

Within Oconee will be multiple redundant busses and tie busses supplying power, loads, instruments and controls. For each unit, the engineered safe-guards, generally arranged on a three-ccmponent basis, are supplied from three separate auxiliary power busses, each of which can be supplied from any of the five principal sources of power.

The sources of power and associated electrical equipment will insure safe functioning of the station and its engineered safeguards.

1.2.8 ONCE-THROUGH STEAM GENERATORS The steam generators are of a new design based on extensive research, develop-li f t -

Olly 1-4 (Revised 4-1-67)

p

() ment and experimental work on boiling heat transfer performed by B & W over the past 10 years. Each generator consists of a vertical shell-and-tube, counterflow heat exchanger with reactor coolant on the tube side and steam on the shell side. Feedwater is pumped into the generator, heated to saturation by direct mixing with steam, converted to steam and superheated in a single pass through the generator. The basic design parameters such as feedwater heating, boiling length, superheat length and performance characteristics have been confirmed by testing of a full-length, 7-tube section and a 37-tube section. Tests are continuing to obtain cdditional data with the larger test section.

With the once-through design, natural circulation flow is adequate to remove full decay heat without the use of reactor coolant pumps. Thus, vith total loss of pumps, the fuel will not reach departure from nucleate boiling.

1.3 TABULAR CHARACTERISTICS Table 1-2 is a comparative list of important design and operating character-istics of Duke's Oconee Nuclear Station Units 1 and 2, Turkey Point Units 3 and 4 (Florida Power and Light Company), Indian Point Station Unit 2 (Consolidated Edison Company of New York, Inc.), and Brookwood (Rochester Gas

& Electric Company) nuclear power stations. These stations have design and operating parameters close to those of the Duke facility.

s The data contained in Table 1-2 represent information presented in available

() station descriptions, and Safety Analysis Reports submitted for licensing. j The design of each of these stations is based upon information developed from operation of commercial and prototype pressurized water reactors over a num-ber of ye ars. The Oconee design is based upon this existing power reactor technology, and has not been extended beyond the boundaries of known informa-tion or operating experience.

The similarities and differences of the features of the reactor stations listed in Table 1-2 are discussed in the following paragraphs. In each case the item number used refers to the item numbers used in the table.

Item 1. Hydraulic and Thermal Design Parameters Most of the parameters listed in this section are similar for each reactor facility, differing according to the thermal power level. The differences of power level are reflected chiefly in the total heat output, core size (fuel loading), coolant flow rate, and total heat transfer surface. They amount only to o scaling down of the above parameters for a decrease in the thermal reactor power level, and do not alter the safety-related characteristics of the reactors. The Departure from Nucleate Boiling Ratio (DNBR) and the maxi-mum ratio of peak to average total heat input per fuel pin (Feh nuc.) are re-presentative of a more conservative design for Oconee than for the other re- l actors presented. These comparisons are discussed in detail in 3.2.3.2.

f-'S Item 2. Core Mechanical Design Parameters U The dimensions, materials, and technology for each of the reactors in this k[

00 002S 1-5

section are similar. This uniformity is again due to optimtzation of the op-erating parameters for this type of reactor, and differences are related to the power levels.

There are also small differences in the mechanical assembly of the fuel rods and the number of control rods used in the individual reactors. The increased number of control rods in the Oconee reactors provides for maneuverability and flexibility cf operation. Oconee utilizes a canned fuel assembly which pro-vides structural integrity and protection of the fuel rods against damage during fuel handling operations.

Item 3. Preliminary Nuclear Design Data Since these reactors have essentially the same core geometrical configuration, the fuel loading differs by an amount that is proportional to the physical size of the reactor core. The .hii 0/u 2 raci thc unit--ceM-Tii'~the Oconee-reactor h gn-is-due-to-the-slightly-greater number of rods in a fuel assembly-and-higher fuel density.-

Each core has a three region fuel loading, but differs in the fuel burnup ratio that is to be used. A choice can be made in the fuel rotation and discharge plan, which, in the Oconee reactors, leads to a relatively low fuel burnup in the first cycle.

The Duke reactor design offers about 3 per cent greater reactivity control in the control rods. This is also reflected in the lesser concentration of boron that is required to control the reactivity over the lifetime of the reactor core. Some slight differences are noted in reactivity coefficients.

Item 4. Principal Design Parameters of the Reactor Coolant System Most of the features in this section are directly related to material proper-ties and the amount of heat produced in the reactor core. The parameters are scaled proportional to the power of the reactor. The major differe2ce is in the number of coolant loops required to remove the heat produced.

For Oconee, only two loops are required since once-through steam generators are used instead of the U-tubes-in-shell design. The greater cooling capacity of these steam generators permits a reduction in the number of cooling loops for an equivalent amount of heat removed.

Item 5. Reactor Coolant System - Code Requirements Code requirements for each system are the same except for the shell side of the steam generator which conforms to the AS}E III Class A specification, whereas in the other reactors they are ASTE III Class C. This is considered to be a contribution to the safety of the vessel as it enhances the integrity because of the more stringent ASFE III Class A design, material, and quality control requirements.

Item 6. Principal Design Parameters of the Reactor Vessel The Oconee vessel design is characterized by a thinner thermal shield and a 1

% 0% I 1-6

relatively large dia. meter. This larger diameter provides for additional water between the edge of the core and the vessel which leads to additional neutron attenuation.

Item 7. Principal Design Features of the Steam Generators The steam generators in the Duke facility are significantly different than the I I

other facilities since they are of the once-through design and incorporate an integral superheat section. j Item 8. Principal Design Parameters of the Reactor Coolant Pumps In each specific parameter the relative number or size is in proportion to the total amount of heat that is removed from the core. The Oconee pumps have higher head and horsepower requirements for approximately the same flow ,

because of the increased flow losses of the once-through steam generators and  ;

the use of only two reactor coolant loops.

Item 9. Principal Design Parameters of the Reactor Coolant Piping The Oconee piping utilizes carbon steel clad with stainless steelsh.- -th -

rci;;ht see t im .

Item 10. Reactor Building Parameters N

] The Oconee Reactor Building is basically the same design and construction as the Turkey Point units. Differences are physical dimensions, amount of co,- n crete shielding needed, and design incident pressures. These differences are a direct result of station layout, engineered safeguards, system capacitites, and site location. The Reactor Building design and shielding offer satis-factory protection to the surrounding population in case of an accident and during normal operation of tie generating units.

Item 11. Engineered Safeguards Engineered safeguards are generally similar, except Oconee includes a pene-tration room ventilation system for each unit (6.5) . Differences in charac-teristics and capacities are due to strong dependence on site. The avail-ability of hydroelectric power for emergency use provides adequate electric power to operate all emergency equipment if offsite power is lost.

O d

00 0004t -

1-7

("% f '\

'w) v/

4 k \ \

f Q/

Table 1-2 Cam,.xarison of Design Parameters (Nr Unit Lusis Unless Noted)

Item

" hn N*k#"# htsti d Turkey Point No. 3 or 4 Indian Point No. 2 Brookwood 1 Hydraulic and Thermal feste Parameters Dated IIeat Output, sr.it 2,k52 2,07f latte.1 IIcat Output, Btu /hr 8,369 x lob 2,758 1,300 f t4ximan Overpower, f.

7,15T x 106 9,423 x 106 4,437 x 106 14 12 System Pressure (nominal), psia 12 12 2,200 2,250 2,250 System Pressure (miniatus steady state), paia .@ 9. / fo 2,220 2,2 >o Power Distribution Factors ' 2,220 2,200 Ifeat Generated in Fuel and Cla.idin6, 5 'll.3 FI 4 rg(nuclear) 1.85 1.75 17.4 FI.%

Fg(nuclear) 1 75 1 75 3 15 3 12 Hot Channel Factors 3 12 3 28 Fg (nue. and mech.) 3 24 3 25 DNB k tio at Rat,4:4 Conditions 3 25 3 41 2.27(W-3) 1.85(W-3) 1.81(W-3) 1 90(W-3)

Y Minimum DIIB Ratio at Desi;n Over;nwcr 1.60 BAW-168) 1 73 W-3) 03 1 30(W-3) 1 38 BAW-168) 13o(w-3) 1 30(W-3)

Coolant Flow Total Flow Rate, Ib/hr 131 3 x 2ffective Flow 120 9 x 1 100.6xif 136.2xick 67.1 x 1 Effective Flow Rate Area for for IIeat

!! eatTransfer.

Transfer,Ib/hr ft" 47 75 91 39 0 5 x 10 324.1 x 106 61.1 x 1 Average Velocity Along Fuel Bods, ft/sec 48.h 25 1 15 70 33 9 16.1 Average Mass Velocity, ib/hr-ft2 3,$3 x 106 15.1 Coolant Temperatures, F 2 35 x 106 2 56 x 106 2.43 x IOb Nominal Inlet 555 tbximum Inlet due to Instrumentation 5k6.5 S3 556 Error and Deadband 557 Average Rise in Vessel 550.5 547 560 Average Rise in Core 47 8 54 53 49 Q Average in Core 49 3 59 57 579 7 54 Q Average in Vessel 578 9 577 572 7 584 Nominal Outlet of !!ot Channel 574 570 581 644.4 647 C Average Film Coefficient, Btu /hr-ft2-F 5,000 5,500 643 644 Average Film Temperature Difference, F 5,900 5,830 O 31 30 30 30 C

A N

9 m

,i -.

L

._a_______.

b N

's ., '%/

tnLie 1-2 (Cont'd )

(ftr Ur.1t Basis Unless Nated)

Item

• Un[ r Turkey Point No. 3 or 4 Indian Point No. 2 Bruokwood Heat Transfer at 100% Power Active Heat Transfer Surface Area, ft2 48,578 42,460 52,200 28,500 Avera6e Heat Flux, Btu /hr-ft2 167,620 164,200 175,600 Haxiasm Heat Flux, Btu /hr-rto 543,000 151,800 533,600 570,800 $17,500 Average Thermal Output, kw/ft 5.4 53 h imum Thermal Output, kv/ft 17 5 17 3 57 49 himum Clad Surface Tearperature at 18.5 16 7 Nominal Pressure, F 6 54 657 Fuel Central Temperature, F 659 659 Maximum at 100% Power 4,160 4,070 4,1W htmum at 114% Overpower 3,920 4,400 4,270 h,250 Thermal Output, kv/ft at himum Overpower b 150 19 9 19.4 20 7 lb.7 2 Core Mechanical Design Parameters Fuel Assemblies Design RCC can RCC canless u HCC canless BCC canless Bod Pitch, in. 0.558 e overall Dimensions, in 0.563 0 563 o.556 8.522 x 8.522 d.b26 x 8.h26 8.%26 x 8.b26 4 Fuel weight (as ug), Ib Total Weight, Ib 201,520 283,200 179,000 215,3'O 7 763 x 7 763 117,530 226,200 273,410 151,630 Number of Gride per Assembly 6 t1 8 Fuel Bods 8 Number 36,016 y ,028 39,372 21,480 Outside Diameter, in, 0.k20 0.422 Diametral Csp, in. 0.006 0.422 0.k22 C Clad Thickness, in. 0.Cc6 0.0065 I 0.0065 0.0074 C Clad mterial Zircaloy 0.Cc43 Zircaloy 0.Tk3 0.0243 Fuel Pellets Zirealoy Zircaloy mterial O Density, ') of theoretical UO2 sintered UO2 sintered UO2 sintered IXh sintered O 95 94-93 Diameter, in. O.362 94-93 9C93 Q Iength, in. 0 3669 0 3669 0 366 0.600 0.600 0.600 4 Control Rod Cluster Assew11es 0.600 Neutron Absorber $ Cd-15% In-80% Ag Cladding m terial 5%Cd-l% In-80% A6 5% Cd-15% In-80% Ag 5% Cd-15% In 80% Ag 304 SS - cold worked 304 SS - coLi worked 304 SS - cold worked C hd Thickness, in. 0.018 304 SS - cold worked Number of Clusters 0.019 0.019 0.019 69 41 Number of Centrol Pins per cluster 16 20 53 32 Core Structure 20 16 Core Barrel ID/0D, in. 147/150 N Thermal Chield ID/OD, in. d'ff/16F /$' ((p/ 133 5/137 25 148.5/152 5 109 0/112 5 141.0/147 5 158 5/164.0 114.5/122 5 H

b O

8 t

O.

%,e

- __ _ ___ A ___ _ - _- --

r_.____ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _

\ )

'Q  %.)

?able 1-2 (Cont'd )

(Fer Unit Basis Unless Noted)

Item D " 1e t12' Turkey Point No. 3 or 4

%n t 1 Indian Point No. 2 Brookwood 3 Preliminary Naclear Design Data Stmetural Characteristics Fuel Weight (as UO 2 ), Ib 201,520 179,000 215,320 117.527 Clad Weight, Ib 43,000 35,600 43,785 24,208 Core Diameter, in. (equivalent) 128 9 119 5 134 97 Core Height, in. (active fuel) 144 144 144 144 Reflector Thickness and Composition Top (water plus steel),in. 12 10 10 lo Botton (water plus steel), in. 12 10 10 10 Side (water plus steel), in. 18 15 15 15 M 0/U 2 (unit cell - cold) 2.54 3.48 3 48 s. 31 Number or Fuel Assenh11es 177 157 193 uo f' . # r b /j~#3 e= /r/,t -Atemte-ftp/tr1TETT ter a tcTT) 208 204 2 04 179 A d *#Ng' fe ' Perfomance Characteristics Ioading Technique 3 region 3 region 3 region 3 region Fuel Discharge Burnup, 76tD'MPU Average First Cycle 8,260 14,000 12,000 12,000 Equilibrium Core Average 28,200 27,000 27,000 21,800 Feed Enrichments, w/o U-235 No. 1 Region 1 2.24 2.28 2.23 2 35 Regior. 2 2.47 2.43 2 38 2.50 7

p-*

Region 3 Equilibrium 2 77 2 73 2.68 2.80 3 09 ---

2,$! 3 05 O Control Characteristics Effective Multiplication (beginning of life No. 1 No. 2 Cold, No Power, Clean 1 312 1.255 1.275 1.275 1.275 Hot, No Power, Clean 1.258 1.201 1.225 1.225 1.225 Hot, Reted Power, Xe and Sm Equ111briua 1.167 l 119 1.170 1.170 1.205 O Control Rod cluster Assemblies a Material Number of ROC Assemblies 5% Cd-15% In-80% Ag 69

$$ Cd-15% In-80% Ag 41 5% Cd-15% In-80% Ag 5% Cd-15% In-80f Ag 53 30 Number of Absorber has per RCC Assembly 16 20 20 16 O Total b d Worth ,% 10.0 70 70 75 C"3 Baron Concentrations Oc 3 To shut reactor down with ruda inserted (clean), cold /hotppa 4 , 1290/1150 2200/250c 3400/3500 2200/2350

' To control at power with no rods inserted N (clean /equilibriumxenogandsamarium) 1950/1430 Boron Worth (bot), % 0 fppm 1/100 1/130 1/150 1/150 Boron Worth (cold), /ppia 1/75 1/100 1/22 0 1/120

% Kinetic Characteristics tg3 Moderator Temperature Coefficient, Ak /F ,1,oxio-4to-l.7x10-b +1.ox10-4to-3 0xlO' +1.0x10 to-3 0x10 4 h Moderator Pressure Coefficient, / psi

-1.0x10 to+2.oxlod -1.0410 tot 3.onlo d 1.oxio 6ge,3,ox194

+1.ox10 4to.3.oxio-4 68

-1.0x10-6go,3,oglo 4 Moderator Void Coefficient, /$ void b *1.Qx10 to-3 0xlO-3 +0 5xlo-3to-2.oxlo-3 *l.0x10~3to-3 0x10'3 +1.oxlo'3to-3 0x10-3 g Doppler Coefficient, /F 1,1xio-5to-l.7x10-5 1.oxio-5 to-2.0mio-5 -1.0x10'Ito-2.0x10-5 ,1,oxto4to.2.oxlo-5 0

of $ $L . ! : .f s L 0l Yf *.. $) u Yf /* l ,

[. H 'ls !t 9 ^ .Y N c'[vlc'b b he b 'l Hi Of N /0

/11 Y f r r f d. Vs't1*&l b :a.KM,Is./u . .% :.) luc. :/ & xs% a mnd id,,,e id:9 m.dtoJ aut, th ut,;a &t,c h i> 3. .U .

n ]

m ^

\

\ 4 U N Table 1-2 (Cont'd )

(Fer Unit Basis Unless Noted)

Item ***

l n r Turkey Point No. 3 or b Indian Point No. 2 P roc kC

% Principal Design Parameters of the Reactor Coolant System System Heat Outpu', awt 2,468 2,097 2,758 1,300 System Heat Output, Btu /hr T */ A) - Arbe9x10k 7,156x106 9,412x106 g,g37,396 Operating Pressure, psig y

2,185 2,2_35 2,235 2,235 Reactor Inlet Tempe mture, F 555 546.5 543 Reactor Outlet Temperature, F 556 603 600.6 596.0 605.4 haber of Icops 2 3 4 2 Design Pressure, psig 2,500 2,485 2.485 2,A85 Design Temperature, F 650 650 650 Hydrostatic Test Pressure (cold), psig 3,12 5 650 3, 11 0 3.110 3,110 Coolant Volume, including pressurizer, ft3 ---11,600 p# y ** ' 9,800 Total Reactor Flow, gpa 352,000 266,400 12,209 355,800 6 38

,000 5 Beactor Coolant System Code Bequirements H Reactor Vessel ASE III, Class A ASME III, Class A ASME III, Class A ASME III, Class A e

Steam Generator Tube Side j ASME III, Class A ASE III, Class A ASME III, Class A Shell Side ASME III, Class A ASME III, Class A ACME III, Class C ASE III, Class C ACME III, Class C '

Pressurizer ASME III, Class A ASE III, Class A ASME III, Class A ASME III, Class A Pressurizer Relief Tank ASME III, Class C ASME III, Class C ASNB III, Class C Pressurizer Safety Valves ASE III, Class C ASME III ASME III ASME III ASME III Reactor Coolant Piping ASA B31.1 ASA B31.1 ASA B31.1 ASA B31.1 O 6 Principal Design Isrameters of the Beactor Vessel O Material SA-302, Grade B. SA-3T , Grade B. CA-300, Grade B.

O Clad with Type 3016 Clad with Type 304 Clad with Type 304 SA-3T, Grade B.

Clad with sustenttic g austenitic w. austenitic SS. austenttic SS. SS.

Design Pressure, psig 2,500 2,485 2,485 Design Temperature, F 2,485 650 650 '

650 650 Operating Pressure, psig 2,185 2,235 2,235 2,235 Inside Diameter of Shell, in. 171 155.5 Outside Diameter Across Nntles, in. 173 IN 249 240/235-3/8 245 220 Overall Height of Vessel and Closure Q

tx3 Head, ft-in.

ginimum 01ad Thickness, in.

41-8 5/8 41-0 42-4 39-0 1/8 $/32 5/E 5/E N3 8

su a

O O

D c.t.

v D

~  %

\ G^-;

v y  %

Table 1-2 (Cont'd)

(ivt Unit Basis Unless Noted)

Item OC0h*'jcycar

, gtation Turkey Point No. 3 or b Indian Point No. 2 Brookwaod 7 Principal Dealgn Parameters of the Steam Generators Namber of Units - 3 b 2 Type Vertical once-through Vertical, U-tube with inte- Vertical, U-tube with Vertical, U-tute with g y ntegral super- gral moisture separator.

integral an21sture sep* integral maisture sepa.

Tute hterial Inconel Inconel Inconel Inconel Shell Material Carbon steel carbon steel Carbon steel Carbon steel Tube Side Design Pressure, peig P,500 2,25 2,485 2,485 Tube Side Design Temperature, F 650 650 650 650 Tube Side Design Flow, Ib/hr 65.66x16 33 53x10 6 3g,o5,1o6 38.53x106 Shell Side Design Pressure, psig 1,050 1,085 1,085 1,085 Shell Side Design Temperature, F M F 4' 600 600 600 Operating Pressure, Tube Side, Nvminal, psig 2,185 2,235 2,235 2,235 Operating Pressure, Shell Side, Maximan, poig 910 1,005 1,005 1,005 Mnxisman m isture at Outlet at Full Ioad, $ 35 F superheat 1/4 1/b 1/4 Hydn, static Test Pressure (tube side-cold)IA 3,12 5 3,110 3.110 3,110 P S 8 Principal Design Parameters of the 7

~

heactor Coolant Nmps N Number of Unite 4 3 4 2 Type Vertical, single stage Vertical, single stage. Vertical, single str.ge. Vertical, single stase, andial flow with bottom boial flow with botton Radial flow with bottom suction and borizontal suction and horizontal suction and horisontal discharge, discharge. discharge.

Design Pressure, psig 2,500 2,L85 . 2,485 2,k85 Design Temperature, F 650 650 650 650 C Operating Pressure, Naminal, psig 2,185 2,235 2,235 2,235 Suction Temperature, F 555 5A6.5 M3 557 Design Capacity, gym 88,000 88,800 89,700 90,000 O Design Total Developed Head, ft 370 2% 272 252 O Hydrostatic Test Pressure (cold), paig 3,12 5 3,110 3.110 3.110 Q mtor Type A-C Induction, A-C Induction, A-C Induction, A-C Induction, single speed single speed single speed single speed Motor hting, hp - 4,000 g et.*. ' 5,500 6,000 5,500 9 Principal Design Isrameters of the Eeactor Coolant F1 ping Carbon steel clad h

w Sterial wtles(IQ,in.

with SS 36 Austenitic SS 29 Austenitic SS 29 Auster.itic SS 29 Coldleg(ID),in. 28 271/2 27 1/2 27 1/2 Between Pump and Steam Oenerator (ID), in. 28 31 31 31 t^

I N

^

O 8

e.-

v

. _ _ _ _ . . fl ._ _ _ _ _ . _ _ _ _ _ _ . _ . _ _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

f')

^ lh ( )

\

Tatle 1-2 (Cont'd)

(Per Unit Basis Unless Nated)

Item Oconee Nuclear Station Unit 1 or 2 Turkey Point M. 3 or 4 Ir lian Point E. 2 Brookwood 10 Reactor Building System Parameters Type Steel-lined, prestressed, Steel-lined, prestressed, Steel-lined, reinforced Steel-lined, reinforced post-tensioned concrete, post-tensiosed concrete, concrete, vertical cylin- concrete, vertical cylin-vertical cylinder with vertical cylinder with der with flat botton and der with flat bottom and flat bottca and shallow flat tattom and shallow bensspherical dome. hemispherical o m .

domed roof, domed roof.

Design Barameters Inside Diameter, ft 116 116 135 105 Height, ft FreeVolume, ft3 af 2 C E 177 2n 146 2,O'AM /,944', Jt'd 1 550,000 2,610,000 972,000 Reference Incident Pressure, peig -ff .f'/ Sd 47 N Reference lacident Energy (E1 ), Btu A,000,000 ]Q ?g & 4 272,000,000 305,290,000 156,030,000 Energy Required to Produce Incident Pressure (E2 ), Btu 133, " ,C'^ *E//. f d ,C N 300,000,000 3k9,880,000 168.300,000 Ratio: E3/E2 'O*900' **-bt# 0 4'? 7 0.907 0.673 0.W8 Batio: (E2 - EI )/El + 118 Ay (A //N 0.20d 0 16 OO Concrete Thickness, ft M

a Vertical Wall w H/a 31// 31/2 51/2 31/2 h Rec.ctor Building Ieak Prevention 4 3g 3 4 1/2 2 1/2 Leak-tignt penetrations Leak-tiEht penetrations

• and Mitigation near we da wi and continuous steel and continuous steel liner. Automatic iso. Ainer. Autmatic iso- weld channals and access channels capable of leak lation where required. lation where required.

openings. Isolation valve test. Automatic isola-Exhauat from penetratim seal water system auto. tion valves in piping, veoms to station vent. natically isolates piping, where required.

O leak rate monitorirg of Q , containment and pres.

,surized areas.

g Gaseous Effluent laarge Discharge vent above top Thzuuc ,h particulate Vent discharge from top Vent discharge from top C of Beactor Building filters and monitors. of containment (~150' of containatnt facade

(- 200 ft above grade) tart of the main exhaust above grade).

O b l15' aoove grade).

system.

W 11 Engineered Safeguards Safety Injection System g No. of High Head Pumps y 3 per Unit and 2 shared. 3 D:= No. of Iow Head Pumps 3 3 tIf '* F per Unit and 2 shared. 2 2 2 Reactor Building Emergency Coolers h No. of Units Air Flow Cap'y. each, at accident 3 3 5  %

y to condition, cfm 54,000 80,000 65,000 38,000

^

O O

3 c+

v

_ . _ _ _ . _ _ .__A._ _ _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

s ]

J v Table I-2 (Cont'd)

(Per Unit Basis Unlesa Noted)

Item Oconee Nuclear Station Unit I or 2 Turkey Point No. 3 or 4 indian Point No. 2 B rook wooit 11 Engineered Safeguerds (Cont'd)

Core Floodtag Tanks 2 3 Postaccident Filters None inside Reactor No. of Units Building. Leakage None 5 4 Air Flow Capacity each, at from penetrations is accident condition, cts collected, filtered N.A. 65,000 65,000 ra Type end discharged through None Roughing / absolute / Roughing / absolute /

8 Filtration Reduction Rate agR/Ve station vent. charcoal charc oal

$[ (ng - 0.9 per pass), hr-l 6.75 (5 units) 8.5 (4 units)

Reactor Building Spray N No. of Pumps 2 2 2 2

'ff

<g Emergency Power Generator Units , No. 2 2 for both units 3 2 g7 Type On-site hyJroelectrit Diesel Diesel Diesel S

c6 C: ) EngineereJ Safeguards Operable All engineered safe.

C::) rrom Emergency rower source guards . uipment is 1 High head Safety (2 of 3 diesels) (1 of 2 diesels) f* (minimum) capable of being lajection (SI) pump 1 High head $1 pump I High head SI pump pa g ;) operated from on-site  ! Low head S1 pump 1 Low head SI pump 1 Law head $1 pump emergency power. 3 containment air 4 Containment air re- 2 Containment air re-CN ([2) recirculation units circulation us.its circulation units

[2! ((() 1 Containment spray pump 1 Containment spray (or I containment spray

(_7) i Service water pump pump pump) 1 Service water pump 1 Service water pump e

cn t*

ts!

s PJ

^

O O

D t1 v

0.

1.4 PRINCIPAL DESIGN CRITERIA The Oconee Nuclear Station is designed to meet the 27 General Design Criteria for Nuclear Fower Plant Construction Permits (l) proposea by the Atomic Energy Cocanission. The principal safety features that meet each criterion are suranarized herein. In the discussion of each criterion, reference is made to sections of this report where more detailed information is presented.

1.4.1 CRITERION 1 Those features of reactor facilities which are essential to the prevention of accidents or to the mitigation of their consequences must be designed , fe,bricated, and erected to:

(a) Quality standards that reflect the importance of the safety function to be performed. It should be recognized, in this respect, that design codes commonly used for nonnuclear applications may not be adequate.

(b) Performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces imposed by the most severe

earthquakes, flooding conditions, winds, ice, and other natural phenomena anticipated at the proposed site.

Answer:

<\

The integrity of systems, structures, and components essential to accident prevention and to mitigation of accident consequences has been considered in i

the design evaluations. These systems, structures, and compongats are:

1. Fuel assemblies

2. Reactor coolant system

3. Reaccor instrumentation, controls and protective systems 4 Engineered safeguards systems

5. Radioactive materials handling systems

6 Reactor building

7. Electric power sources 1

(a) guality Standards

, The fuel assemblies are designed to maintain their integrity when sub-jected to the mechanical and thermal stresses resulting from anticipated

! operating conditions during their design life. _The design is based on technology which has proven successful in existing nuclear power plants and is substantiated by test data. The fuel assemblies will be manu-factured to high quality standards and subjected to a series of rigorous

tests during fabrication. (Section 3.2.4.2) 3 (1) The criteria as proposed by the AEC in its press release H-252 of November 22, 1965.

c 1,13 00 00052I o

Components and piping in the reactor coolant system are designed for a pressure of 2,500 psig at a temperature of 650 F. The nominal operating conditions of 2,185 psig and 579 F allow an adequate margin for normal load changes and operating transients. The reactor coolant system is designed to meet applicable portions of the following principal codes:

(Section 4.1.5)

Piping and Valves - ASA B31.1 (Piping Code) including nuclear cases.

Pump Casing - ASME Boiler and Pressure Vessel Code,Section III.

Steam Generators - ASMC Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

Pressurizer - ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

Reactor Vessel - ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

Quality control, inspection, and system testing will insure integrity of the reactor coolant system. (Section 4)

The fast neutron exposure (3 x 1019 nyt) of the reactor vessel results in a maximum NDTT of not more than 260 F at the end of station service life. Operating procedures for the station will be compatible with these temperature limitations. (Section 4.1.4)

The instrumentation, reactor control system, and protective systems will be fabricated using high quality components and workmanship. All com-ponents will be selected on the basis of reliability, durability under -

service conditions, and proven application. In the absence of applicable industry standards and codes regulating quality in this type of equip-ment, quality standards are established based on extensive e>perience, study and testing. (Section 7)

(,ogg fwoiNS GNM Engineered safeguards , ie[the emergency injection systems , the Reactor Building spray system, the Reactor Building isolation system, and the Reactor Building emergency cooling units are designed and will be fabri-cated to high quality standards and tested to insure proper operability.

All piping in these systems is designed to the applicable ASA Code for pressure piping. (Section 6)

Radioactive material handling systems are designed and will be fabricated in accordance with the applicable design codes listed in the introduction to Section 9. Pressure vessels are designed to either ASME Boiler and Pressure vessel Code Sections III or VIII, as applicable.

The Reactor Building is designed and will be constructed in accordance with applicable sections of appropriate ACI and ASTM codes and specifi-cations as well as criteria described in Section 5.

Five sources of electric power will supply the electrical load require-ments of either unit. One of these will be an underground cable from the two-unit Keowee Hydro Station located on the site. Four redundant battery backed busses will be provided for vital instrumentation and control. Electrical equipment will be purchased and tested to stringent -

e 1-16 M

requirements for reliability and quality, including appropriate NEMA,

,O ASA and IEEE electrical standards. (Section 8) d (b) Performance Standards All equipment and structures having a safety function >>ill be designed, constmeted, operated and maintained without loss of capability to protect the public under all environmental conditions anticipated at the site.

l.h.2 CRITERION 2 Provisions must be included to limit the extent and the ccnse-quences of credible chemical reactions that could cause or materially augment the release of significant amounts of fission products from the facility.

Answer:

Two emergency injection coolant systems and core flooding tanks are provided to prevent core melting for the complete range of postulated reactor coolant system rupture sizes up to the maximum size of a 36 in. ID pipe. In the process of cooling the core, the amount of metal-water reaction is limited to an insignificant amount. (Section lb)

A high pressure injection system with a capacity of 1500 gpm will pump water to the reactor coolant system at all pressures up to full design pressure.

O This system is primarily effective early in the accident while the reactor coolant system pressure is above 100 psig. (Section 6)

A low pressure injection system provides 9,000 gpm to recover the core when it is partially or totally uncovered, and the reactor coolant pressure has dropped below 100 psig. (Section 6) 1.h.3 CRITERION 3 l Protection must be provided against possibilities for damage of the safeguarding features of the facility by missiles gen-erated through equipment failures inside the containment.

Answer:

Protective walls and slabs, local missile shielding, or restraining devices will be provided to protect the Reactor Building liner plate and engineered l safeguard systems within the Reactor Building against danage from missiles renerated by equipment failures. The con rete enclosing the reactor coolant system serves as radiation shielding and an effective barrier against I missiles. Local missile barriers will be provided for control red drive j mechanisms. (Sections h and 5.1.2.7.1)

For those parts of the safeguards systems susceptible to missile damage, redundant equipment is provided to assure required operation. (Section 6)

O 1 00 0005F 11 1-17 (Revised 4-1-67)

1.4.4 CRITERION 4

~

The reactor must be designed to accommodate, without fuel failure or primary system damage, deviations from steady state norm that might be occasioned by abnormal yet enticipated transient events such as tripping of the turbine-generator and loss of power to the reactor recirculation system pumps.

Answer:

The reactor is designed with a margin above normal operating conditions to accommodate anticipated abnormal deviations from the steady state operation.

This margin allows for deviations of temperature, pressure., flow, reactor power and reactor-turbine power mismatch. The reactor is operated at a constant average coolant temperature above 15 per cent power and has a negative power coefficient to dampen the effects of power transients. The reactor control system will maintain the reactor operating parameters within preset limits and the reactor protective system will shut down the reactor if normal operating limits are exceeded by preset amounts. (Sections 7.1 and 14)

The reactor plant is shut down automatically by the reactor protective system if a complete loss of electrical power occurs. Upon loss of electrical lo-d, a reactor power reduction occurs, and the reactor continues to generate sta-tion power needs at reduced load. The resultant reactor coolant system tem-perature and volume increases for both of the above are '-!A within design limits by relieving steam through the bypass to the condenser and/or secondary system relief valves to the atmosphere, thereby preventing excessive reactor coolant system pressures. Accordingly, these transients will not produce fuel or reactor coolant system damage. (Sections 7.1 and 14.1.2.8)

The reactor coolant pumps are provided with sufficient inertia to maintain adequate flow to prevent fuel damage if power to all pumps is lost. The criterion for core protection following loss-of-coolant flow is to maintain a Departure from Nucleate Boiling Ratio (DNBR) equal to or greater than that at the design overpower level for initial power conditions up to and including the maximum operating power level of 107.5 per cent awer. Natural circula-tion coolant flow will provide adequate core cooling after the pump energy has been dissipated. (Section 14.1.2.6 and Figure 9-7) 1.4.5 CRITERION 5 The reactor must be designed so that power or process variable l oscillations or transients that could cause fuel failure or primary system damage are not possible or can be readily suppressed.

Answer:  !

l The ability of the reactor control and protective system to control the oscillations resulting from variation of coolant temperature within the con-trol system dead band and from spatial xenon oscillations has been analyzed.

Variations in average coolant temperature provide negative feedback and en-hance reactor stability during that portion of core life in which the modera-tor temperature coefficient is negative. When the coefficient is positice, l

1

[. i i k) U i : '

l-18

(^

\

rod motion will compensate for the positive feedback. The maximum power change rate resulting from temperature oscillations within the control system dead band has been calculated to be less than 1 per cent / minute. Since the station has been designed to follow ramp load changes of 10 per cent per minute, this is well within the capability of the control system.

Control flexibility with respect to xenon transients is provided by the combination of control rods, incore instrumentation and out-of-core instru-centation. Within control rod limits, transient xenon related to load enanges is controlled by the automatic control system. Axial, radial, or azimuthal neutron flux changes will be detected by both out-of-core and in-core instrumentation. Individual or groups of control rods can be positioned to suppress and/or correct flux changes. (Section 3.2.2.2.3) 1.4.6 CRITERION 6 Clad fuel must be designed to accommodate throughout its design lifetime all normal and abnormal modes of anticipated reactor aperation, including the design overpower condition, without experiencing significant cladding failures. Unclad or ventet fuels must be designed with the similar objective of providing control ove fission products. For unclad and vented solid fuels, normal and abnormal modes of anticipated reactor operation must be achieved without exceeding design release rates of fission products from the fuel over core O lifetime, b

Answer:

Fuel clad integrity is insured under til normal and abnormal modes of antici-pated operation by avoiding clad overstressing and overheating. The evalua-tion of clad stresses includes the effects of internal and external pressures, temperature gradients and changes, clad-fuel interactions, vibrations, and earthquake effects. The free-standing clad design prevents collapse at the end volume region of the fuel rod and provides sufficient radial and end void volume to accommodate clad-fuel interactions and internal gas pressures.

(Section 3.2.4.2)

Clad overheating is prevented by satisfying the following core thermal and hydraulic criteria: (Section 3.2.3.1.1)

(a) At the design overpower no fuel melting will occur.

(b) A 99 per cent confidence exists that at least 99.5 per cent of the fuel rods in the core will be in no jeopardy of experiencing a DNB during continuous operation at the design overpower of 114 per cent.

1.4.7 CRITERION 7 The maximum reactivity worth of control rods or elements and ,

the rates with which reactivity can be inserted must be held to values such that no single credible mechanical or electrical control system malfunction could cause a reactivity transient I

'M capable of damaging the primary system or causing significant fuel failure. <

l-19 000005q l

.I

[

Answer:

Reactivity control will be accomplished by movement of control rods and by changes in soluble poison (boron) concentration in the reactor coolant. Each control rod consists of a cluster of 16 poison pins. The rod drive mechanism and controls will have an inherent feature to limit overspeed in the event of malfunctions.

Approach to criticality and low power operation will be by means of manual rod withdrawal. The remaining rods (or rod groups) will be interlocked to permit withdrawal on automatic control only after the rod groups used for approach to criticality and low power operation have been fully withdrawn.

Rods used for automatic control will be arranged in four groups and inter-locked to prevent simultaneous withdrawal of more than two groups. That is, simultaneous withdrawal of two automatic groups will be permitted over approx-imately the first 25 per cent of the second rod group stroke and the last 25 per cent of the first rod group stroke.

The maximum reactivity insertion rate associated with simultaneous withdrawal of a regulating twelve rod group is 5.8 x 10-5 o.k/k/sec. Assuming a single electrical failure occurs that invalidates the interlock and permits the 25 control rods on automatic control to be withdrawn simultaneously, a maximum reactivity insertion rate of 1.9 x 10-4 tk/k/sec. could result. Reactivity transients of this magnitude have been analyzed, and the resultant power transients will not produce reactor coolant system or fuel failure.

(Section 14.1.2.3)

A reduction in the reactor coolant soluble poison concentration will require operator initiation, and will be prohibited by interlocks until the control rods are in an acceptable pattern for dilution. A second safety feature will physically limit the maximum rate at which dilution water can be added to the system. A third safety measure will consist of a relay which will limit the total time of dilution. The maximum reactivity insertion rates from moderator dilution will be 7.0 x 10-6 ok/k/sec. These rates are not sufficient to produce damage to either the fuel or reactor coolant system.

1.4.8 CRITERION 8 Reactivity shutdown capability must be provided to make and hold the core suberitical from any credible operating con-dition with any one control element at its position of highest reactivity.

Answer:

The reactor is designed with the capability of providing a shutdown margin of at least 1 per cent ak/k with the single most reactive control rod fully with-drawn at any point in core life with the reactor at a hot zero power condi-tion. The minimum hot shutdown margin of 1.1 per cent Ak/k occurs at the end of the first cycle of Unit 1. The margin for Unit 2 is greater.

Reactor suberitical margin is maintained during cooldown by changes in soluble poison concentration. The rate of reactivity compensation from boron addition c

l-20

/R \ is greater than the reactivity change associated with the maximum allowable U reactor cooldown rate of 100 F per hour. Thus, subcriticality is assured during cooldown with tne most reactive control rod totally unavailable.

(Section 3.2.2.1) 1.4.9 CRITERION 9 Backup reactivity shutdown capability must be provided that is independent of normal reactivity control provisions. This system must have the capability to shut down the reactor from any operating condition.

Answer:

Soluble poison addition will provide an independent backup to the control rods for reactivity shutdown. Poison addition will be accom lished using the high l pressure injection and purification system. There are high pressure injection pumps to insure flow availability under all credible operating con-dicions. These pumps take suction from th. borated water storage tank which contains water with 2,270 ppm boron or from the letdown storage tank. In the 1 latter case, a solution containing 8,750 ppm boron is supplied to the letdown )

storage tank from a mixing tank. Two transfer pumps are provided. l (Section 9.1)  !

i l

The high pressure injection pumps and the two sources of concentrated boron solution insure the capability of being able to shut down the reactor without D any control rods from any operating condition. The following table demon-strates the capability of shutdown with control rods for two modes of high pressure injection and purification system operation. ,

l l

Soluble Poison Shutdown Capability l Feed Negative Reactivity Time to Shut Down from 1007.

Concen- Feed Insertion Rate, Full Power to Hot Zero Power tration, Flow Rate, 7. Ak/k Minute Condition (1), Minutes PPM Boron GPM BOL EOL BOL EOL l l

8,750 20 0.0179 0.0217 56 83 )

4,121 70(2) 0.0224 0.0353 45 51 3,196 140(3) 0.0278 0.0543 36 33

( ) Reactivity balance on Doppler and Moderator equal to 1.0 per cent Ak/k for BOL and 1.8 per cent Ak/k for EOL.

( ) Makeup to letdown storage tank at 20 gpm of 8750 ppm boron from boric acid mix tank plus 50 8Pm at 2270 ppm boron from storage tanks.

( ) Makeup to letdown storage tank at 20 gpm of 8750 ppm boron from boric acid mix tank plus 120 gpm at 2270 ppm boron from storage tanks.

000006$ 1 1-21

l.4.10 CRITERION 10 s

Heat removal systems must be provided which are capable of accommodating core decay heat under all anticipated abnormal and credible accident conditions , such as isolation from the main condenser and complete or partial loss of primary coolant from the reactor.

Answer:

Reactor decay heat will be removed through the steam generators until the reactor coolant system is cooled to 250 F. Steam generated by decay heat will supply the steam-driven feedwater pump turbine and can also be vented to atmosphere and/or bypassed to the condenser. The steam generators are supplied feedwater from either the main steam-driven feedwater pumps, which can be operated at a reduced flow rate for decay heat removal, or from a steam-driven emergency feed pump sized at 5 per cent of full feedwater flow.

The main feedwater pumps supply water contained in the feedwater train and the upper surge tank (condensate storage) to the steam generators. The emergency feed pump takes suction from the condenser hotwell and the upper surge tank. These sources provide approximately 1.5 million pounds of water storage which is sufficient for decay heat removal for about two days after reactor shutdown with the condenser isolated. The condenser is normally available so that water inventory is not depleted. (Section 10)

Without use of reactor coolant pumps, decay heat will be removed by natural circulation through the reactor coolant system. (Section 14.1.2.8)

Under conditions of complete or partial loss-of-coolant from the reactor, ,

decay heat will be removed from the core by coolant supplied by the emer- c,gs fe,<mvg gency injection coolant systems. The source of injection water will be the rug 5A uo TA borated water storage tank. When this source is exhausted, the low pressure injection pumps will take suction from the Reactor Building sump. The return flow is. cooled and pumped to the reactor vessel to continue core cooling.

This system contains redundancy of equipment to insure availability of flow when required. If complete loss of external electric power occurs, on-site sources supply sufficient electric power for all engineered safeguards and cooling water systems. (Section 14.2.2.3) 1.4.11 CRITERION 11 Components of the primary coolant and containment systems must be designed and operated so that no substantial pressurc or thermal stress will be imposed on the structural materials unless the temperatures are well above the nil-ductility temperatures. For ferritic materials of the coolant envelope and the containment, minimum temperatures are NDT + 60*F and NDT + 30*F, respectively.

Answer:

The reactor vessel plate material opposite the core is purchased to a speci-4 ,' ' . t r - I b l-22 hh

9

/ fied NDTT of 10 F or less, and is tested to verify confcrmity to specified

(,-)/ requirements. (Section 4.2.5)

The end of station life NDIT value of the reactor vessel opposite the core will be not more than 260 F. Station operating procedures will be established to limit the operating pressure to 20 per cent of the design pressure when the reactor coolant system temperature is below NDTT plus 60 F throughout station life. Surveillance specimens of the reactor vessel shell section material will be installed between the core and inside wall of the vessel shell to monitor the NDTT of the vessel material during operating lifetime.

4 (Section 4.1.4)

The reactor vessel material is protected from excessive radiation damage by coolant water annuli between the core and the reactor vessel. The thickness of these annuli limits the total fast flux greater than 1 mev incident on the reactor vessel wall to an nyt value of 3 x 1019 in 40 years at an 80 per cent station capacity factor. The thermal shield contributes to a further reduc-tion in vessel material radiation damage. (Section 4.1.4)

The Reactor Building steel liner plate (ASTM A-36) will be maintained at temperatures above the 30 F nil-ductility temperature plus 30 F, or above 60 F. The average normal operating temperature for the atmosphere inside the Reactor Building will be in excess of 100 F. The liner plate is completely enclosed by the thick concrete walls, slab and roof of the Reactor Building, and will thus not be subject to sudden variations due to changes in external l temperatures. In addition, the bottom liner plate is protected by a minimum

\

thickness of 12 in. of cover concrete. Nil-ductility is not a consideration at the higher temperatures associated with accidental conditions.

1.4.12 CRITERION 12 Capability for control rod insertion under abnormal conditions must be provided.

Answer:

Control rods will provide the normal means for changing reactivity to shut down to a hot subcritical condition. They may be inserted independently of the normal reactor control system by the reactor protective system or by manual means. Both modes of insertion override reactor control system sig-nals by interrupting power to the rod drives. Without power the control rods insert into the core by gravity. Soluble poison is added to maintain sub-criticality from a hot to a cold zero power condition.

The principal safety criteria for the control rod drive assemblies are:

(a) No single failure in the drive system shall result in the loss of safety function.

(b) Trip action shall not require power, and no single failure or chain of failures shall prevent trip action to more than one mechanism.

s (c) The trip command shall override all other commands. Trip action shall be nonreversible.

}

00 0006S l-23

The reactor vessel, reactor vessel supports, reactor vessel internals, fuel assemblies, control rods, and the control rod drive mechanisms are all de-signed to resist, without loss of function. the effects of seismic loadings established by the seismological analysis of the site, The control rod is never withdrawn completely from its guide structure. The guide structure is oriented with respect to the fuel tssembly by a common grid structure which maintains full stroke control rod guidance into the fuel assembly. The drive line is designed and will be tested to be fully operable under conditions of the maximum misalignment specified. (Section 3.3.3.4.1) 1.4.13 CRITERION 13 The reactor facility must be provided with a control room from which all actions can be controlled or monitored as necessary to maintain safe op_erational status of the plant at all times. The control room must be provided with adequate protection to permit occupancy under the conditions described in Criterion 17 below, and with the means to shut down the plant and maintain it in a safe condition if such accident were to be experienced.

Answer:

Each reactor unit will be controlled from a separate control panel located in a single control room. The control room is designed to permit continuous occupancy following an accident. (Section 7.4.5)

All controls and instrumentation required to monitor and operate the reactors I and electric power generating equipment will be located within the control room. This includes indication of power level, process variables such as temperatures, pressures, and flows, valve positions, and control rod positions.

All engineered safeguards equipment will be controlled and monitored from the control room. The status of all dynamic equipment (pumps , valves , etc) as well as pertinent pressures, temperatures and flows will be displayed. The station radiation monitoring system will have instrumentation readouts dis-played in the control room.

During MHA conditions the concrete Reactor Building and control room walls and roof provide adequate protection against direct radiation to control room personnel. The direct dose to control room personnel for the first two hours after the accident is less than 1 rem. Control room personnel on eight hour shifts during a 90 day period following the MHA would not receive an inte-grated direct dose in excess of 3 rem, which is approximately equal to the calendar quarter dose permitted in 10 CFR 20. Shielding provided by the Reactor Building permits restricted access to all areas around the site after an accident without excessive direct dose exposures. (Section 11.2)

The control room is provided with indegegdgngeglation and filtration systems to limit exposure to airborne en a n= n a escaping from the Reactor Building. (Section 9.8)

C k I 1-24 P O

i

t 1.4.14 CRITERION 14

' (v)

Means must be included in the control room to show the relative reactivity status of the reactor such as position indication of

mechanical rods or concentrations of chemical poisons.

Answer:

The position of each control rod will be displayed in the control room. The reactivity status of soluble poison will be indicated by the position of the control rods. The soluble poison concentration will be adjusted to be con-sistent with specified rod patterns and control rod group position. Accord-ingly, continuous indication of soluble poison concentration will not be required. The operator will receive results of laboratory analyses of the soluble poison concentration. (Section 7) 1.4.15 CRITERION 15 A reliable reactor protection system must be provided to automatically initiate appropriate action to prevent safety limits from being exceeded. Capability must be provided for testing functional operability of the system and for deter-mining that no component or circuit failure has occurred. For

, instruments and control systems in vital areas where the po-tential consequences of failure require redundancy, the redundant channels must be independent and must be capable

'- of being tested to determine that they remain independent.

4 Sufficient redundancy must be provided that failure or removal from service of a single component or channel will not inhibit necessary safety action when required. These criteria should, where applicable, be satisfied by the ,

instrumentation associated with containment closure and l isolation systems, afterheat removal and core cooling I systems, systems to prevent cold-slug accidents, and other i

vital systems, as well as the reactor nuclear and process )

i safety system.

1 Answer:

The reactor protective system is designed to provide the features specified in this criterion. A minimum of four sensors is provided for each trip variable except start-up rate. Two sensors are provided for start-up rate monitoring. Reactor trip is provided when the following parameters exceed preset values:

(a) Reactor power (b) Reactor outlet temperature (c) Reactor pressure (d) Reactor start-up rate j If a portion or all of an instrumentation channel is removed from service, g the channel assumes a tripped condition. One channel in a tripped condition places the protective system in a half-tripped mode such that a trip of any l s

1 ,

00-00065

one of the remaining channels causes a reactor trip.

Reactor Building isolation and engineered safeguards are initiated from a 3 channel system described in Section 7.

The power supply for each individual channel will be from one of the 4 redun-dant battery-backed vital busses. (Section 8) The components and channels are normally energized and loss of power causes a reactor trip. (Section 7)

Provisions will be included for testing the protective systems and/or com-ponents under administrative control on a periodic basis. Normal testing will include the insertion of a simulated signal to dynamically check response and performance of each channel's components except detectors. Tests of each protective system channel will insure a high confidence level of system operability. (Section 7.1.3.5) 1.4.16 CRITERION 16 The vital instrumentation systems of Criterion 15 must be designed so that no credible combination of circumstances can interfere with the performance of a safety function when it is needed. In particular, the effect of influences common to redundant channels which are intended to be independent must not negate the operability of a safety system. The effects of gross disconnection of the system, loss of energy (electric power, instrument air), and adverse environment (heat from loss of instrument cooling, ext:.eme cold, fire,

)

steam, water, etc.) must cause the system to go into its  ;

safest state (fail-safe) or be demonstrably tolerable on some other basis.

Answer:

Protective systems instrumentation is designed to operate in Reactor Building ambient conditions ranging from 49 F to 140 F without adverr,e effects in accuracy. Reactor Building temperature will be normally ccntrolled in the range of 60 F to 104 F. The protective system instrumentation, exclusive of the neutron detectors in the Reactor Building, will withstand the external pressure and temperature for the duration of a loss-of-coolant accident and still be operable (but subject to several per cent inaccuracy). The out-of-core neutron detectors are designed for continuous operation in a temperature of 175 F and a pressure of 150 psig.

Redundant instrument channels are provided for all reactor protective and engineered safeguard systems. Loss of power to each individual reactor pro-tective channel will trip that individual channel. Loss of all instrument power will trip the reactor protective system thereby releasing the control rods and will activate the engineered safeguards system controls with the exception of the Reactor Building spray valves. (Section 7.1)

Manual reactor trip la designed such that failure of the automatic reactor trip circuitry will not prohibit or negate the manual trip. The same is true with respect to manual operation of the engineered safeguards equipment.

(Section 7.1) 7{ nei 1-26 hh

1.4.17 CRITERION 17 d

The containment structure, including access openings and penetrations , must be designed and fabricated to acconunodate or dissipate without failure the pressures and temperatures associated with the largest credible energy release including the effects of credible metal-water or other chemical reac-tions uninhibited by active quenching systems. If part of the primary coolant system is outside the primary rt. actor containment, appropriate safeguards must be provided for that part if necessary, to protect the health and safety of the public, in case of an accidental rupture in that part of the system. The appropriateness of safeguards such as isolation valves, additional containment, etc. , will depend on environmental and population conditions surrounding the site.

Answer:

The Reactor Bupiding, iLcluding access openings and penetrations, has a design pressure of $s psig at lHFT'F. The greatest transient peak pressure, assoc-iated with a hypothetical rupture of the piping in the reactor coolant system and the effects of a credible metal-water reaction, will not exceed these values.

The Reactor Building and engineered safeguards systems have been evaluated h

V for various combinations of credible energy releases. The analysis accounts for system energy, decay heat, metal-water reactions, and the burning of the resultant hydrogen. The cooling capacity of either Reactor Building cooling system (see Criterion 18) is adequate to prevent over-pressurization of the structure, and to return the Reactor Building to near atmospheric pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The details of this evaluation are discussed in Section 14.2.2.3.5.

The use of injection systems for core flooding will limit the Reactor Building pressure to less than the design pressure. If a metal-water reaction is un-inhibited by the active quenching systems the resultant peak Reactor Building pressure is less than the design pressure.

No lines which contain high temperature, high pressure reactor coolant pene-trate the Reactor Building except the sampling lines. These small sampling lines are normally isolated by two valves in series. Therefore, it is only during a sampling operation that a line failure would require operator action to prevent escape of coolant external to the Reactor Building. This is a procedure that the operator would normally perform.

The high pressure injection and purification system diverts a small amount of reactor coolant outside of the Reactor Building. This high pressure and high temperature coolant is cooled before it leaves the Reactor Building. Lines serving this function contain isolation valves that can be closed to prevent uncontrolled release of reactor coolant in the event a line fails external to the Reactor Building. The letdown coolers are supplied with water from the Reactor Building and component cooling system. Any leakage of reactor coolant V through the letdown coolers will be into this system rather than -to the en-l 00 00067 1-27 l

vironment. The Reactor Building and component cooling system is monitored to detect leakage of reactor coolant.

Leakage of contaminated coolant from engineered safeguards equipment located external to the Reactor Building has been evaluated, and the resultant envir-onmental consequences are well below 10 CFR 100 limits at the site boundary, and have been included in the total accidental dose calculations.

The high pressure injection and low pressure injection systems have redun-dancy of equipment to insure availability of capacity. (Section 6.1)

Some engineered safeguards systems have both a normal and an emergency func-tion, thereby providing nearly continuous testing of operability. For example, one high pressure injection pump is in continuous use for seal in-jection and makeup; the low pressure injection pumps are in use for decay heat removal during each shutdownyand cn; n;mcLus Buildius ovmpvuuui ccclius p=p ic in continueus usm.~

During normal operation the standby and operating units will be rotated into service on a scheduled basis. In cases where separate equipment is used solely for emergency conditions, such as the Reactor Building spray pumps, recirculating lines are provided, and instrumentation is installed to provide means for conducting tests. The equipment is located to facilitate inspec-tion during operation. (Sections 6 and 9)

Electric motors, valves, and damper operators, which must function within the accident conditions, will operate in a steam-air Reactor atmosphere Buildiog,.during at M_ F and $8 psig.

~

1.4.18 CRITERION 18 Provisions must be made for the removal of heat from within the containment structure as necessary to maintain the integrity of the structure under conditions described in Criterion 17 above.

If engineered safeguards are needed to prevent containment vessel failure due to heat released under such conditions, at least two independent systems must be provided, preferably of different principles. Backup equipment (e.g. , water and power systems) to such engineered safeguards must also be redundant.

Answer:

Reactor Building cooling following the loss-of-coolant accident is provided by two independant systems: (1) the Reactor Building spray, and (2) the Recctor Building emergency coolers. The capability of either of these cooling sys-tems, or both at partial capacity, is sufficient to prevent excessive Reactor Building pressure during loss-of-coolant accident conditions.

The Reactor Building spray system supplies 3,000 3pm from the borated water storage tank into the Reactor Building. After the borated water storage tank is emptied, recirculation from the Reactor Building sump begins. This recir-culated water is cooled in heat exchangers by the service water system. The service water system is always in operation, and therefore has continuously ,

c f

i t 1-28

)hh

indicated availability.

Two sets of nozzles, located in the upper portion of the Reactor Building structure, are arranged to provide a uniform spray pattern. Redundancy in both pumping and heat exchanger capacity exists. (Section 6.2)

, To prevent excessive temperature rise following an accident, the Reactor Building emergency cooling system has three cooling units which reject heat to the service water system. Pumps and heat exchangers are redundant to .

insure availability. (Section 6.3)  !

l

! Upon loss of the normal source of electric power, any one of the four re-maining sources will permit operation of all engineered safeguards equipment.

(Section 8.2.3) 1.4.19 CRITERION 19 f

I The maximum integrated leakage from the containment structure l under the conditions described in Criterion 17 above muqt meet  !

the site exposure criteria set forth in 10 CFR 100. The con-tainment structure must be designed so that the containment can  ;

be leak tested at least to design pressure conditions after completion and installation of all penetrations, and the leakage rate measured over a suitable period to verify its conformance with required performance. The plant must be designed for later tests at suitable pressures.

Answer:

The Reactor Building leakage rate will be determined at design pressure after completion and installation of all penetrations. The leak rate test will verify that the maximum integrated leakage does not exceed 0.5 per cent by volume of the contained a!r per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design pressure. (Section 5.1.2.2)

The environmental hazards from the maximum hypothetical accident, assuming the above specified maximum integrated leakage from the Reactor Building, are within the guideline values of 10 CFR 100. (Section 14.2.2.4)

In order to maintain the specified leakage rate of the Reactor Building dur-ing the operating life of the station, a program of leak rate tests at suit-able pressures will be established. Periodic leak rate tests of the entire Reactor Building are contemplated along with more frequent tests of certain elements such as the access lock, equipment hatch, penetrations with re-silient seals and bellows, and isolation valves. (Sections 5.1.2.6.1 and 5.4)

All Reactor Building penetrations, except the main steam lines and permanent

hatches, are enclosed in penetration rooms. The permanent hatches are pro-l vided with double seals. The volume between the seals is piped to the i penetration ventilation room. Following an accident, these rooms are held i at a slightly negative pressure to collect and filter Reactor. Building l penetration leakage, thereby reducing environmental activity levels result-l / ing from penetration leakage. (Section 6.5) i 00 00069 1 1-29 (Revisell 4-1-67)

1.4.20 CRITERION 20 All containment structure penetrations subject to failure such as resilient seals and expansion bellows must be designed and constructed so that leak-tightness can be demonstrated at de-sign pressure at any time throughout operating life of the reactor.

Answer:

All Reactor Building penetrations with resilient seals or expansion bellows will be constructed so that leak tests can be conducted at any time.

(Sections 5.1.2.6.1, 5.4, 6 and 9) 1.4.21 CRITERION 21 Sufficient normal and emergency sources of electrical power must be provided to assure a capability for prompt shutdown and continued maintenance of the reactor facility in a safe condition under all credible circumstances.

Answer:

The design of the electrical system for this two-unit nuclear station will provide five sources of electric power. These are: (1) power to a nuclear unit from its own generator; (2) 230 kv transmission system; (3) power to a nuclear unit from the other nuclear unit af ter Unit 2 is installed; (4) 100 kv transmission system; and (5) two on-site hydro units. (Section 8.2.3)

All of these power sources are large in capacity and, with associated equip-ment, will insure safe reliable functioning of the station under all oper-ating and shutdown conditions.

1.4.22 CRITERION 22 Valves and their associated apparatus that are essential to the containment function must be redundant and so arranged that no credible combination of circumstances can interfere with their necessary functioning. Such redundant valves and associated apparatus must be independent of each other.

Capability must be provided for testing functional operabi-lity of these valves and associated equipment to determine that no failure has occurred and that leakage is within acceptable limits. Redundant valves and auxiliaries must be independent. Containment closure valves must be actuated by instrumentation, control circuits and energy sources which satisfy Criteria 15 and 16 above.

Answer:

The isolation system closes all fluid lines (except those associated with engineered safeguards systems) penetrating the Reactor Building in the event of a loss-of-coolant accident. Reactor Building isolation occurs on a signal

- e 1-30 6

indicated availability, v

Two sets of nozzles, located in the upper portion of the Reactor Building structure,are arranged to provide a uniform spray pattern. Redundancy in both pumping and heat exchanger capacity exists. (Section 6.2)

To prevent excessive temperature rise following an accident, the Reactor Building emergency cooling system has three cooling units which reject heat to the service water system. Pumps and heat exchangers are redundant to insure availability. (Section 6.3)

Upon loss of the normal source of electric power, any one of the four re-maining sources will permit operation of all engineered safeguards equipment.

(Section 8.2.3) 1.4.19 CRITERION 19 The maximum integrated leakage from the containment structure under the conditions described in Criter!.on 17 above must meet i the site exposure criteria set forth in 10 CFR 100. The con-tainment structure must be designed so that the containment can be leak tested at least to design pressure conditions after completion and installation of all penetrations , and the leakage rate measured over a suitable period to verify its conformance with required performance. The plant must be

, designed for later tests at suitable pressures.

Answer:

The Reactor Building leakage rate will be determined at design pressure after completion and installation of all penetrations. The leak rate test will verify that the maximum integrated leakage does not exceed 0.5 per cent by volume of the contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design pressure. (Section 5.1.2.2)

The environmental hazards from the maximum hypothetical accident, assuming the above specified maximum integrated leakage from the Reactor Building, is with-in the guide line values of 10 CFR 100. (Section 14.2.2.4)

In order to maintain the specified leakage rate of the Reactor Building during the operating life of the station, a program of Icak rate tests at suitable pressures will be established. Periodic leak rate tests of the entire Reactor Building are contemplated along with more frequent tests of certain elements such as the access lock, equipment hatch, penetrations with resilient seals and bellows, and isolation valves. (Sections 5.1.2.6.1 and 5.4)

All Reactor Building penetrations, except the main steam lines, and termanent hatches are enclosed in penetration rooms. Following an accident, these rooms

, are held at a slightly negative pressure to collect and filter Reactor Build-1 ing penetration leakage thereby reducing environmental activity levels result-ing from penetration leakage. (Section 6.5) wi J

00 000JQ 1-29 L

1.4.20 CRITERION 20 All containment structure penetrations subject to failure such as resilient seals and expansion bellows must be designed and constructed so that leak-tightness can be demonstrated at de-sign pressure at any time throughout operating life of the reactor.

Answer:

All Reactor Building penetrations with resilient seals or expansion bellows will be constructed so that le:A tests can be conducted at any time.

(Sections 5.1.2.6.1, 5.4, 6 and 9) 1.4.21 CRITERION 21 Sufficient normal and emergency sources of electrical power must be provided to assure a capability for prompt shutdown and continued maintenance of the reactor facility in a safe condition under all credible circumstances.

Answer:

The design of the electrical system for this two-unit nuclear station will provide five sources of slectric power. These are: (1) power to a nuclear unit from its own generator; (2) 230 kv transmission system; (3) power to a nuclear unit from the other nuclear unit after Unit 2 is installed; (4) 100 kv transmission system; and (5) two on-site hydro units. (Section 8.2.3) j All of these power sources are large in capacity and, with associated equip-ment, will insure safe reliable functioning of the station under all oper-ating and shutdown conditions.

1.4.22 CRITERION 22 Valves and their associated apparatus that are essential to the containment function must be redundant and so arranged that no credible combination of circumstances can interfere with their necessary functioning. Such redundant valves and associated apparatus must be independent of each other.

Capability must be provided for testing functional cperabi-lity of these valves and associated equipment to determine that no failure has occurred and that leakage is within acceptable limits. Redundant valves and auxiliaries must be independent. Containment closure valves must be actuated by instrumentation, control circuits and energy sources which satisfy Criteria 15 and 16 above.

Inswer:

The isolation system closes all fluid lines (except those associated with engineered safeguards systems) penetrating the Reactor Building in the event of a loss-of-coolant accident. Reactor Building isolation occurs on a signal ,

~O.

ts. -

1-30 g

of approximately 4 psig or by manual actuation from the control room.

[V The criterion for isolation valve requirements is:

Leakage through all fluid penetrations not serving accident-consequence-limiting systems is to be minimized by a double barrier so that ne single credible failure or malfunction of an active component can re-sult in a loss of isolation or intolerable leakage. The double barriers take the form of closed piping systems both inside and outside the Reactor Building and various arrangements of isolation valves.

(Section 5.2)

Fluid penetrations serving engineered safeguards systems also meet this criterion, but the actuators are manually operated from the control room.

The control circuitry.that initiates Reactor Building isslation is part of the engineerad safeguards protective system and is designed to meet Criteria 15 and 16. (Section 7.1.3.2)

Isolation valves and valves which control other engineered safeguards equip-ment have test provisions, and periodic manual application of test signals is used to verify functional operability.

1.4.23 CRITERION 23 In determining the suitability of a facility for a proposed v site the acceptance of the inherent and engineered safety afforded by the systems, materials and components, and the associated engineered safeguards built into the facility, will depend on their demonstrated performance capability and reliability and the extent to which the operability of such systems, materials, components, and engineered safeguards can be tested and inspected during the life of the plant.

Answer:

l All engineered safeguards systems are designed so that a single failure of an  !

active component will not prevent operation of that system or reduce the I capacity below that required to maintain a safe condition. ' Two independent J Reactor Building cooling systems, each having full heat removal capacity, are used to prevent overpressurization. (Sections 6.2 and 6.3) l The high pressure injection and low pressure injection systems have redun-dancy of equipment to insure availability of capacity. (Section 6.1) l Some engineered safeguards systems have both a normal and an emergency function, thereby providing nearly continuous testing of operability. During normal operation, the standby and operating units will be rotated into service on a scheduled basis. The answer to Criterion 17 (Section 1.4.17) gives more detail regarding redundancy, testing, and normal and emergency operation of engineered safeguards.

p!

t

'" Engineered safeguards equipment piping, which is not fully protected against m

1-31 00 000$3

missile damage, utilizes dual lines to preclude loss of the protective func- _

tion as a result of the secondary failure. (Section 6) 1.4.24 CRITERION 24 All fuel storage and waste handling systems must be con-tained if necessary to prevent the accidental release of radioactivity in amounts which could affect the health and safety of the public.

Answer:

The spent fuel storage pool is located within the fuel handling and storage area of the Auxiliary Building. The liquid waste holdup tanks and the gassous waste storage and disposal equipment are located within a separate area of the same building. Both of these areas provide confinement capability in the event of an cccidental release of radioactive materials, and both are venti-lated with discharges to the station vent. Analysis has demonstrated that the accidental release of the maximum activity content of a gaseous waste storage tank will not cause doses in excess of the limits set forth in 10 CFR 100. (Section 11.1.2.5.3)

Radioactive liquid effluent leakage into the service water system will be de-termined by monitors on the cooling water discharge lines. Any accidental leakage from liquid waste storage tanks will be collected in a sump and trans-ferred to other tanks to prevent releases to the environment. (Section 11.1.2.4) i 1.4.25 CRITERION 25 The fuel handling and storage facilities must be designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel under all anticipated normal and abnormal i conditions, and credible accident conditions. Variables upon '

which health and safety of the public depend must be monitored. i l

Answer:

All refueling operations will be carried out with the fuel under borated water to provide cooling and shielding for the fuel assemblies. Visual control of all fuel handling operations will exist at all times except during fuel trans-fer from the Reactor Building. Spent fuel is transferred under water through a spent fuel transfer tube to the spent fuel storage pool. Storage space is provided in the pool for two full cores and the spent fuel shipping cask.

Additional underwater storage for large internal components is provided inside l the Reactor Building refueling canal. l To avoid accidental draining of the spent fuel storage pool, there are no penetrations that would permit the pool to be drained below a safe level. The fuel transfer tubes between the spent fuel storage pool and the refueling canal are provided with gate valves and gasketed closure plates to prevent leakage.

A % i.

l

! 3 h't !10 t' i '

l-32 g 1

(\ Water depth in the spent fuel storage pool provides sufficient shielding for j x- normal occupancy by operating personnel. The spent fuel storage pool cooling l system contains provisions to maintain water cleanliness, temperature and water level. A 21 in by 21 in. lattice arrangement is used ,for the spent i fuel storage racks to insure fuel assembly suberiticality.  !

l.4.26 CRITERION 26 Where unfavorable environmental conditions can be expected to require limitations upon the release of operational radioactive effluents to the environment, appropriate hold-up capacity must be provided for retention of gaseous, liquid, or solid effluents.

Answer:

The radioactive waste disposal system will collect, segregate, process, and dispose of radioactive solids, liquids, and gases in such a manner as to insure compliance with 10 CFR 20.

Solid wastes will be processed in a batch manner for off-site disposal.

Liquid and gaseous wastes released to the environment will be monitored and discharged with suitable dilution to assure tolerable activity levels on the

site and at the site boundary. Liquid wastes that have activity levels too high for direct discharge will be held in storage tanks for decay and dilu-tion or for evaporation. Ample holdup storage capacity for liquid waste is sized and will be installed to store all effluent from operation. Wastes' will be sampled to establish release rates consistent with environmental con-l ditions.

The gaseous waste system will store accumulated gas that is released during operation. The contents of the storage tanks will be sampled, and a release rate established consistent with the prevailing environmental conditions.

In-line monitoring will provide a continuous check on the release of activity.

(Section 11) 1.4.27 CRITERION 27 The plant must be provided with systems capable of monitoring the release of radioactivity under accident conditions.

Answer:

Radioactive gaseous effluents which may be released into enclosed areas are

collected by the ventilation systems and discharged to the station vent.

Permanently installed area detectors and the station vent detectors are used to monitor the discharge levels to the environment. In addition, portable monitors are available on site for supplemental surveys, if necessary.

Radioactive liquid effluent leakage into the service water systems will be determined by monitors on the cooling water discharge lines. These monitors t

O are used for normal operational protection as well as for accident conditions.

The effluent from the liquid waste disposal system is sampled prior to dis-e

, 00 00075

charge and the release to the environment is monitored to insure compliance .

with 10 CFR 20.

Leakage through the Reactor Building penetrations following a loss-of-coolant accident is collected in the penetration ventilation rooms and discharged to the station vent through an absolute and charcoal filter assembly. Radiation detectors monitor the residual activity prior to discharge to the stack and environment.

1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS The research and development programs that have been initiated to establish final design, or to demonstrate the capability of the design for future oper-ation at a higher power level, are summarized as follows:

1.5.1 ONCE-THROUGH STEAM GENERATOR TEST The test unit for this program consists of 37 full length tubes fabricated in accordance with the production techniques anticipated for the full size unit.

This program consists of performing steady state and load changing operations to demonstrate the ability of the unit to follow the transients and the inter-action of the control system with the water level, steam pressure and flows.

The latter portion of the program includes tests to determine the natural frequency of the tubes and other parts in the steam generator. This will be accomplished by artificially induced vibrations from an external source, and the tubes will be examined for evidence of tube-to-tube contact and wear at support points.

1.5.2 CONTROL R0D DRIVE LINE TEST The test assembly for this program consists of a full size fuel assembly with its associated control rod, adjacent internals and control rod drive assembly.

The unit is being tested under conditions of temperature, pressure, flow and water chemistry specified for the full size reactor installation. This pro-gram will embrace a prototype phase in which the unit will be subjected to misalignment, varying flow, and temperature. The second phase of this program is one of 14 fe-testing where the unit will be continuously cycled to cover the number of feet of travel and the number of trips anticipated for its life in the reactor. Both phases of thi: program will confirm the operability of the drive line in normal and misaligned conditions, confirm the rod drop times and load carrying characteristics of the netating actuator, indicate vibration, fretting corrosion, and hydraulic lift characteristics of the control rods and fuel assemblies, and finally determine the wear characteristics of all the drive line components.

1.5.3 SELF-POWERED DETECTOR TESTS The test units for this program consist of the self-powered detectors de-scribed in 7.3.3. These units have been tested in the B&W Test Reactor at conditions of temperature and neutron flux antic: pated in a central station reactor. These units are currently being tested in the Big Rock Point Nuclear Power Plant where they are exposed to temperatura, neutron flux, and flow for e

~' \

1-34

(N conditions approximating those in the Oconee Nuclear Station. The results of these programs will provide a detector system with predictable characteristics of performance and longevity under incore conditions.

1.5.4 THERMAL AND HYDRAULIC PROGRAMS B W is conducting a continuous research and development program for heat transfer and fluid flow investigations applicable to the design of Oconee.

Two important aspects of this program are:

(a) Reactor Vessel Flow Distribution and Pressure Drop Tests A 1/6 scale model of the vessel and internals is under test to measure the flow distribution to the core, fluid mixing in the vessel and core, and the distribution of pressure drop within the reactor vessel.

(b) Fuel Assembly Heat Transfer and Fluid Flow Test Critical heat flux data has been obtained on a single channel for both tubular and annular test sections. This work has been extended to obtain data on multiple rod fuel assemblies with both uniform and nonuniform heat fluxes at reactor design conditions of pressure and mass velocities. Additional mixing, flow distribution, and pressure drop data will be taken on models of various reactor flow cells and on a full scale fuel assembly.

1.6 IDENTIFICATION OF CONTRACTORS Duke Power Company will be responsible for the design, purchasing, construc-tion and operation of Oconee, a practice successfully followed for all of the Company's major generating facilities now in service or planned.

The Engineering Department has the responsibility for specification of  !

materials and equipment, design of structures and systems and preparation of  !

installation drawing. Procurement is the responsibility of the Purchasing Department. Ihe Construction Department has the responsibility for all site construction activities. Shop inspections and witness testing are the  :

responsibility of the Engineering Department while field quality control and I testing are the responsibility of the Construction Department. The Steam l Production Department has the responsibility for preoperational testing as I well as operation and maintenance of the station. All other departments of Duke are available as needed to assist in the design, construction or opera-tion of the station.

Duke has contracted with BW to design, manufacture and deliver to the site two complete nuclear steam supply systems and fuel. In addition, BW will supply competent technical and professional supervision of erection, of initial fuel loading, of testing and of initial start-up of the complete nuclear steam supply system with coordination, scheduling and administrative direction by Duke.

O The Bechtel Corporation has been retained by Duke as a general consultant to

~

provide such engineering assistance as needed during the design and construc- l tion of the station. Layout, engineering and design of he R or Buildings  ;

1-35

have been assigned to Bechtel. In addition, they are furnishing assistance in the hydrology studies.

As consultants on seismology and meteorolor , the firm of Dames & Moore has been retained. Mr. Joseph Fischer is in charge of seismology studies. Mr.

F. E. Courtney, Jr., is performing the meteorology studies.

Duke has also retained Mr. William V. Conn from Atlanta, Georgia, for geology studies and the Law Engineering Testing Company for subsurface investigations under the direction of Dr. George F. Sowers.

1.7 CONCLUSION

S The personnel assembled to design, construct and operate Oconee are competent.

It is their combined intention to make this a conservative design and one which can be operated to produca electric power safely and economically.

Toward this end -

(a) The site has been examined and found to be suitable for the two-unit nuclear station. The station at this site is compatible with surrounding population and land uses, present and expected. Site characteristics of meteorology, hydrology, geology and seismology are favorable.

(b) The reactor system chosen is a practical design of prove type, and its expected performance will not require fuel exposures or nrgy-release rates higher than those presently proved achievable usia., materials now available. Its shutdown margin and performance characteristics are com- ,

parable to those used in existing reactors. Before it commences commer-cial operation, the reactor system will be thoroughly tested to confirm the desirable features designed into it, and that it will perform as expected with full safety margins.

(c) The reactor will be installed in an enclosure both modern and conserva-tive in design, which will be able to contain and control all materials, vapors or energies which could conceivably be released as a result of an accident under any coincident condition. Supplementing the enclosure capability will be engineered safeguards which will reduce to a very ednimum the consequences of any accident and insure that the dynmnic conditions existing after sa accident are kept well within safeguards design parameters.

(d) A hydroelectric station having a capacity of 140,000 kw is being con-structed contiguous to the Oconee site. Such an arrangement will provide an abundance of quickly available energy to operate the auxiliary and emergency equipment in the nuclear station through redundant connections.

In addition, the hydro station reservoir will serve as a source of cooling water for the nuclear station without adverse thermal effects on the lake, and this use has already been approved by the applicable Federal and State agencies. Although not required by the nuclear station design, streamflow regulation provided by the hydro station will provide dilution capabilities. Thus , the suitability of the site is enhanced to a considerable extent by the hydro project, and there are no interactions between the nuclear and hydro stations adversely affecting public safety. ,

i f D [i p 1-36 ']P {f

(e) Likewise the operation of two nuclear units in one station has been in-vestigated , and their interactions enhance their individual safety.

Joint use of some station utilities and services provides extra redun-dancy and gives some operating flexibility not otherwise obtainable.

Those systems and components vital to the safe operation of a unit will be provided for each unit, and completion of the second unit will in nowise jeopardize the safe operation of the first unit. Sequential con-struction of the two units realizes cost savings.

(f) The station waste and emergency systems will be designed to release only effluents permitted by the AEC Regulations. Where practicable, liquid and gaseous wastes will ba treated so that the effluents contain a mini-mum of radioactivity and significantly less than that allowed by applic-able regulation.

(g) A training program is planned which will adequately prepare operating personnel so that they will be qualified to test, start-up and operate the nuclear units. Experience gained in the design, construction and operation of the CVTR will be of considerable value, i 1

} In consideration of the above circumstances and plans, it is concluded that the proposed Oconee Nuclear Station can be designed, constructed and operated in a safe manner; that the propoced design will provide adequate protection to the public from any sequence of events resulting in disablement of equip-ment from causes , natural or mechanical; and that Duke Power Company is quali-O fled to design, construct, start, operate and maintain these proposed nuclear generating units in accordance with all applicable laws and regulations add in a manner satisfactory to the Atomic Energy Commission, to the public )

interests and to itself. l l

l l

l-37 l

l 0

l t

I 1

l O

l O _

g '-

i

• l}

't - i !- ,jl

~ _ _ evw---p-.,,, _

O DUKE POWER COMPANY SERVICE AREA VIRGINI A

~

_. > )'_p ,-

TE NN ESSEE __b **

N ijhWi l

$ ," {

~

l^'>'.

^

• :, , }

gy N*t

, s/

g , <-^ ~ ,r l_ _ p$ '

g CHA LOTTE f .!' .hi 1. '.

OCONEE E0*EE. ,

<, , (L; ,

NUCLEAR RryER e r J - '~+s 9 .

, ,. ;. a , j STATION GMW3LLE ' N ';.g .'

0 ANDER90se ~>

' q w ,

- . _ A,_ .

50 0 50 10 0 15 0 200 250 APPROXIMATE SCALE IN MILES outrows OCONEE NUCLEAR STATION FIGURE 1-1 00 000dfL C

s ET& TIN NosTM

\

i O Aun= e es ve-T ---

q m~ux mu2 l H

o .

0

~ ~

o O 0 _

._ver a romc Ma w '

wo -w . _ _ . .

0 O _ _ 0 E rv. % e evow I 6 h h 6 6 l I l uemu !

' i o i 1  ;

c =--- + y y 9- i

-n 6

, l --

i @ @ .

l= c0 S<

[.t, ji- com M 4_ , _4_,, os &.,

===> GO

f'y

,.d5,

o. ;

j'll l g /g ?~ g 'Nh _*  ::M DJ~'[

sem

~ -wassas ei.

a,m }a, Ns 1

c--

(:. w.1?.4:;

---=

o

.Mt4-- dj 1

, . p-- -

LMrT d Nb} i f r- I

1. " k FLoo2 et EL7fde$k

!i!I'9! \al e' y f5/  :

\ k E'f'(r  :

i x =

! pml I M  !

i n!

n.

s

- "s -

i

• mas 4

!w, tp ol: a m cu 2v  :

1

~.

O

\

d F

. _ _ _ .. _ . ~ _ _ _ _ _ _ _ _ . - . _ . - _ _ - -

q r9 L] ii .

i i

] O 1 in , ./w * [ L/J A ,i, fi E LPI&~VJ4 , t Gm^4uSE26 e i

{7 A'I1LLARY J 97II,DI%

f

' A

• Shipping Recalving

__) L) 3 - Decentaeanattaa Area C - Storage D = Stair I - Passenger Elevator F - Freight Elevator G - Tunnel to not behine Shop 5 Cooling heter Ploe str7tNwI I - Purification Dominera11aer 2 - Righ Pressere Injectica Pump o -<h s--@-- de----e-  ! : Po '~;." ;;;*a*

3 - toric Acid Pump i

l eex t

I l

4 9

l

- sa u.3 Ts. A,e.

7 - Reseter Coolent Sample Cboler l 8 - Im Pressure 1sjection and Decay Beat temoval Pump f

p j

-<p-- -G--o- - -G- p a : '<; ;;: 701:',':;" ~

11 - Debosating Deatmere tteer Y$

4

@ A[g r 12 Reactor Coolant 51eed Rolaup Tank 13 - Was t e TV ano f er Pump I,. ,

~

14 - merellaneous Waste Holdup Tank f 1) - A=ax111ery Buildtag Sump Tank O. kTM _

m d-

~6) h 16 - Aust11ery ha11 ding Sump Tank Pump 17 - 6 pent Resin storase Tank 3"[ ] L.J e

- 4 "E" 2---

w J' ,

IS, 3 .- tvaporater Evapo-starWacuum Pump Condenser 20 - Evaporator Condensate feet Tank

, e y - p-. e rg -

,l t, f~

i 21 - waste Evaporator 22 - Evaporator Cameestrate Puur C-Trm S wa m

'.g. g [

23 - Evaporator Condenaste P uy 24 - Evaporator (bedenaste EMaeraliser 5 f- j '%, w l 25 - Beactor Coolant Pep Seal Beters cooler 26 - 1eer Preaeore tsjectice and Decay Heat temoval Cuoters

/

] s' +\ '\ l 't i 27 - Steam Generator Sample Cooler I l 28 - R h:1 at Cooling Pumpe

. -.g - p. [ n..a g l 2 - .eactor.e et- ,141ng C,

..r,e.

e v,-e ].j. ..

s n

p

.ed Pu.y Q %73.g.y),

ni , ,

u s' ,

@]$ " t, lIl,j i i

art '-l a-o j a ~.

n.a I

! ll L. ugp iYV-

- t tr 7 , 1 - i l f' r.2kITd sea i 4 n ' .- . . -_ __

w en yria , d i i-o

>J'so !F Fj

\l Q (Q<

% LJ 6 f llNIl GENERAL ARRANGEMENT FLOOR PIAN e' ~ # /  !,hll u.h k k([k

d s  %

l ele 775+0 N \w h = f. [l h l li t

4'..-

i ,

N ; % ... '

l' l h, ! ,

Nc /

~

[ l. , '

6 el i u' l .. c) !. o um OCONEE NUCLEAR STATION gg"'" ~

FIGURE 1-2 00 000 $ 63 ,

I e.

J

r'

\

{

M STATiou w2T4

! [Q) A m' caf EJM ve%T ~~~

g i L_ ._ -- .---- -

li I

'M

-4> 0 O o O -o bEATE %

l O O Q O - O E S i

I

,, g ,- V s

,p =A,

,- . . . --e -

I '

OT n e1 a D?yse g

? -

CC g, l f - - -

0- oc g/ P I R n . .

-w n>w-a .-

VLbT I t 51.c06L m t-l /

v 0TEuh Q%

CA.E2V

I I

' t i

}

. - . . ~ _ . _ -

Svi >W guRItlaFT SUnDDG a . che= teel necetvies and storese O O o 9-- -e -- - o i

- .tei-C . Passenser Eleestee o . rreisha ries.ter I

• hatch j F Tennel to het Machine Shop

-+

match t o O o 0 -o' C. e-e.nes,oie GeteiMachtse Shop

.-t ie.e p

• i' e ; t ,

em -

I - ~ " '"

1

• Caustie His Tonk

' [*

EEI 2

• Po.tasetum r.. .e .Dydreside ItLa Teah e .e , . a , . . 3 . r e.

4

• Canette Pg V

% j -

F = JL- -- .r---- , ,

I* *'**""'I F 8 r--

' p- g., --y d ', W h j

• 8 'la'"*"own td Storage Tank 9 . Serie acid Ilia Tank N*Q g C

i or .  ;

• io w ste te =t r.u..ti- T. m

~ ~ 31 We Gas Compressor

_ 1 1 '< 12 Weste Gas Muistere Separator 31 - Weste Gee Decay Tsak

_. . ] O p.g3,, / i . w t. ces riiter

.- g' .

- / IS

• Reseter (helaat Bleed Helday Tank

,,4) 16

• Spent Pwel Coelant Pg

,_d. p

~ ,

/

[ 17

• Spent Feel Costere

\, 14 Spent Puel toelant Deelneraliser hogg .

19

• Spent Peel Coelant Filter

"~*' 9 e -

$_ 20 Seectes Building ? . et Cooler

.. , i 4 21 - Reseter tellding Componeet Coeling surge Tank E 6;; 3 i II - asmanium mydreasde stia Tank e I

- a . .-..

.I . .

auwa evwaa n-a-oi-=rr-idee-, e m e.1  :; ; ,. **

wt c RET; 4 18-67 Il l

1-i! STATION GENERAL ARRANGEMENT

. ~,

5 '

/ FLOOR PLAN dnN EL. 785+9 l '

l x OCONEE NUCLEAR STATION EnBy'"f"'""M M FIGURE 1-3 00000$g

b w e.s ~as~s:A uta vs.u

(

^

g

/L  ;

i_.y s'atew w .

. e. vt.a t . v. T w sc.s..w m- .

a.= s se< l [

W A4F.# W4 21 i

u. t-i _

'( Ass ur< s>al wm - - - - - - - - - - - . -

. .- .,--,m. , . - , _ . = - =----:.-----~~ - -

1 i e x t ;om se.am> =. As.s caas t

- - . .e y -- 3 LJ L- t I I

'< #r m

I 9

4 ,e Ch 'A" a. Si 6 .*

f

,.a f I MM.jd2 J tlAT'J6I 6 ,

AW % TGk"*Ou -, <,

g (( g r

[7 e .a nz ._ ,

i,

(- u .~, ev.m 64A4 hM6 ti CLWW*a - G.t aF A% JG

+

i O O  :. o O O 44.%

4

\ - e n  ;-

U V v 0 U 8 s ,

.l l' F __7, b,_----- 4 r_._ ~ ..

f.1e i

i o _s.2 e

( l 5 L - y

, r

v. t e

{I 9

. ---_ il' f. * '

~

• C 3 ,

_ _._ _ _ _.. t *'

E+ t P-r 9, # _.

-  :=*

\ f*'l ==i' y'.

lANl' ZE _ _ _ _ ,

.g u!. .mL.

-~.s , m_n._m _

• f

S ,%- 'J e- *J [ M l )

y g ic :. @;.greq_ g g-;g

i> u p.. -

m

,..., j/

s g-- -

a cu.: u_o % .c-- l i . ; . ; .. ;jd- j}

i o, Q[

mi _;1 2J14_ .f g

J.. i; - _ -

,-i

m. - . . l ,

/:

, !u_=' ? U ;- e n $_1t

_q

_ .a , s_ e i...p t e-y s i g

.___g

\ ,

(.. *

.,L' J' j e ,,

v v.-m "

j

! q .

\ t xN b / </ a t

\ . f-s , f

/

l

\x , ', g

, __w_____..=+- - -- -

x .- I

\, .- . p'A- 1 4

6 e-l A TEmah OatitJT L;l e.w i,

. ~ .

was,w su.,sa ; . ,

I swac Tw f l i

\ %/ /

\

/ \ -

x

\  %

m.e-

A t90*** K f

~'

gfELLIART $7titimC 3 . laver Egetpe.ne Rees

• Lenit 1 g . Lower Bgistpuems Saum e tantt 2 A,u 1 Ate d sLbgrattartl MV ge Drwuuni.g . Decooteesmet tee Ares

,.4 3 . Woth Area s . Leedtog Dock y . treek Asee

-T 4 . n.4i. . me . ,, un.r. .,, . a o.e.ne. i.nor.t.e r

~ ~

I b . Realth physics lahumrat.r.

) g . Cantentsi. sed feilet and Clothee th= ease are.

~~ ~

y J . Centaenanted Showes r 5

• Cl*** Shar**

' ~

$ gr

,  ; g

. . towenies.sa.ee

. no v .

t '

3 . Close tasker tese e . Generet Steesse 3- - - - p . Poseemsee Eleveter j

I l

} q . Pressht tieveter i U i- i a . sistr.

s . stor.se f g masseia ,

i LAe T C TVS Ob- ,sbe s.42a%,q q l' t . Eheet Sbaf t 3 . Cable Shaf t

.t..

i

- o a  !

4

, . . . , .._..r...-...r..

.....u_.e, .

~

EtB IP'WNT 1.Stettenv.ent i

O o O o o j ,,,...,...,

t i

3 O o O O o i 1

_. . _ _ _ ._ =__

,"==m-r . 9 4 .e

^~" c

_ _F, l

- l 5 l

• j I ,!

s (( D' k,I[s h ,q' __

                                                                                                           }                  l l

f/([l-

                                     - .O '- r-m
,/-                                                                                                                             i 0 ,t h " 1 ~ '
                                                             'S ,\g   . \.

g,'- <m,i i j ,f -,.4I 7 e y c f G iese-2 wT e c ar v j c- 6 *t s (L y> l N , (" J

                        .ui
                             -a                                       ,l l .,i . mye
                                                                       ~^

1 0- ' ' ~~-N

 \,o$     ( \K   -

_1' i, 3,,t .,

                                                                                                             '. s I

d-A - t GENERAL ARRANGEMENT

         'N

[] i FLOOR PIAN J J gw wawa tm4.#2E d iL*2 f' EL. 796+6 i vs. > l ( / .m.m l- . . . E .0CONEE NUCLEAR STATION FIGURE 14 00000$f ;l r. m

f a

 \

e

              ,f' %

f,,T A T 6N MCA2 T6J MM YJ l ! l l O ' ~ '2< es n2 m, -=mmm.-==_=.-e-.-r

                                              -n=

l

           @@dt4TE ATIOd Esu4pedG                                                                                                       itAI%e k,

b iThii 4,H W - , (['

                                                                                        ]              -~.

l @ .%

                                                                    \@'
                                                                    !f

[_j_-@ i.4 s

                                                               /

7 {

                                                                                 '                  s s    ' .s-Te,    t?. -
                                                                                                              \

stEAC7oG. @)@i4G , tJCIT I l i l '- < k x

                                                                                                    /j/                !\ ~e
                                                                                          /                ri i
                                                                                                                     .]
                      -        - . .      .                               .~'N            \1 C         T[-- 1 3D                                            ,

f f I @$

l 7 l 3 I

                                                                    "          AUXILIARY EUILDING A - Cable Room B - Passenger Elevator C - Freight Elevator D - Stair E - Storage F - Laundry G - Duct Shaft H - Cable Shaft I - Spent Fuel Pool s e uo                                                                          J - Fuel Cask Loading Are K - Penetration Room L - Roof EQUIPMENT 1 - Station Vent REV:   4-18-67 f                       @

l D32F S e_

               ,,    r=        @           --

I_

               ' ~

e i D c _l+ 4

                                                                "**cic'" e mdo 1                     --. _ _ __ _._ _ __      ___

9n _____. _ - us N l STATION GENERAL ARRANGEMENT

             @             \                                                               FLOOR PLAN 4,                                                                                                                         !

EL. 809+3 g a- a f at DCONEE NUCLEAR STATION I i FIGURE 1-5 00 00089 j 1

f G 6

              . 6TETIM b/M T d O ^ n ' u " ca **

M)67)JE f tpaat.TCiG - 2E-E A't 4 6 (-- -- nm:157 . e. w vauvec O O O __ _

                                               . . ~
                                                                                                                   "                                                   it r              p "p-   b _                  _ . _

fw < uvT 1 Tus4 2 .ssetap p { l i

                                                                                                                                                            "-j wm JiGTsATieu
       ~, _                                                                                                                                                        ~

O O O n '

                                                                        )

MoisTuce seeA2ATce esset.Tsas Tuaw

                                          <                             j N     /   _f             _6_ _ _g _ __f _ _ g _ _ . 4 _

i (- == i u. 2! l i

                                 )P                            b               ,,(

_4_ _ ) .4

                                                                                                                                   .I                         (

e. m . _ =a f

a im resg L.si 4f((. n i

                                                          .P'                                                                                  @j eb (s r-o
                                                                                /

m,L+ s g J

                                                                               /c1 e            o         (         @ O                       ,          a 1           e d
                                                                                                                                                               ,            t I-ceAcm wu [.     - _ ..

a , Q 4, ,-

                                                                                                               ) ,'

i s.l.i_N .

                                                                                                   'f                                                                -

GEE F.4J CL.E l 1" v

                                                                          \\                                                 J ,7,/ t, s                            ;

y\ i o , 9' o l

              % e. a       o_. g                                          % Qi                  a      - 8po./                          ur' ij h s.~ -s_     ._           .

s _. Q.... __.: i 1

   'l

P I k 4 .. --

                                 -=-=s-w=                                                                   _ . _ . . _                                :...---_---.--                                                       --

(~ ~ ]

( TD O O__ O _.

                                                                                              -s _                                                            ,j                               [                    _                                                               AUXILIARY BUILDING 3
                       "-' _ a                        ,

3

4. wt q Tune + meAnegu

    }                                                                                                 l.                                                                                                            _       __ _

{ -i u -- - j --_ _ p A - Control Room B - Electronic Maintenance 1 -- C - Kitchen I, O O O D - shif t Supervisor ( ) E - Operators Toilet j g o, F - Toilet g C - Storage (

     ,                                                                                                                                                                                                                                                                              H -  Passenger Elevator
      ,                                                                                                                                                                                                                                                                             I -  Freight Elevator j                                                                                  q') __ - %g_ __ __Q3_                                                                                                           . _ .           -
                                                                                                                                                                                                                                           \                    ,'                  J-S     ir

~k'* T "  ! . w uess .

                                                                                                                                                                                                                                            ,.y                                     L - Cable Shaf t
                                          )                                                        I f

h-~ f /\ M - Spent Fuel Pool J j- -C y 9. + [ N - Fuel Cask Loading Area O - Penetration Room P - Roof Q4 , d4 EQUIPMENT I A' LG - t

f. 1 - Station Vent l ?'

ei / r]{

h. e

                                                                                                ,7 'r / [

s e 2 ', 4 W REV: 4-29-67 [p, a' O'/b'y,ou-- . b ' U

                                            '                                      /3                                                                                                                                 -

y' ^

                                                                                                                                                                                                                                        ~~
               ' -                                                               .I V'                                                 (                            G                    . -. ....
                                                                                                                                                                                                                          ??s"                                   ,

s j , .

                                                                                                                                                                                                                          *e p.s w a ut                          -

STATION GENERAL ARRANGEMENT 3 [ f;.m --

                                                                       ,                   \,v                                l O ^, y FLOOR PLAN EL. 822+0 sav    N                      st_re --

s_

                                                                                                                                                                                           ..o..y        '

g y, t 3 .

                                                          -%[                                                                       ,
                                                                                                                                               ~.,_ y .

LH ' E OCONEE NUCLEAR STATION FIGURE 1-6 00000f(l { d . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . . _ . _ _ . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _.____..__..__.____________.___o

P l M c,T AT ich J E4174 O ^~"; ro a2 *T <------==-:==z.. . _ _ , _ -

                           --IC .m _ ---:-

J l I d

                                .i l

i i A W ajsT#ATioQ P>UI LD'O.1 TU E 6%E 6 3

                                                                                                                                      ~~          -

L n[ N $1 f _, i e - ay9 o d s,_ u

                                                                              #+               1
                                                                                                            =[]'

a li . l A

                                                                                                  =e V              c                      h_ !

y/ - 4 s R$. ~l ,l ,

                                                              'f        , M - - g,' s4                                          A' I
                                                                            ',.                                                                 qp
                                                                                           -(A f' - hG

[/]. . y : 7

                                                                      ~

o' (' ,a x1 h r ate:02 owes ,T-

                                                                                    > m '.

C \ w{n = -_

                                                                                                    ~ ~ ~~~)

va.y , .._.....a_p_a... I 4, 4 cc a ps pt M5 l 'Ge i .;5 e d 6 --

                                                                                    -% ..a - - - - l}

7-t 'j ,- ~ ) I ' i ,/ f "I l 3 N'4O.l.,~~%._ ,' i, ' : b'U

                                                                                                                   %j.. - -_ .

r; bNN t= r -1

                                                                                             /              '       4 e

i ti

O 1 1 J l L ei.ar . . . .. 5LO 44 AURELIABY BUI121NG

                                                                                                             & - Upper Equipesnt Room

• Floor Elevottos 837+0 3
• Spent Fuel Fool - Floor Elevatten 841+6 C
• Feel Cash 1mediae aree D = Stair 8 - Fossenger Elevater F = Freight tievatar C = Dset Shaf t B
• Ceble Shaf t

                                                                                                             ! = Caetaminered Tottet J - coches Ownse E - Coatsmiasted Ebaser g3                                                                                   L

• Cleam Shower N
• Toweling Area 7 N = Clean Locker tone and Clean Teilete I PfC'  : : " o,'..e p

W* - 0

1. Station West g 2 - mata steen Lines

                      .C'~ -
                                                                       /                                     3 = Feedwater Lines vW ,e                                       O f                                    4 . Fuel 11asef or Bridge S

• Fuel haandias crane m

w i 5 e

                    .=,         w*                   +

y x _ 4, == +-i n-u a .v',._ - A . U m q pr> _ _ i Of 4 Qltf-m_L9 l ,O

                     ,- )                                  l g k

yr . .-e i nur u o a -- ...,,,,, _....- L.. . ., ,

h. .._ . - . .__. _% c

                                     . pI \ \ ..'

I .' i sta r.q ue I tS l . ks bIi ( 'v) h j r j STATION GENERAL ARRANGEMENT K x\ \c(_f=uAO y -' vtoon etAN ele 837+0 & ele 841+6 p/ t [

                                                                                                                 - OCONEE NUCLEAR STATION GE                           FIGURE 1-7 0000;G                                ;

i e

 /

( g, -,s.. ,_

                                                    ~c -                               = = -.

O ! ".' =< _ w_ .= = == ==. - i i 87M stdlOTAlnT LM 6DIL9f*JG T UG 6'* .

    !                                                                                    Ei Y

L ._. i e Q: d h'~

                                                                                      'Db;
                                                           . [__ -                      G H

i ,- N+Q -

                                                                                               's +

d/ s g

                                                    /'                                       , . ,

MC'CA __ _ . _ _. _ ' 1 i... .__ j voix i MN$ n_ \. _ i t /i 1

                                                                                                            -=~~
                                                                                                                 .J g\
                                                                                            ,/                   ]'
                                                    \\N                      i
                                                                                       /

i

                                                                                                            ,l
                                                           $ ~ __ w;.. //                           *-
        %fM=1                                                      ~_   __.-

J f,

n .= -_ _ --- p l' b 1 V

                                                                                                         ~ ~ ' ~ ~~^
                                                                             ' ~ ~ ^ ~ ~ ~

____,,;ww-.-. -m-a=?""~'-~"' h i L I

 }          DWLAM i
                                                                                                                                                                                          ?

I

                                 @                                                                                      ~~

i AUXILIARY SU11DDIG r

                         ~ ~ -

i lim .; A - Ventilation Equipment - Elevation 857+0 3 - Spent Fuel Fool - Floor Elevation 841% C - Fuel Cask leading Area

        }

l (p D - Duct Shaf t E = Cable Shaf t I

                                 ^

h F = Passenger Elevator i g C Freight Elevator it - Stairs

            }                 ,

g J R g, { _s sm treet i G 1 Station vent

                                    )
 --                                                                                                                               RETT 6 16-67 p

I nac.tx ev- u>wc. . g, , Wrf'"r k H!

.~

[ ['O]) 77-GENERAL ARRANGEMENT I d i-FLOOR PIAN

            '                                     '                                                                                           EL. 857+0 y.

1 - Ot e OCONEE NUCLEAR STATION FIGURE 1-8 00000ff i

                 ,.                                        .-      -                  , . . _ ,     c- ,
                                                                                                                    , y              ,-         ,                ,4 # .-,.     ---e,'

.f ) s e . is s

                                                ,\ '\ \                                                          .        .             !

LotiuAM goLE2, VEuT

                                               ..j,'
                                                 .t x                                                                            /

weea MoGURE SURGE TMb6

                                                                                .              FE9A2ATo26 TV2Bo -C.E UE:2AToA, E              gaggA7gg4            I oO                                    O mA,a o,,uiu;               -

f{ b4,\fo-  ! - - -m_ C-+-

                               - -l                       oO                     ,,             Co                           a,

_.. _ ws .y li \ . I 3

• a I

( YV N ' l i '3r R n E'~6id "Qi} E pp b \ \ ,

                                                                                                                                         \
                                    .        ~                                                    '
                                                   .._~___ '.-

m_-________________ I Tocese suw.m 4 l l I 1  %

                                                                                                                                                                   \

STt.Ti':ma VeuT l

                                              >g:-                                                 A A - Basement Floor - Elevation 775+0
                                                                         !                                                 B - Mezzanine Floor - Elevation 796%

C - Operating Floor - Elevation 822+O h --- -~

                                                                                      /                      7             D - Roof High Point - Elevation 906%

l  ! E - Floor at Elevation 764+3 I F - Floor at Elevation 775+0

                                                   '                                                                       G - Floor at Elevation 780t3 i                                                                  H - Floor at Elevation 785+9 I - Floor at Elevation 796%
                                                   .                                                                       J - Floor at Elevation 809+3

K - Floor at Elevation 822+0 L - Floor at Elevation 837+O M - Floor at Elevation 857+0

         ,                                                                                                                 N - Floor at Elevation 841%

O - Crade at Elevation 796% l

      ,I                                            i
  ~

r , l 1 - Fue1 Handling Area I 2 - Spent Fuel Pool 3 - Hot Nachine Shop

 ]

[ $- - - 0 4 3

                                                                                                                             -  Stair Tower
                                                                                                                                .1unnet to Hot Machine Shop r                                                     %/

P = W- = i

                                                                                     @                                                REY: h-18-67
                                            -                                                -.a i

i. - _._k " I

_a P l _ (" Io UF_ tl ,,

                                                                                                               , :'. t          __f**".="

maae e ~4 ___J L.s/r -] U L - __________a _.--__.__a s. . . . . _ a. ;.

                                                                                              ' L.,.L.1 l
     ~-

lL -k=:.-Y W T'" "'M

.::: \"~k-i

                                                                                                     \=             . !? 'ne44 e.o va t.cv 6u n_p4G                                       gl SDHON CENERA1. ARRANCEMENT, CROSS SECHON 4_                  ::e t.cT:a. Agatcwa                              y m irowie OCONEE NUCLEAR STATION FIGURE 1-9 0000077 J,

i 4 3varson warw (' cag i 4- ~cm- If.__ __:::_ _::::::::[-

                                                                                    ==

a c on sv

                                                                      ..,,e                                                                                                                 C'.'.~-~

v ons s

                                                                                                                                                ~~                                    x
                                                                                                                   ,~.
                                                                                 -                          -                                                      x r
                                                                            /                                                                                      %                 Ns
                                                                         ,7                         7                                   j

- ..:.3.%

2. 2.. w

                                                                                                                                                                   ~.                      'sbs l '           ,a                                                                                                   .
        ',((f,     /w                            /             '
                                                                 ',-           ' Y UY ,Y)_]f.*4 l-                   'p[/                      STE AM GE NE R ATOR
                                                             /                                                                                     ,

ll / g . . .u,, --- . 4N, --

     .<woroce mr -                                              ere soe              m o rw *>      ...
                                             /                                                                                    w                          w
                                     /.                                                                                                                                                               '
                                                       - ui f                                                                                  
                                                                                                                                                                           \                        /
                                                                                                                                                                                            /

l *4 ,,/: - 4 .. , / f*.s!? s' / n* I

                                                                                                                                /              /                               \
                                   .                                                                                                       /        e,,es-e ,'
                    ' b,               ,0                                              /            ^
                                                                                                                                                                                ,i          _
                       .        ,te mrvnwr,                                                                         f       ---
                                                                                                                                             ,            (                s
                                         **"'-cEQ7[ y?3                                              --.-m                          .
                                                                                                                          -.._ra

_ \ _ i

                                       \                               :
                                                                               .. N                                                 \                      j g                            7

x. 2/gz

[LH .4 s 4 g - r.; f' /

                                       \                                  l              O. ,..
                                                                                                                                                                                                 /
                                        \ \.L e                                           ai                                                                                                4 (N&     \s N s
                                                                                            . .v- . e .

i,;. se t#TY? / I i I i

                                                                                                                                                                                             $c
                                                                                                                                                                                           . .W, A. -

s,- ,-

                                                                   \                                  */f , *,7f                             $TLAM GENEFATOR N'O \.                            C g,,,

N /.

                                                                                             's
                                                                                                                                                                                 ,,. //                 .

D - mricw mr o p p es e.s to Tre II.]ili l' ^ \ :. i aMcc E GRAPHIC SCALE 6 y TE mej o* :-D REAcTae stor. 1

i

                                                                                                     'J
                                                                                                      \

i GFj

    , , ,          cm , a                1                    ,

t::,:,:.:V ~< = ,- n k M) 1 . 2cfCl,. .s ca:s s,

                  ..N N)\. \ /,N.                                        3 ii.'fa

/

                          \. s I                              \
                                \             4:a        7 Em
                                               --3

~~--

bun ** t b~,,,

                      ~-

g j

                                            .  . . .o 9..                                             3 L         Au .i:,            . .i l.
      , m $5                     !

rY $ , t .~,z O N/

           /

f // / REACTOR BUILDING BASEMENT FLOOR EL 777+6 NOTES 2

                                                           ,pa. mry g ersa

• 3 Mckn.!?v"l,T'*

                                                           'IUr% % "Y,i"*"    OCONEE NUCLEAR STATION
                                                         -: xo e.ie ar FIGURE 1-10 00000fi              i i

s

9e 4

 /

e y sw,0 1-e m *

                                                                                                    .-   Y.--

r _2. f[- 5

                                                                            ~
                                                                                         >;r ' -                         ~

47')? r k i l en e wsa u

                                                     <-              \

s7 f

                                                                                               /

e x~

                                                             )                  ,5                                                    -

x

                                                    /XD    ,                                             /      f%                                                         ;'                                      *
                                              / /                                .

_/ ,, 8

                   \,

n% /

                                                                                   \'

p ... [ ,.l '-

                                                                                                           "0'Tll,L(m                                                  A                                             -

t. \-

t. u.....

j ,s ~ ~-1 . . . . -

                                                  ,                  y                                                     -
                                                                                                                                     -- )                                        c
                                      ,/     t    ,                                                                                         ,                                                             ,

ces

                                                                                                                                                     /k t--.Yt-ml r-                                                                                                                                                        \             \
                                                  #                                                                        J
                                   ' 'f-a ,

[ yp 5

                                                                                                                                                                                                               \
                                                ' .!J'4 . sD                                                                                                                        J'                           \

s

                                                                                                                                                             . y\                                        \ 2_,

4==: - I \ y, h-n

                                                                                                                                       .g+

j ,. - g.

       "'Oc'~h

_. -y*'y, - " l , .

                                                                                                                                                                                                              -aE   (

___ .;:--= --

                                                                                                                                                                                                                ; cm y                                   ,

As we u x"- hw .

                             ,                                                                                                                                                            j

i. .9

                                                                                                                                                  ..o$p'                          ;j
                                                ,i~     ~-. .
                                                                           ~
                                                                            \

( (. e .

                                                                                                                                                                                                        /

I g . ,-

                                                                                                                                                                                                            .f, <.y.-

q,1. t, . ,

                                                                                                                                                                        *~

i ,g )s,  ! *

                                                                                                                                 /                             -                   t 4 ' ~~ '. ' T',,f/          iIf       if               ", E 'O'4'.usY
                                                                                                                                                      ..- 3 u naven f
                                                                            ,-       .      ,.                                                                                            , /           '
                                                                                    .                                                                                              /

T;'t' -- $i a fp, .

                                                                                                                                                                  ]r#5,fl?h.
                                                              \
                                                                              *          /                                                                                                                               l ta(AsstnKT enatu                s                                                 t
                                                                                                                  . u n n , n J p,

! l 9

w_ s a

i h leg-C* 1.0 1 RE ACTOR BLDG. UNIT No I 5 o S to i5 a,o IIIII' I  ! I m e scAtt i CO

1

                                                                                                                                                  \

1 l l l 1;

                      i r
                                                                                                                                                    )

l

                   ==}

frd y__- .. s E

   }w?g-     a
           . ,w          0.w:

'fi .b m-

                             .e L' g             ,
                          -avd          ,
                    \                 m
         ;lp T _                     q. , vu-en x, Ni p/!:                  ~R            !
      .:\

1 l-V l

r i k'o7,/e j  ! I 1 ----I

                                               '"*"-+

I.}{u1= 11/ni~23*d.

                                  $fvdML

• I t

I f I t- - 1 l l,

                                            }

l l 1 1 le REACTOR BUILDING "5 t101E3: GROUND FLOOR

                                                  ' E A'll4" 7'~~'*'                             EL, 797+6 n m scrm v,ea aam mer~

SEE PM t-c e S !"W MCT*Lmeel VrEw LOCntNG 14:? Mi FM FfS R$VISCO 4 4 69 ont me OCONEE NUCLEAR STATION FIGURE 1-11 k 00 00@2SEiEe  !, 10' ia L

74 i st., o e ., - if

                                     ._f._                                                                                                                         /

I y##[

                                                          ,                              u. e u u ti e.L o.                                       l.lNi\
                                                                                                                                                     ,\
                                                                                                                                                               . < st een    rt rsy.o  s
                                              }'                         gS                               .

1., f.g - ' x .. b+pf ,

                                        - n.       3. .-                                            ,                                                                   .,*g, x'

i

                                                                                                                       ?T*:                                                 L              Jg
                         ~we@                  I                                               [ ,3                         l Z

N, ' s i s "'o * [. . _ y i x ) C""

                                               ,         L                !                           :1                   ;                   l_ , a                                       ,..c,.
                                                                                                                                                      /j M,         2
                                                                     ...., ~ ,.            g/,_                                                '

W=

            .p                                                                                                                                                          4           f. ,

j' -h*n'"NAU* l I pf, , G.': ,  ! x  ::gge

                                                                                                                             ~\
                                                                            ~
       . d.h ,                                                                                              /                                 '
                                                                                                                                                                 'l (c

[. ,,_. ,: ,r'j P. r-- M 1- ~ , \>gN,::i , i '

                     .cre. . -                                                                    '

a$'Oa ,'  ! ' i:_ - <

                . ,c.cy . nisis        ray,_.e .:q
                                       -c,..      ,.
                                        ..' pf AM N s.g
                                                                     ' v, li .'_                  ,

i lI li , 8 e- 2.< Masur( { j ( ((,', # . il / V- \ \ C.'n.%@ .'

                                                                                          \                                                      /                  e
                                                                                                                  %..      -                 l                             .

i - ,i g l

                                                                                                                         ,           ,/                          e

( l l E i m i . i N- r-y - 4  ; t j q y~

                                            -l                            .m.                                       /s                T j.M                                  =

a',3l r ca yJ'a,a,,,;

                                                                                                                 . w;,4,
                                                                                                                                                                                ^

l s t......; s#..4s/' 3 , 1- < -- s, 8 p *

                                                                                                                                                                              /

na er se s- .,f ma .m a i s .++,w

                                                                                                          \
                                                                                                                                                     \'y, s
                                                                                                                  ~~                                 ,N k asmerarmte y , e ,,3                                                   \-                        .

1 ( legit I.D. l ukT** No i o imd i [] I ' ' ' zi* I f, W . .W IC $CALE

  !                                                                                                                                  10 A

Y I A I l r- l l

                            ,         1 s                                     6 J'

s

                                 /

N '/'s '

                /                    ~

d i g+ A..L A  :.r.i) *m?' C 3 ' '.') ) ,

         .a

<

• l> .

2.k, T. f ~ v f(-- -

                          . - . ' _5_34T'_*3 4.c

[ .- . _ b_ _Ei5.

• Of"'

s T .. 6. s

         .b'
           -                                      s
     #(
   .h".'4 -
                                                   \  s 6
                                                 /
                                                /
                                             /

L REACTOR BUILDING UPPER CANAL FLOOR 4

                                  ,, ,j                                        EL. 816+6 NOTES s ,re3wmney mm :r ur
                                               ' =19 ,12.M"ll**

• in!'1",TTJ" i'0" HEv/ SED 4 /3-67 Ilt! OCONEE NUCLEAR STATION FIGURE 1-12 00 00070~ 103  ;

                                                                                                .f

i, I I f4 n. e3s'.o' s-w .

                                                                                       - =w . ,ut-, h                      qs
                                         / . , ..,>>,#>                                                                             s s
                                        / 

x: i ur

                                                                                                             %{

I.rr * - x f 4 t: ~Q,g-n ..n-Nx

                               . -} '  . . . . .

(k~*;lew argey **'i:N '....,~1d .. 4_ y .-ffIfuw= l. r . s/.y

           <ya        . h,n:=. ......' (e                                 .

Q. w,. \

                                                                                                         /. dot     .e .1,@3       f, ~ < ,,

0,

                                                 ......                                                               {                     .

(@: GbC } L

          '.'.s g ,e                                                                       , ?.Me
                                                                                                                                   '.f._ gf. ~'P x,,

{

                             ,-.-..         - . . . ,      s           m v.n.

x p, I.

                                                                                                                                    + w-                         7 v,.                                       ,

u__,,

                                                        =         I
                                                                                                                  .t.               !

n

                                                          ?                                                        i
                                                                - . -                _                         .i  .        ~ . ,. ..

x

                                           . . _                  i                                   !. A. : :l x /.
                                                                                                   .j,-            ,. _

s et sc_m. sca neerg er o rer.. e s (/ ~ \  !' l

                                                                                                       '\

LY i i

                                                                                                                                                         ~m
                                                                                    -m. = 1_ .                                                _m i

3 llilip 1m > m

4. m w --

                                                                                                                                                            /

j ;. ...

                                                                                                                                      ==

M,J 3 lT.

                                                                                                                                                          .w I

t" s l

                                               +                     x                               > .. w ...
                                                                                                                   ,c=>

( 4 ,,13.-i+ N s, i 3/ n e senev retoogn gausca;

                                                                                                                .I
                                            'st ta esS'-v V 1,6 O' l D.

REACtonBLOG UNT Peo.1 s e y e es no U fi d I . .. t.- a [u u s' ,:1 s .7  : n GAAPHC SCALE t \ loL-

 \

9 9 9 4 N N , y-- c

  - ~#7y ,,           - -
                             ...u
                                . sw OT:'J.~i ks.>af      .

f I~" :

b -

, i

                        \

v i I

. . , , ^ ~ ~ ~ fi
    .,         .....r.-

X., ) /4

--% / .:;%s

'L:L,7;;f,!,E h A 4 6

/

REACTOR BUILDING OPERATING FLOOR EL. 841+6 (AND ABOVE) O OCONEE NUCLEAR FIGURE 1-13 00 00077 i g i, i s .

1 a

                                                        /

N TrirQ

       ,, . c
           .             i n ..                                                                                     l                _ . - . . .           . _ _ _ . . -

_ .y! p ll, 4 r l /

                                                                                                                                                                                                                \

( b

                                        /
                                  //

qqJ s. 6 .

     -. u .-              .-

6[/

                                                                                                             $Ekj i

tt

                                                                                                             +i '* N" "

l

                                                                                                                  ~

u. s .- C, %*a*.I. tie' f' ,[ IW l I h,

l. .-.=

W I l-

                                           -tN/iN -                           .
                                                                                                  ?

I i . 11, .

                                                                                                                                                                        .s                        j                '

4%

                                  .m mm====

t,

                                                                                 .\3                                                                                f         ,'

j

                                                                                                                                                                   /             '
                                                                      /               _L) jii j                                            fi                                     _

w [ c-h h  ! 4..a . 4.. ;  ! I I! \ f I{A

                                                                                                         ]

j;,r' ,t .' 1 . . ! t

                                           '3 C. ' . '    f        ____n                      \

i g<.%

                                                                                                                                                                                                             ~
                                    ---                        , _4 ._41                {

I-I!  !

                                                                             ... .n, q: .

at:- - n # w-! y= WN i. . _' _Y f .. ~ _ . ._

                                           .a u[-.                            -

at +rS i g 4, li .H c: a ( fsitMll i

                                                                                            '.i                ..

J:fp$ns- Il-q '!,t-4 WlLw i Q ,p 3_ '

                                                                                                                   -, f., ,, . -. wwJ               :.-
                                                                                                                                                     % yy                 pl -

i i

                                                    ..e.
                                                     .               s,,

I . nsm i 1.

                                                                                                                                                                           . awa.

1. ,.

               -e
                                         =                       uw .n =                                  .                                    ,
             }"        g               .t"                               l
                                                                                  -   " " " ' * *                                              !     i                     ff I

[

                                                                                                ~

i

                                                                                                                                                                         < .8               . i
       .m !..                                                  i . ._
                                                                  . . _ . _        1j
                                                                                                                                                                          'T M = 9 l                                                                                                                                                        -:0            .
              .J .                                    _
                                                                     -@.   .                             . = .i g. JJ..a.3

• t-

                                                                                                                                                 ..p f.                                                   -
                                    ,, =                                                                                             l
                                                                                                                                                  =wl
                                     \'                                                                                              I
                                                                                    . . . . =

( hF.E..i.. .u h.m2

e

                                                                    ]
                                                                  \
  .......,.,n 4 ~<ig s . sz -

<: :. . ;,,1 e.Af t.f v d:. 4 q 1

                       .on' V

y. __. _ g ,1, Y' li.

In t.y1 III

           -.       + 1> u<

5- __. _ .-us _. _ _- me U __ w.2 m s_ REACTOR BUILDING SECTIONAL VIEW LOOKING NORTH - $e. ;s m i m m c- _ ,2  !)!Ahli m.: m .u l' i ' st at OCONEE NUCLEAR STATION 6 FIGURE 1-14

                                                          'O          a UUu[h

a k t

                                                                                                         ..,f     .....,.N*4                        .-

e #re e-

• 5 * <. t . e o. w._ _ _ .. _ __

                 , y }                                      ,
                                                                    /' -                                                l                                                                     ~

N 5 6 w w.n g -ptp - - ,-

                                                                                                   . =. .wn~ .. _. ,, . .. .. ,.=                                 .      m .w         -                      -.

4;

                                                                                                                        +

P f ( . _m .mt, come'

            'I41' & '

(T1F'n r f fl..

                                                                                                                     - - -                        ~                                                                                  i f

f, ge e = A to'.' ./ U eo l

                                 $~-                                                                                         i&

h. '-

                                 .s.

L'.O w

                                                                                                                                                                                                                 .-r_.

_.,. e. _ - - - . _ ,. y __7 4-j -_4 _o _} _ [__ 9 ._ Le n - p a ' n > L s _. ._-4 . > _. .

   ?

f .

                                                                            .u                             - - _ +                                                         p 7 __ _ m -t_,

l#- - I i ;i I i I - --h

                                               .1i73                    .. _J        .
                                                                                                                                                                                   ;a m                                          f
                                                                                                                  - o . .a4. .,,,. -

l ~ kA1 Tj -w if, g~ [l;] =ZOes

                                              ~ ~~, , , ,

17 ) E

                                                             -                          1 J,             i                                                                - i bl Y                                                    I                                               I                 n== = cai =

wy g; ,. ,- l e g:.r1c

                                                                                                             ,'s nw r            h M.,,

Janj s'

                                                                                                               '".f'.7L.

R

                                                  'LH1 - __.                                                                                                                                   W                             ~5
                                            ,                                                                     4.I Of-$ff                                          g wtasm staan e ass l

i .

                                                                                            > ..                       g ma artte              h'* **' "                               ${-[

Q WP i

                                                                                                                                                                '"*                 j
                  ,,,,,,,,m                   /,                   .~              1,k ,.s$.,,J,                                                                       i nt
                                                                ~~-

l j+

'- d. .J (h.3 t;

                                                                                                                                                                                        , . -           .a l
                                 . %'/w Ly 2        ,A               f                  /..)JH..(,.'c s

6*.7pg. Ird - - - I avil [-

                                                                                                                                                                                           <p.Ej.fqt
               **  *m   *'
                                           *4                        .Jf0
                               .r 3:.,
                                        ,'s,4 ,
                                                                                                       ~=~                                                              ; -==-~                               -~~ -
                                 /.Y.' *!'f,                                                                                                                                                        ~                         /~

ii

                                               -1I                                                                                           .

• o-

                                                                                                                                                                      .( % Egm
                                               ~y
                                                                            .,..                                     . c a. .,_                                     ._,

a _,,,M_, ,, u--i m-P ms'ma! */ *

• r w vtVn d. ppm. s 7 I O (6 i

l 1 (iff r,y sy. r a w

       .f y,4.
        .~,

I ss I I

                 .q
                 ~I
            ~
                 ._ J b

REACTOR BUILDING SECTIONAL VIEW 0V

  .h ;4 g;1                            LOOKING EAST             <

l

   .~05l(-lE
       'Ep                                                      l b         t rew   OCONEE NUCLEAR STATION       l

. ;. ,gg .. . . FIGURE 1-15

                                 .D n nnq7g vm   ..     ,    lo}}