ML19326B449
| ML19326B449 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/28/1968 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8004150797 | |
| Download: ML19326B449 (63) | |
Text
{{#Wiki_filter:.~~ -- --- d BEFORE THE UNITED STATES ATOMIC ENERGY COMMISSION In the Matter of ARKANSAS POWER & LIGHT COMPANY ..--,; g , c ;, RUSSELLVILLE NUCLEAR UNIT Docket No. 50-313 THIS DOCUMENT CONTAINS POOR QUALITY PAGES l
SUMMARY
DESCRIPTION OF APPLICATION FOR REAC'IOR CONSTRUCTION PERMIT AND OPERATING LICENSE September 28, 1968 8oo4150777 k
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9 TABLE OF CONTENTS i Page
- 1. INTRODUCTION. . . . . . . . . . . . . . . . . 1
- 2. DESCRIPTION OF SITE AND ENVIRONMENTAL CHARACTERISTICS WHICH INFLUENCE DESIGN. . . 5 2.l' Location . - . . . . . . . . . . . . . . . 5 2.2 Population . . . . . . . . . . . . . . . 5 2.3 Meteorology. . . . . . . . . . . . . . . 6 2.4 Surface Water Hydrology. . . . . . . . . 7 2.5 Ground Water Hydrology . . . . . . . . . 9 2.6 Geology. . . . . . . . . . . . . . . . . 9 2.7 Seismology . . . . . .. . . . . . . . . 10 2.8 Dardanelle Lock and Dam. . . . . . . . . 11 2.9 Environmental Radiation Monitoring . . . 12
- 3. DESCRIPTION OF RUSSELLVILLE NUCLEAR UNIT. .- . 14 3.1 Introduction . . . . . . . . . . . . . . 14 3.2 Reactor and Primary Coolant System . . . 15 3.3 Reactor Building . . .. . . . . . . . . 19 3.4 Engineered Safeguards. . . . . . . . . . 21-
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3.5 Instrumentation and Control. . . . . . . 23 3.6 Electrical Systems . . . . . . . . . . . 25-A 3.7 Auxiliary Systems. . . . . . . . . . . . 26 3.8 Steam and Power Conversion System. . . . 28 3.9 Radioactivity Control Systems. . . . . . 29 l 1 i
f ** 1 i TABLE OF CONTENTS (continued) ! Page
- 4. SAFETY ANALYSES. . . . . . . . . . .. . .. . 30
- 5. TESTS , INSPECTIONS , AND QUALITY CONTROL. .. . 33
- 6. RESEARCH AND DEVEIDPMENT PROGRAMS. . . . ... 37
- 7. TECHNICAL QUALIFICATIONS . . . ... . . . . . 43 7.1 Arkansas Power & Light Company. . . . . . 43 7.2 Bechtel Corporation . . . . .. . . .. . 45 7.3 Babcock and Wilcox Company. . . . . .. . 46
- 8. COMMON DEFENSE AND SECURITY. . . .. . . .. . 47
- 9. CONCLUSION . . . . . . . . . . .. . . ... . 49 APPENDICES APPENDIX A - List of References. . . . . .. . A-1 APPENDIX B - Figures . . . . . ... . ... . B-1 LL
1 1. INTRODUCTION 2 This document is a Summary Description of the Appli-3 cation, as supplemented by Supplements 1 through 10, of 4 Arkansas Power & Light Company (referred to as "the 5 Applicant") for a construction permit and facility 6 License to construct and operate the Russellville Nuclear 7 Unit on a peninsula in Dardanelle Reservoir on the 8 Arkansas River in Pope County, Arkansas. This Summary 9 Description includes information on the site and environ-10 ment, a description of the Russellville Nuclear Unit, 11 analyses of the safety aspects of the plant, a summary 12 of quality assurance procedures, a summary of the research 13 and development programs necessary for the final design, 14 the technical qualifications of the Applicant and its 15 principal contractors and considerations relating to the 16 common defense and security of the United States. 17 This Summary Description will constitute a portion 18 of the prepared testimony of the Applicant to be presented
-19 at its hearing before the Atomic Safety and Licensing 20 Board and is therefore being sponsored by an Arkansas 21 Power & Light Company witness, Mr. Harlan T. Holmes, i
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1 Assistant Manager of Production and Nuclear Project 2 Manager. 3 To assist Mr. Holmes in answering questions on cross-4 examination by the Board or another party, several techni-5 cat witnesses reprecenting the Applicant, its engineers 6 and contractors will make up a panel of technical expert 7 witnesses whose unprepared testimony will become a part 6 of ti.e Applicant's testimony before the Board.
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9 The Russellville nuclear generating unit will employ a l 10 pressurized water nuclear steam supply system furnished by 11 The Babcock & Wilcox Company (referred to as "B6H") and is 12 similar in design to the nuclear steam supply systems wi tch 13 are being furnished by B&W to Duke Power Company for its 14 Oconee Nuclear Station (AEC Docket Nos. 50-269, -270 and 15 -287), Metropolitan Edison Company for the Three Mile 16 Island Nuclear Station (AEC Docket No. 50-289), Florida 17 Power Corporation for the Crystal River Plant Unit 3 (AEC LE Docket No. 50-302) and Sacramento Municipal Utility District 19 for its Rancho Seco Nuclear Generating Station, Unit No.1 20 .(AEC Docket No. 50-312). A construction permit authorizing 21 construction of the Oconee facilities was issued in l 2 4
1 November 1967 and a construction permit authorizing con-i' struction of the Three Mile Island Nuclear Station was 3 issued in May 1968, both pursuant to Section 104 (b) 4 of the Atomic Energy Act of 1954, as amended. The nuclear 1 5 steam supply system will operate initially at core power 6 levels up to 2452 MWe, which corresponds to a gross l l 7 electrical output of about 850 MWe. An ultimate core 1 8 output of 2568 MWe is expected, and all steam and power 9 conversion equipment is designed accordingly. All plant 10 ! safety systems, including containment and engineered safe- ' 11 guards, are designed . nd evaluated for operation at this 12 higher power level. The higher power level is also used in 13 the analyses of postulated accidents to establish the suit-14 ability of the site under the guidelines set forth in 10 . 15 CFR 100. 16 The Applicant's construction permit application 17 including the suppl,ements thereto, has been reviewed by 18 staff of the Atomic Energy Commission, which has prepared 19 a safety analysis of the Application. The Advisory 20 Committee on Reactor Safeguards (referred to as "ACRS") 21 has also reviewed the Application, as amended through 22 Supplement No. 9, and reported its findings to the J 3
1 Chairman of the U. S. Atomic Energy Commission in a 2 lette.: dated September 12, 1968. The ACRS concluded "the 3 proposed reactor can be constructed at the Russellville 4 site with reasonable assurance that it can be operated 5 without undue risk to the health and safety of the public." 6 .The AEC staff concluded similarly. 7 The principal architectural and engineering criteria 8 which will govern the plant design are set forth in Section 9 1.4 of the Volume I and Supplement No.1 of the Applicant's 10 Preliminary Safety Analysis Report. These criteria 11 together with the engineered safeguards and other incor-12 porated systems provide assurance that the proposed 13 7ussellville Nuclear Unit can and will be constructed and 14 operated at the proposed location without undue risk to 15 the health and safety of the public. 1 4 I i 5 a , - - - , , , , - . , , - ~ - . -
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, i 1 2. DESCRIPTION OF SITE AND ENVIRONMENTAL CHARACTERISTICS 2 WHICH INFLUENCE DESIGN 3 2.L Location 4 The Russellville Nuclear Unit will be constructed 5 in the Southwestern part of Pope County, State of 6 Arkansas. The site of the unit is located six miles 7 West North-West of Russellville and 57 miles Northwest 8 of Little Rock, as shown in Figure 1, Appendix B. The 9 site and immediate vicinity are shown in Figure 2, 10 Appendix B.
11 All land comprising the site will be controlled to 12 the extent necessary by Arkansas Power & Light Company. l 13 This area includes certain portions of the bed and banks 14 of Dardanelle Reservoir which are owned by the United 15 States. An easement has been obtained which entitles the 16 Applicant to exclude all persons from these areas during (1) 17 periods when Applicant feels it is advisable. Land l 18 use is shown in Figure 3, Appendix B and dairy animal l 19 population is shown in Figure 4, isppendix B. 20 2.2. Population 21 The site exclusion area, which is under control of 5
I the Applicant, has a minimum radius of 0.65 mile. The 2 distance to the boundary of the low population zone has (2) 3 been established as four miles. The nearest popula-4 tion center of 25,000 or more is Hot Springs, located 55 5 miles South of tne site. There are no population centers 6 of 25,000 or more located within a 50-mile radius of the 7 site. 8 It is expected that the Dardanelle Reservoir will be 9 a major contributing factor to the part-time population 10 within a five mile radius. It is anticipated that the 75 li miles (approximately) of shoreline of the Dardanelle Reser-12 voir and Arkansas River will be developed as recreational 13 areas and week-end and holiday population will increase. i 14 Figure 5, Appendix B shows this estimated transient popula- l 15 tion within five miles of the plant site in 2012. 16 2.3 Meteorology 17 The site meteorology has been extensively investigated l 18 to provide an assessment of environmental consequences of , 19 routine and accidental releases of radioactivity. The 20 climate of the Arkansas River Valley in the region of the 21 site is primarily continental in character. The Boston 6 l l
1 Mountains, with elevations up to 2700 feet and oriented 2 generally east-west on the north side of the valley, have 3 an influence on the annual precipitation. The annual 4 precipitation on the south slope is on the order of 2.4 5 inches greater than in the valley. Within the valley , in 6 an east-west direction, the climatology is homogeneous. 7 A study was made of the site atmospheric diffusion 8 characteristics, utilizing conservative meteorological (4) 9 conditions. A meteorological program for the site was ! 1 10 initiated in the Fall of 1967. l 11 2.4 Surface Water Hydrology I l 12 In connection with the safety aspects of the proposed 13 nuclear power plant, surface water investigations were made. 14 These included the source and dependability of the cooling 15 water supply, magnitudes of possible floods and possible (5) l 16 failure of upstream dams. l 17 The plant will require 1700 cfs cooling water. This 18 water will be taken from the Dardanelle Reservoir down-19 stream east of the plant. The discharge will flow into 20 the Arkansas River southwest of the plant. No domestic 2L water supply is taken downstream of the plant to the mouth 7
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(5) L of the Arkansas River. 2 ' The minimum pool elevation in Dardanelle Reservoir is 3 336 feet. The highest experienced flood occurred in 1943, , 4 with a peak flow of 683,000 cfs. The levees along the 4 5 river channel in this area are generally designed for flow 6 of 830,000 cfs. The Dardanelle Dam is designed to hold a 7 water level no higher than 338 feet and to discharge 8 900,000 cfs. The maximum probable flood level was computed 9 by the Corps of Engineers as 1,500,000 cfs with 358 feet 10 flood level. Failure of Ozark Dam, immediately upstream 11 from Dardanelle Dam, during a maximum probable flood would 12 result in a maximum 361 foot water level at the site. 13 Nominal plant grade elevation will be 353 and ground floor 14 elevation for the building will be 354. Durinc - J.2< mom 15 probable flood the plant will be shut down. All Class I 16 structures are designed to resist this flood and all Class 17 I equipment is either located above elevation 361 ft. or 18 protected from flooding by the Class I structures. (Access
-19 to the plant would be by boat and/or helicopter.) The-20 minimum daily average flow compute'd by the Corps of
'21 Engineers during the driest critical month of the year is ;
(5) 22 -4,000 cfs. 8-
1 2.5 Ground Water Hydrology 2 The site is located on compact clayey soil overlying 3 dense shale bedrock and adjacent to the Dardanelle 4 Reservoir. This clayey overburden is generally impermeable 5 and hence ground water is not available. Ground water is 6 available in the bedrock fracture systems. It is confined 7 water which flows toward the reservoir under a relatively 8 flat gradient. 9 Water discharged at the surface and ponded will percolate 10 very slowly downward through the clayey soil overburden while 11 migrating toward the reservoir. In the unlikely event of 12 an accident, the clayey soils at the site will react with 13 any dissolved radionuclides and inhibit their migration. 14 The proximity of the site to the Dardanelle Reservoir l 15 will not adversely affect construction conditions. Domestic 16 wells obtain supplies from confined water in bedrock which 17 is under pressure. Tnus infiltration from the surface is (6) 18 not a problem. 19 2.6 Geology 20 The recent exploration program which included core, 21 auger, and wash-bore holes in addition to geologic mapping, 9 I
I a geophysical survey, and testing program were sufficient 2 to delineate the fouriation conditions relative to con-3 atruction of the proposed plant. The exploration and 4 testing program enabled construction design criteria to be (7) 5 formulated. i 6 Critical structures will utilize the underlying 7 Pennsylvanian McAlestec formation shale bedrock as founda-8 tion material. Other structures may be placed on the over-9 lying clayey material. These materials are adequate for 10 properly designed structures and should present no unusual (7)
'll construction problems.
12 2.7 Seismology 13 No active or recent faulting has been mapped in the 14 area of the proposed site. The London and Prairie View 15 faults located five and six miles,(respectively,
- 8) from the 16 site are the closest known faults.
17_ The proposed reactor structures will utilize the 18 shale bedrock as a foundation. This rock has good strength 19 properties and will result in no amplification of ground 20 motion from an earthquake. 10
s _. ___ 1 The area is not seismically active ; however, the 2 of fects of earthquakes from distant sources may be expe-3 rienced at the site. The New Madrid earthquake of 1811-4 1812, the epicenters of which were located about 220 miles 5 north-east of the site, is the type which would be felt at 6 the site. The maximam epicentral intensity for this event 7 was estimated at XII which probably decreased to about VI 8 in the area of the site. 9 Therefore, because of the above described site condi-10 tions and seismic history of the area, the maximum probable 11 intensity of VII is assigned to the site. This value is 12 conservative and corresponds to a design spectrum of 0.10g 13 for plant design with a factor of 0.20g for safe shutdown. 14 2.8 Dardanelle Lock and Dam 15 Dardanelle Lock and Dam forms the Dardanelle Reservoir 16 which provides cooling water for the Plant. An investiga-17- tion was performed to determine if this structure would 18 withstand the " Maximum Earthquake" of 0.2g without losing 19 its functional integrity. This investigation included a 20 stability and structural analysis of the following , 21 components: 11
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s e 1 a. Non-Overflow Section 2 b. Generator Section 3 c. Overflow Section 4 d. Lock Gates and Tainter Gates e. 5 Lock Walls 6 f. Earthfill Section 7 The investigation indicated that the " Maximum Earth-8 quake" could cause some distress and limited damage, but 9 the dam would not lose its functional integrity, and the (9) 10 normal control of pool level would not be interrupted. 11 An emergency cooling water pond of about 100 acre 12 feet will be dug at the location shown on Figure 2, 13 Exhibit B, to provide cooling water in the unlikely event of 14 destruction of Dardanelle Dam. 1 15 2.9 Environmental Radiation Monitoring ; i 16 Environmental radiation monitoring programs s'il be { 17 l conducted at the site with assistance from the State Health i 18 Department to establish existing background radiation levels 19 and to detect any changes which may occur. Lakt water, air, l 20 milk, lake bottom, soil and silt, vegetation and fish 21 . samples will be collected and analyzed for gross alpha and 12
i I gross beta-gamma activity. If any significant amount of i 2 activity is found, the samples will be analyzed for 3 specific radionuclides. Sampling points will be located 4 both C -site and off-site. This monitoring progre.m l 5 has begun and will continue after operations begin. I 1 l 13
d 1 3. DESCRIPTION OF RUSSELLVILLE NUCLEAR UNIT i 2 3.1. Introduction 3 A description of plant features and layout, as well 4 as an evaluation of plant safety are set forth in the i 5 Application, as supplemented. The plant description 6 emphasizes the concepts, guidelines and criteria which 7 will govern final design. The station will consist of a 8 reactor building, an auxiliary building (including control l
- l 9 room and radwaste area), a turbine structure, a fuel i
10 storage building, a shop and storeroom, an administration 11 building, a cooling water pond, a switchyard and various ! 12 other auxiliary structures and equipment. A plot plan of ' 13 the Russellville Nuclear Unit, indicating the general 14 station layout, is shown in Figure 2 of Appendix B. Table 15 1-2 in the Application sets forth a comparison of the 16 design parameters of the proposed Russellville Nuclear 17 Unit with the Duke Power Company's Oconee Units 1, 2, and 18 3; Florida Power and Light Company'J Turkey Point Units 3.9 3 and 4 ; and Florida- Power Corporation 's Crystal River - 20 Plant Unit 3. The following is a summary of the principal 21 features of the plant which are significant with respect i 22 to safety considera tions : l l l l 14
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1 3.2 Reactor and Primary Coolant System 2 The reactor for the Russellville Nuclear Unit is of 3 the pressurized water type. It has an initial rating of 4 2452 MWt, corresponding to a gross electrical output of (11). 5 about 850 MWe. The nominal operating pressure for the 6 reactor is 2185 psig, with an average temperature of 579 F. 7 The reactor coolant system is designed for 2500 psig 8 pressure and 650 F temperature. 9~ The reactor core is approximately 129 inches in (13) 10 diameter, with an active height of 144 inches. It is 11 made up of 177 fuel assemblies, each consisting of a 15 by 12 15 array of rods enclosed in a square, stainless steel, 13 perforated envelope. The array of rods consists of 208 14 zircaloy tubes containing uranium dioxide, 16 control rod 15 guide tubes and a center tube available for an in-core (14) 16 instrumentation assembly. There are approximately (12) 17 201,520 pounds of uranium dioxide in the core. 13 The thermal and hydraulic design limits of the core 19- are conservative and are consistent with those of other 20 pressurized water reactors currently in operation or (12, 15) 21 under construction. I t L5 i
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l . Core rcactivity is controlled by a combination of 69 2 movable control rod assemblies and a neutron absorber 3 dissolved in the coolant. The control rods are an alloy 4 of. silver-indium-cadmium encapsulated in stainless steel. (16) 5 The dissolved neutron absorber is boric acid. t 6 The control rods are used for short-term reactivity 7- control associated with the changes in power level and also 8 with changes in fuel burn-up between periodic adjustments of (17) 9 dissolved boron concentration. The reactor can be shut 10 down by the movable control rods from any power level at (18) 11 any time. Each movable control rod assembly contains 16 12 control pins, and is actuated by a separate control rod L3 drive mechanism mounted on the top head of the reactor 14 vessel. Upon trip, the 69 control rod assemblies fall into (19) 15 the core by gravity. 16 Systems are provided so that the concentration of 17 dissolved neutron absorber in the reactor may be adjusted la to maintain the reactor shutdown at room temperature and to (20) 19 provide a safe shutdown margin during refueling. The 20 concentration of dissolved absorber is reduced to compen-21 sate for long-term reactivity changes, burn-up of fuel and 16 d e- -
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1 buildup of fission products over the core cycle. 2 The core is contained within a cylindrical reactor 3 vessel having the dimensions of 14 feet 3 inches inside 4 diameter and 37 feet 4 inches in overall inside height. 5 The vessel has a spherically-dished bottom head with a (21) . 6 bolted, removable, spherically-dished top head. The 7 reactor vessel is constructed of carbon steel with all 8 interior surfaces clad with austenitic stainless steel. The 9 reactor vessel is manufactured under close quality control, 10 and several types of nondestructive tests are performed 11 during fabrication. These tests include radiography of 12 welds, ultrasonic testing gggneticparticleexamination 13 and dye pentrant testing. During operation, specimens ; l 14 of reactor vessel materials will be placed in the reactor 15 near the inside surface of the reactor vessel. These ; 16 specimens are subject to irradiation similar to that to 17 which the shell of the reactor vessel is exposed. They 18 will be removed periodically and tested to ascertain the 19 (23) effects of radiation on the reactor vessel material. 20 Two coolant loops are connected to the reactor vessel 21 by nozzles located near the top of the vessel. Each loop 17
I contains one steam generator, two motor-driven coolant 2 pumps and the interconnecting piping. The reactor coolant 3 piping is carbon steel clad on the inside surface with (24) 4 austenitic stainless steel. Reactor coolant is pumped 5 from the reactor through each steam generator and back to 6 the reactor inlet by two 88,000 gpm centrifugal pumps 7 located at the outlet of each steam generator. 8 The steam generator is a vertical, straight-tube-and-9 shell heat exchanger which produces superheated steam at 10 constant pressure over the power range. Reactor coolant 11 flows downward through the tubes, and steam is generated (26) 12 on t he shell side. 13 The reactor coolant pumps are vertical single-speed, 14 shaft-sealed units having bottom suction and horizontal 15 discharge. Each pump has a separate single-speed top-16 mounted motor, which is connected to the ,, ump by a shaft (25) 17 coupling. 18 The pressurizer, a vertical surge tar.k approximately 19 half-filled with reactor coolant and half-filled with 20 steam, is connected to the reactor coolant system to 18
7 1 control system pressure. The operating pressure of the 2 system'is maintained by operating electric immersion 3 heaters to increase pressure or by spraying reactor 4 coolant water into the steam within the pressurizer tank 5 to reduce pressure. Self-actuated safety relief valves 6 connected to the pressurizer prevent overpressurization (27) 7 of the reactor coolant system. 8 3.3 Reactor Building 9 The reactor building is designed to completely enclose 10 the reactor coolant system and portions of the auxiliary 11 and engineered safeguards systems (see Figure 6, Appendix B). 12 It is a reinforced concrete structure in the shape of a 13 cylinder with a shallow domed roof and a flat foundation 14 slab. The cylindrical portion is prestressed by a post-15 tensioning system, consisting of horizontal and vertical 16 tendons. ' The dome has a three-way post-ten.=ioning system. 17 The building will have three buttresses to which tendons 18 will be anchored instead of six in order to facilitate the 19 arrangement of penetrations and of other equipment within 20 the building. The foundation slab is conventionally re-21 . inforced with high-strength reinforcing steel. The entire 22 structure is 1.ined with' welded steel plate, 1/4-inch minimum 19
1 The building is designed to sustain safely all 2 internal and external loading conditions which may reason-3 ably be expected to occur during the life of the station 4 or which could result from the postulated design base 5 accident to the reactor's primary coolant system. The 6 tendon system used in the structure is of the unbonded type 7 with a protective compound used to prevent corrosion. Prior 8 to construction, a test will be conducted on t he liner 9 place anchorages to verify certain factors of design 10 analyses. 11 The reactor building is so designed that, u.th the ' 12 engineered safeguards systems provided, any leakage of 13 radioactive materials to the environment will result in 14 doses well within AEC's 10 CFR 100 guidelines for any of 15 the postulated accidents. The integrated leak rate at 16 design pressure will not exceed two-tenths of one percent (28) 17 by volume, within 24 hours. 18 Prior to operation, the reactor building will be 19 subjected to a structural integrity test and leak rate test. I 20 The structural integrity test will be conducted at 115% of 21 design pressure. Periodic leak rate tests will be performe A ' 20
-,o 1 rate test. The structural integrity test will be conducted 2 at design pressure. Periodic leak rate tests will be performed i
3 to assure integrity at the reactor build,ing. A tendon 4 ' surveillance capability will be available to provide - 5 assurance that the tendons are free from harmful corrosion 6 an,d that excessive steel relaxation has not taken place. 7 3.4 Engineered Safeguards 8 Engineered safeguards are provided to fulfill the 9 following functions in the unlikely event of an accident: 10 a. Minimize the release of fission products from the 11 fuel to the reactor building atmosphere. 12 b. Ensure reactor building integrity and reduce the 13 driving force for building leakage 14 c. Remove fission products from the reactor building 15 atmosphere 16 The engineered safeguards systems can be grouped into 17 an emergency core cooling system, reactor building cooling 13 systems and fission product control systems. 19 The emergency core cooling systems contain both passive , 20 flooding and pumping equipment. The passive flooding 21 equipment consists of two pressurized core flooding tanks 22 'which automatically discharge borated water into the 21-
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I reactor vessel in the event the reactor system pressure 2 drops below 600 psi. The pumping equipment consists of 3 two completely independent sub-systems. Each sub-system 4 contains both a high pressure and a low pressure injection 5 pump. Either sub-system, in conjunction with the core 6 flooding tanks, is capable of protecting the core for any 7 size leak up to and including the double-ended rupture of 8 the largest reactor coolant pipe. Either sub-system can 9 supply coolant directly from the borated water storage tank 10 or by recirculation from the reactor building sump through 11 heat exchangers which cool it before it is returned to cool (30) 12 the core. 13 The reactor building cooling system, which is made up 14 of two separate and independent heat removal systems, limits 15 the pressure in the reactor building following a loss-of-16 coolant accident. One system contains three separate fan 17 and cooler units. The other system c'ontains redundant spray 18 headers which spray low temperature borated water into the 19 reactor building to cool it. Each of these systems inde-20- pendently has the heat removal capability to maintain the (31) 21 reactor building pressure below its design pressure. 22 Control of fission products following a loss-of-coolant 22
Wes5 h U A I accident is provided by the reactor building itself and by 2 a second separate engineered safety feature for limiting 3 release of fission products from the reactor building. The 4 second means for fission product control is the iodine 5 removal spray system which utilizes sodium thiosulphate 6 mixed in the reactor building spray water to absorb the 7 iodine released from the reactor during an accident and J 8 renders it unavailable for leakage from the reactor building. i
- 9 The reactor building and the iodine removal chemical spray 10 system will limit radiation doses at the exclusion radius 11 and low population zone boundary to values within the 10 CFR 12 100 guideline values. In addition, room has been pro-13 vided for charcoal filters if it is subsequently determined 14 that they are needed.
15 3.5 Instrumentation and Control , I 16 A complete and dependable network of instrumentation' 17 and controls will be provided to ensure safe operations l 18 of Russellville Nuclear Unit. The reactor protective system 19 monitors parameters related to safe operation and shuts l 20 -down the reactor if an operating limit is reached. (33) This 21 will be accomplished by interrupting power to the control i i 23
1 rod drive clutches and allowing the control rods to drop (34) (35) 2 into the reactor core. Alarms are provided to 3 alert the operator to abnormal operating conditions, and (36) 4 interlocks are provided to prevent abnormal operations 5 which could lead to potentially unsafe conditions. 6 The nuclear instrumentation system monitors reactor 7 power from start-up level through 125 percent of full power 8 operation. There are separate, overlapping instrumentation 9 channels for the start-up power range, the intermediate (37) 10 approach to power range, and the power operation range. 11 A control system automatically monitors reactor system con-12 ditions and the load requirements on the turbine-generator 13 unit, and adjusts reactor power, steam generator feedwater (38) 14 flow and the turbine throttle for safe, efficient operation. 15 The engineered safeguards protective system monitors 16 plant conditions and automatically initiates operation of (39) 17 the engineered safeguards systems, if required. 18 Following proven power station design philosophy, all 19 control' stations, switches, controllers and indicators 20 necessary to start-up, operate and shutdown the nuclear unit i I 24 I t -
I will be placed in the centrally located control room. 2 There will be sufficient infornation display and alarm 3 monitoring to ensure safe and reliable operation under 4 normal and accident conditions. Design is such as to 5 permit shutting down the reactor from outside the control 6 room. 7 The report of the Advisory Committee an Reactor 8 Safeguards for the Russellville Nuclear Unit indicated 9 that the instrumentation design should be reviewed for 10 common failure modes, and that it should be shown that 11 the interconnection of control and safety circuitry will 12 not significantly affect safety considering the possibility 13 of systematic component failures . During the detailed 14 design of the instrumentation systems their immunity to j l 15 common failure modes will be evaluated. The possibility ! 16 of systematic, non-random, concurrent failures of redun-17 dant devices, not considered in the single failure 18 criterion, will be taken into cccount in the evaluation. 19 The instrumentation signals sent to control and safety 20 circuits from common transmitters are made fully inde-21 pendent by the use of isolation amplifiers. The 25
I effectiveness of these devices has been demonstrated by 2 analysis and by actual test of prototype equipment as 3 described in Supplement 3 to the PSAR, Question 6.4. 4 3.6 Electrical Systems 5 The design of the electrical s ystems for the 6 Russellville Nuclear Unit is based on providing the 7 required electrical equipment and power sources to ensure 8 safe, reliable operation and safe, orderly shutdown of 9 the unit under any normal or emergency conditions. Four 10 sources of power, each possessing various degrees of 11 redundancy, are available to ensure a supply of electrical 12 energy to the station safety systems under any accident 13 conditions, including the loss-of-coolant accident, as 14 outlined below: 15-i
- a. Two 500-kv transmission lines can supply power l i
16 for the station auxiliary load chrough Start-Up 17 Transformer No. 1 connected to the 22 kv tertiary 18 of the 500 kv-161 kv bus tie autotransformer. 19 b. Start-Up Transformer No. 2 will provide an 20 alternate off-site power source from the 161 kv 21 ring bus, supplied by two 161 kv transmission linas. 25 -A l l'
l l i i i c. Tne main generator will continue to supply the 2 station auxiliary load upon abrupt separation 3 from the 500 kv and 161 kv systems. 4 d. Upon loss of all sources of power described in 5 (a), (b) and (c) above, power will be supplied 6 from the two automatic, fast start-up diesel 7 engine generators. These are sized so that either 8 can carry the required engineered safeguards load. 9 The unit will generate electric power at 22 kv, which 10 will be fed through an isolated phase but to the unit main 11 transformer where it will be stepped up to 500 kv trans-12 mission voltage and delivered to the switchyard. The 500 kv 13 switchyard, in turn, is linked to the existing 500 kv trans-l 14 mission network by two 500 kv circuits, and is tied to the 15 161 kv system by a bus tie autotransformer. 16 3.7 Auxiliary Sy; cems 17 Auxiliary systems are provided to supply reactor 18 coolant makeup and seal water, to cool the reactor during 19 shutdown, to cool components, to ventilate station spaces, 20 to handle fuel and to cool spent fuel. 26
L . Reactor coolant makeup and seal water is supplied 2 by the makeup and purification system. This system, which 3 also serves the engineered safeguards function of providing 4 high pressure emergency core coolant, maintains the proper 5 coolant inventory in the primary system, maintains the seal 6 water flow, adjusts the concentration of dissolved neutron 1 7 absorber in the reactor coolant and maintains proper (40) 8 water chemistry. 9 The decay heat removal system cools the reactor when the 10 reactor system is depressurized for maintenance or refueling, 11 This same system serves the engineered safeguards functions 12 of providing low pressure emergency core coolant and of 13 recirculating borated water to cool the core in the unlikely 14 event of a loss-of-coolant accident. The chemical addition and sampling system adds boric 15 16 acid to the reactor coolant system for reactivity control, 17 potassium hydroxide for pH control, and hydrogen and 18 hydrazine for oxygen control. This system is also used (42) 4 19 to take reactor coolant and steam generator water samples. 20- The cooling water systems maintain temperatures 21 throughout the equipment and structures of the station.( 27
1 Appropriate normal ventilation systems are provided in (44) 2 the station. 3 A fuel handling system (45)provides the means for 4 safe, reliable handling of fuel from the tLme it enters 5 the station as new fuel until it is shipped from the station 6 as used fuel. Irradiated fuel is handled under water at all 7 times until after it is placed into a shipping cask. The ! 8 water provides'a radiation shield as well as a reliable
! 9 source of cooling for the irradiated fuel assemblies. A 10 spent fuel cooling system maintains the temperature and 11 purity of the spent fuel storage pool water within acceptable 12 limits.
13 3.8 Steam and Power Conversion System
)
l 14 The steam and power conversion system is designed to 15 remove the heat energy generated in the reactor core by 16 producing steam in the two steam generators. This heat 17 energy is converted to electrical energy by the turbine-18 generator. A cooling water system utilizing Dardanelle 19 Reservoir water will be used to dissipate the thermal 20 energy rejected by the turbine condenser. This cycle, , 21 including the necessary equipment to achieve safe and l l l 28 l
I reliable operation, is similar in concept and design to 2 turbine-generator cycles in successful use for many years. 3 3.9 Radioactivity Control Systems 4 Radioactive gaseous, liquid, and solid wastes in the 5 station are handled by the waste disposal systems. These 6 systems contain the equipment necessary to safely collect, 7 process and prepare for disposal the radioactive wastes 8 which result from reactor operation. These systems are 9 designed to minimize the release of radioactive material 10 from the station to the environment and will maintain 11 releases below the limits of 10 CFR 20. 12 A process radiation monitoring system monitors effluent 13 released to the environment and provides an early warning 14 of possible equipment malfunction or potential radiological 15 hazard. The radiation monitoring system includes a com-16 bination of continuous-automatic-monitoring and periodic-17 sampling. 18 Shielding throughout the station ensures that radiation 19 doses to the general public and to operating personnel
. 2.0 during normal operation are well within the limits of .21- 10 CFR 20.
l l 29 l _ ._ - , - _ - - . . - . . . - \
~ -- _
a 1 4. SAFETY ANALYSES 2 Potential malfunctions or equipment failures have been 3 analyzed to provide a safety evaluation of the Russellville 4 Nuclear Unit. This evaluation demonstrates that the public 5 will not be exposed to radiation in excess of the limits 6 established in the AEC's regulation for siting requirements, 7 10 CFR 100, even in the very unlikely event that one of the (47) 8 accidents postulated in the Application should occur. 9 Two categories of malfunctions or equipment failures 10 have been analyzed : those in which the core and coolant 11 boundaries are protected, a nd those in which one of these 12 boundaries is not effective and standby safeguards are 13 required. The core and coolant boundary protection analysis i 14 shows that in the event any of the postulated malfunctions 15 were to occur, the normal protection systems operate to 16 maintain the integrity of the core and of the coolant (48) 17 boundary. The standby safeguards analysis demonstrates 18 the capability of the engineered safeguards systems to 19 assure protection of the public for postulated malfunctions 20- in which the normal protective systems may not maintain the 21 integrity of the core and coolant boundary. These j 30
I analyses show that for all credible malfunctions the 2 radiation exposure to the general public is well below the 3 limits prescribed in 10 CFR 100. 4 Of the postulated equipment failures, a loss-of-5 coolant accident is the most severe. Emergency core cooling 6 equipment is provided to prevent clad and fuel damage that 7 would interfere with continued core cooling for reactor 8 coolant system failures up to and including the complete 9 severance of the largest reactor coolant pipe. The core 10 cooling system ensures that the core will remain in place (50) 11 and intact. The reactor building spray or emergency 12 cooling units maintain the integrity of the reactor building. 13 The iodine removal sprays in conjunction with the reactor 14 building assure that the public is protected from radiation (52) 15 and radioactive material. Emergency electrical power is 16 available on-site to ensure operation of these systems even 17 if all external sources of electric power to the plant are 18 assumed to be unavailable at the ' time of the accident. (53) 19 Results of the safety analyses show that, even in the 20 unlikely event of a loss-of-coolant accident, no core (52) 21 melting will occur. However, in order to demon strate 31
f
)
I that the operation of a nuclear power station at the pro-2 posed site does not present any undue hazard to the 3 general public, a hypothetical accident has been analyzed 4 involving release of 100 percent of the noble gases, 50 5 percent of the halogens, and 1 percent of the solids in the 6 fission product inventory. The analysis evaluated both the 7 direct radiation exposure and the potential total dose to 8 the thyroid from the inhalation of fission products which 9 are assumed to leak from the reactor building. The low 10 leakage rate of the reactor building and the iodine removal 11 spray system reduce the potential radiation dose to the 12 thyroid to below the 10 CFR 100 guidelines even in the (54) event of such a hypothetical occurrence. 13 1 + 6 l 32
. . . _ .- ~~ .
i l 5. TESTS, INSPECTIONS, AND QUALITY CONTROL 2 Pressure containing components of the reactor coolant 3 system will be designed, fabricated, inspected and tested 4 in accordance with Section III, Nuclear Vessels, of the 5 American Society of Mechanical Engineers Boiler and 6 Pressure Vessel Code. The piping will meet the applicable 7 provisions of Power Piping USA Standards and associated 8 nuclear code cases. Non-destructive testing, including 9 radiography, ultrasonic, magnetic particle, and liquid pene-10 tration examinations will be performed during fabrication of 11 the nuclear vessels. 12 Auxiliary systems and equipment will be designed, 13 fabricated and tested to the appropriate provisions of 14 recognized codes and standards of organizations such as the 15 American Society of Mechanical Engineers, American Society 16 for Testing Materials, USA Standards Institute and Institute 17 of Electrical and Electronics Engineers. 18 A comprehensive field testing program will be conducted 19 to ensure that equipment and systems perform in accordance 20 with design criteria. l l 33 l i i-
-f, - , w + , - -
L 5. TESTS. INSPECTIONS , AND QUALITY CONTROL 2 Pressure containing components of the reactor coolant 3 system _will be designed, fabricated, inspected and tested 4 in accordance with Section III, Nuclear Vessels, of the 5 American Society of-Mechanical Engineers Boiler and 6 Pressure Vessel Code. The piping will meet the applicable 7 provisions of Power Piping USA Standards and associated 8 nuclear code cases. Non-destructive testing, including 9 radiography, ultrasonic,. magnetic particle, and liquid pene-10 tration examinations will be performed during fabrication of 11 the nuclear vessels. 12 Auxiliary systems and equipment will be designed, 13 fabricated and tested to the appropriate provisions of 14 recognized codes and standards of organizations such as the I 15 American Society of Mechanical Engineers, American Society 16 for Testing Materials, USA Standards Institute and Institute 17 of Electrical and Electronics Engineers. 18 A comprehensive field testing program will be conducted 19 to ensure that equipment and systems perform in accordance 20 with design criteria. 33 l l
1 The reactor building will be designed and built in 2 accordance with applicable pcreions of the Building Code 3 Requirements for Reinforced Concrete, ACI 318-63: Specifi-4 cation for Structural Concrete for Buildings, ACI 301-66; 5 AISC Manual of Steel Construction; ASME Boiler and Pressure 6 Vessel Code, Sections III, VIII, and IX. Materials and 7 workmanship will be inspected to ensure compliance with 8 appropriate codes, specifications, and standards. Materials 9 to be inspected and tested include concrete, liner plate, 10 prestressing system materials, hatches, penetrations, 11 structural and reinforcing steel. 12 The reactor building will be structurally tested at 115 13 percent of design pressure by pneumatic test. In addition, 14 it will be leak tested to ensure compliance with a maximum 15 allowable gross leak rate of 0.2 percent by volume per 24 ; 16 hours at the design pressure. Provisions have been included l 17 for in-service pressure testing of equipment and personnel { 18 hatches and other penetrations. 19 Consideration has been given to the inspectability of ! 20 the reactor coolant system in the design and arrangement of 21 components. Access for inspection of the reactor coolant 34
, , - . ~ . ~ . I i l l 1 system includes access for visual examination by direct 2 or remote means. 3 The Applicant's contractors and major equipment suppliers 4 will provide required quality control functions, procedures 5 and techniques to assure manufacture and construction in 6 accord with the plant design and specifications furnished 7 to the Applicant by its architect / engineers, Bechtel 8 Corporation. B&W has an extensive quality assurance program 9 organized and functioning with respect to both equipment of 10 its own manufacture and equipment purchased by it from other 11 vendors. The general contractor has not been finally 12 selected, but this contractor will be required to provide a 13 satisfactory quality assurance program. 14 In addition, Bechtel Corporation, in its construction 15 management function, will provide a complete q uality 16 assurance program covering tests and inspection both in 17 suppliers' shops and on the site of construction and 18 erection. 19 Applicant hcs 9 quality assurance organization which 20 is separate and indepcndent from its vendors, contractors 35
-,- - .a.
1 and construction manager. Through its own employees and 2 independent consultants it will monitor the adequacy of 3 quality control procedures followed in the design, fabrica-4 tion, construction, erection, transportation and testing
-5 of reactor components, equipment and structures.
I l - l 36
4 er> Di 6 " E I 1 6. RESEARCH AND DEVELOPMENT PROGRAMS 2 The nuclear steam supply system for Russellville is ) 3 similar in concept to several projects already in operation, 4 under construction or recently licensed by the Atomic 5 Energy Commission. The preliminary design is based on 6 technical data which has been developed in the nuclear 7 industry and on data developed by B&W which is specifically 8 related to the Russellville Nuclear Unit design. To 9 complete the final detail design of some components addi-10 tional technical information will be obtained. i 11 The following are the areas of the plant design in which 12 additional technical data will be developed to finalize 13 design details. 14 a. Once-Through Steam Generator 15 The design of the once-through steam generator 16 is based on experimental work on boiling heat 17 transfer and data obtained by B&W in full length
'8
. model tests of the unit. The testing of a proto-19 type unit has been completed but is not yet 20 documented. It incluied performance, mechanical, 21 vibration and blowdown tests, and control system 37
1 development. The results have confirmed the 2 analytical predictions of performance, and suffi-3 cient data on the performance and structural design 4 has been obtained from operation of the test models (55) 5 to finalize the design of the steam generators. 6 b. Control Rod Drive Unit 7 The design of the control rod drive mechanisms is 8 based on a principle which has been used in operating 9 reactors and which has been extensively tested by 10 B6W. Test programs have included full scale proto-11 type testing under no-flow conditions, full scale 12 prototype testing at operating conditions, including 13 flow, and components testing. Testing of a proto-14 type mechanism was carried out for a full-lite cycle ' l 15 of strokes and trips, and major design parameters were confirmed. 16 Life cycle testing has been repeated i l 17 using a miter gear of improved material and daowed { 18 satisfactory performance. Data from these test 19 programs are being incorporated into the final l 20 design of the control rod, its guide structure and . 1 (56) 21 the control rod drive mechanism. 1 i 38
L c. In-Core Neutron Detectors 2 The performance and .'ongevity of the self-powered 3 detectors are being demonstrated by detectors 4 installed in the Babcock and Wilcox Test Reactor (57) 5 and in the Big Rock Point Nuclear Power Plant. 6 The tests have demonstrated that the detectors 7 perform successfully. Tests are being continued in ! 8 order to demonstrate detector longevity. At the l 9 present time, the Big Rock Point detectors have 10 accumulated operational experience equivalent to 11 approximately three and one-half years of full , l 12 power operation in the Russellville Nuclear Unit 13 reactor. 14 d. Core Thermal and Hydraulic Design 15 The PSAR as originally submitted contained, in 16 Section 3, an evaluation of the core thermal capa-17 bility in which the heat transfer limits were 18 predicted based on a correlation of experimental 19 DNB (Departure from Nuclear Boiling) data developed ' 20 by The Babcock & Wilcox Company. In order to 21 completely substantiate the BSW correlation additional 39
L These research and development data is necessary). (58 2 requirements are described in the PSAR. 3 Subsequent to submittal of the original PSAR, core 4 thermal performance was also evaluated using the 5 W-3 correlation for predicting DNB. This correla-6 tion is available in the literature and has been used 7 and found acceptable in establishing thermal design 8 limits for other large pressurized water reactors. 9 The thermal evaluation using the W-3 correlation is 10 also presented in the PSAR and its supplements. With 11 the use of this correlation, vessel model flow tests 12 are necessary to substantiate operation of the plant 13 within acceptable thermal limits. Flow testing which 14 demonstrated acceptable flow distribution for the 15 rated power level without internal vent valves in 16 the model has been completed. Flow testing with 17 internal vent valves installed and with open internal 18 vent valves must still be performed. 19 e. Emergency Core Cooling and Internal Vent Valves 20 Analytical evaluation of the effects of blowdown 21 forces on the internals and of the performance of 40
4 1 the internal vent valves installed in the core 2 support shield to insure adequate covering of the
- 3 core by emergency coolant is in progress. A proto-4 type of these valves is being tested to demonstrate (59l 5 their operating characteristics.
6 f. Fuel Failure 7 A study, including testing, is underway to assure 8 that there are no failure mechanisms which might 9 interfere with the ability of the emergency core 1 10 cooling systems to accomplish their objectives. The , 11 results of the work to date demonstrate the ability 12 of the design to accommodate potential fuel failure 13 mechanisms. This work will be continued to assure 14 that fuel rod failures will not significantly affect 15 the ability of the emergency core cooling system to (60) - 16 prevent clad melting. 17 g. Xenon Oscillations 18 The possibility of the occurrence of xenon oscilla-1 19 tions throughout core life is being evaluated. If 20 it is determined that such oscillations may occur, 21 appropriate design-changes to eliminate or control 41
a 1 the oscillations will be incorporated. The 2 design of a means to eliminate or control such 3 oscillations is being carried out in parallel with 4 the studies of the possibility of such oscillations. 5 h. Chemical Spray Additive 6 One of the radiological protection systems of the 7 Russellville Nuclear Unit provides for spraying 8 chemical solutions into cl.e reactor building to 9 remove iodine uader accident conditions. Testing 10 to demonstrate the ability of the chemical sprays 11 to remove and retain iodine effectiv. ly, and to 12 demonstrate solution stability and chemical (62) 13 compatibility with plant materials is in progress. 42
t 1 7. TECHNICAL QUALIFICATIONS 2 7.1 Arkansas Power & Light Company 3 Applicant has over 45 years c 2rience in the design, 4 construction and operation of electric generating plants. 5 Personnel of the Engineering Department of the Company 6 have supervised and made final decisions on the design and 7 construction of its generating plants. It has been the practice 8 of the Company, however, to retain independent engineers to 9 design and manage the construction of its generating plants 10 under the supervision of the Company's engineers. The 11 Production Department, which is a part of the Engineering 12 Department of the Company, operates all of the generating 13 plants with its personnel. 14 On October 1,1968, Applicant operated five steam ! 15 electric generating plants containing a total of 12 units 16 with a net capability of 1,659,000 kilowatts, two hydro-17 electric stations with a capability of 69,000 kilowatts j 18 and diesel generating units with a total capability of 19 6,000 kilowatts, for a total net electric generating 20 capability of 1,734,000 kilowatts. At the present time the
- 2L Company is constructing one additional generating unit, a 43 l
-_ . __ .=_
i I 1 530,000 kilowatt gas-fired unit, which is scheduled to l l 2 be completed in 1969. ) 3 Applicant was one of the founders in 1957 of South- l 4 west Atomic Energy Associates which was created to conduct 5 research in nuclear fuels. In addition to other projects, l 6 SAEA is now one of the participants in the Southwest 7 Experimental Fast Oxide Reactor Facility near Fayetteville, 8 Arkansas, which is expected to begin operations in 9 December 1968. Various officers and employees of Applicant l 10 have actively participated in the activities of SAEA and 11 SEFOR since 1957 as trustees, officers, committee members 12 and observers. Applicant has also been a member of and a 13 contributor to High Temperature Reactor Developnent Associates, 14 Inc. and has participated in the sponsoring of the HTRDA : I 15 operation at Peach Bottom, Pennsylvania. Applicant is also ; 16 a contributor to and participant in Southern Inter-State 17 Nuclear Board and the Atomic Industrial Forum. 1 l 13 Applicant recognizes the importance of the early train-19 ing of sufficient personnel to assure adequate operating 20 manpower, which is the subject of a comment in the ACRS l 21 letter. Applicant will initially train enough employees so ' 22 that there not only will be enough trained employees for 44
,7 l
l l l l 1 regular work on each shift, but also there will be l 2 adequgaly trained personnel to substitute during illness, ! 3 vacations and ther absences. Applicant 's training pro-4 gram for operators is described in full in PSAR, Volume II, 1 l 5 Appendix LA , Section 1. 7. This program will include 600 6 hours of classroom work in nuclear ~ engineering and reactor 7 theory, three to five months of training in operations at 8 an existing plant or on a simulator, about two months 9 instruction on the design characteristics of reactor systems 10 furnished by The Babcock & Wilcox Company and approximately 11 seven months of on-the-job training at the Russellville 1 12 Nuclear Unit. 13 7.2 Bechtel Corporation 14 Bechtel Corporation has been retained by AP6L as 15 Architect / Engineer and Manager of Construction for the 16 Russellville project. l 17 Working closely with AP6L, Bechtel is responsible for l l IP project studies and conceptual design, specification of 19 material and services, project detailed design, construction 20 management, quality control programs and assistance in plant 21 testing and start-up. 1 45 l l l l
l 1 Bechtel Corporation has been continuously engaged in 2 construction or engineering activities since 1898. For the 3 last 20 years, Bechtel has been active in the fields of 4 petroleum, power generation and distribution, harbor develop-5 ment, mining and metallurgy, and chemical and industrial 6 processing. 7 Since the close of World War II, Bechtel has been 8 responsible for the design of over 165 power generating 9 units, representing more than 38 million kilowatts of new 10 generating capacity, which includes units of the largest and 11 most modern types. Of this number, more than 11 million 12 KWe is produced by 20 nuclear-fueled units. 13 For over 18 years, Bechtel has been engaged in the 14 study, design and construction of nuclear installations. 15 Their experience includes design or construction, or both, of 16 such facilities as accelerators, nuclear research laboratories 17 hot cells, experimental reactors and nuclear fuel 18 processing plants, as well as nuclear power plants. A 19 summary of experience is listed in the Application. 20 7.3 Babcock and Wilcox Company 21 BSR's participation in the development of nuclear power 22 dates from the Manhattan Project. B&W's broad nuclear 46
L activities include applied research to develop fundamental 2 data; design and manufacture of nuclear systems, cores, and 3 components ; and design, manufacture, and erection of complete 4 nuclear steam generating systems. Through the B6W Company's 5 several divisions, a wide range of equipment for nuclear 6 application is designed and manufactured. The B&W Company's 7 major nuclear contracts, in addition to manufacture of a 8 substantial percentage of components for the nuclear Navy, 9 have included Indian Point No.1; NS Savannah; Advanced 10 Test Reactor; Oconee Nuclear Station Units 1, 2 and 3: Three 11 Mile Island Nuclear Station; Crystal River Plant Unit 3; 12 and four other units in various stages of licensing in 13 addition to the Russellville Nuclear Unit. 14 8. COMMON DEFENSE AND SECURITY 15 There is no indication that construction and operation 16 of the Russellville Nuclear Unit will in any way be inimical 17 to the common defense and security of the United States. 18 As stated in the Application, AP6L is a private utility 19 with statutory authority for the production, transmission 20 and sale of electric energy. All of the directors and 2 '. principal officers are citizens of the United States, and 22 AP6L is not owned , controlled, or dominated by an alien, a 47
1 l 1 foreign corporation, or a foreign government. 2 The Application contains no restricted or other defense 3' information and Applicant has agreed that it will not permit j 4 any individual to hav? access to Restricted Data until the l 5 Civil Service Commission shall have made an investigation ; l 6 and report to the Atomic Energy Commission on the character, l 7 associations and loyalty of such individual, and the Atomic i 8 Energy Commission shall have determined that permitting such 9 persons to have access to Restricted Data will not endanger 10 the common defense and security. 11 As a licensee, Applicant will be subject to regulations 12 of the Atomic Energy Commission relating to the transfer of 13 and accountability for special nuclear material in its 14 possession. Recent amendments to the AEC Rules and Regula-15 tions (10 CFR 50.60) under which the AEC will discontinue 16 allocating quantities of special nuclear material to 17 reactor licensees evidence that such material is no longer 18 scarce. Moreover, in the event of a state of war or national 19 emergency, the AEC may order the recapture of special nuclear 20 material, as well as the operation of any licensed facility.
.21 (10 CFR 50.103) 48
1 1 9. CONCLUSION l 2 On the basis of the foregoing and the Applicatien, the l 3 Applicant respectfully submits that: 4 a. Arkansas Power & Light Company's Application, as l 5 supplemented, describes the proposed design of the ' 6 Russellville Nuclear Unit, including the principal 7 architectural and engineering criteria for the I i 8 design, and identifies the major features or l l 9 components incorporated in the plant for the pro-10 tection of the health and safety of the public. 11 b. The Application, as amended, identifies the 12 technical and design information necessary to complete 13 the final safety analysis. Such information can 14 reasonably be left for later consideration and will l 15 be supplied in the final safety analysis report. l 16 c. Safety features which require further research and ! 17 development, and the research and development programs 18 to be carried out, are identified in Section 1.5 of 19 the PSAR. The research and development' program is l 20 reasonably designed to resolve any questions 21 associated with such features at or before the s 49
1 latest date stated in the Application for completion 2 of construction of the facility. 3 d. Taking into consideration the characteristics of the 4 site and environs and the proposed design of the 5 Russellville Nuclear Unit, such facility can be 6 constructed and operated within the limitations 7 established by 10 CFR 20, within the site criteria 8 set forth in 10 CFR 100, and without undue risk to 9 the health and safety of the public. 10 e. The Applicant is technically qualified to design and 11 construct the proposed facility. 12 f. The issuance of a construction permit for the , 13 Russellville Nuclear Unit will not be inimical to 14 the common defense and security of the United States , l 15 or to the health and safety of the public. ! 1 l i ~ 50
- x. . _ _ _ _ _ -. .. _ . . - _ . . . - . . . . . . . _ ..-.,.._ - .
l l l l l l l 1 i APPENDIX A l LIST OF REFERENCES l l l l
+
l APPENDIX LIST OF REFERENCES
- 1. PSAR, Volume I, Section 2.2.2
- 2. PSAR, Supplement 3, Questio n 2.3
- 3. PSAR, Volume I, Section 2.2.5
- 4. PSAR, Volume II, Appendix 2A
- 5. PSAR, Volume I, Section 2.4 ,
- 6. PSAR, Volume I, Section 2.5
- 7. PSAR, Volume I, Section 2.6
- 8. PSAR, Volume I, Section 2.7
- 9. PSAR, Volume I, Section 2.8
- 10. PSAR, Supplement 3, Question 2.9
- 11. PSAR, Volume I, Section 1.2.2
- 12. PSAR, Volume I, Table 1-2
- 13. 'PSAR, Volume I, Table 3-2
- 14. PSAR, Volume I, Table 3-1
- 15. PSAR, Volume I, Section 3.2.3
- 16. PSAR, Volume I, Section 3.2.2.1.2 and Section 7.2. 2.1
- 17. PSAR, Volume I, Table 3-6 and Figure 3-1
- 18. PSAR, Volume I, Section 3.2.2.1.3 A-1
~
- 19. PSAR, Volume I, Section 3.2.4.3.2
- 20. PSAR, Volume I, Section 3.2.2.1.3
- 21. PSAR, Volume I, Section 4.2.2.1
- 22. PSAR, Volume I, Section 4.1.4.4
- 23. PSAR, Volume I, Section 4.4.3
- 24. PSAR, Volume I, Section 4.2.5
- 25. PSAR, Volume I, Section 4.2.2.4
- 26. PSAR, Volume I, Section 4.2.2.3
- 27. PSAR, Volume I, Section 4.2.2.2
- 28. PSAR, Volume I, Section 5.9.1.2
- 29. PSAR, Volume I, Section 6
- 30. PSAR, Volume I, Section 6.1
- 31. PSAR, Volume I, Section 6.2
- 32. PSAR, Volume II, Section 14.2.2
- 33. PSAR, Voluue I, Section 7.1
- 34. PSAR, Volume I, Section 3.2.4.3
- 35. PSAR, Volume I, Section 7.4.3
- 36. PSAR, Volume I, Section 7.2.3.2
- 37. PSAR, Volume I, Section 7.3.1
- 38. PSAR, Volume I, Section 7.2.2.2
- 39. PSAR, Volume I, Section 7.1.2.2 and Section 7.1.2.3 A-2
- 40. PSAR, Volume I, Section 9.1
- 41. PSAR, Volune I, Section 9.5
- 42. PSAR, Volume I, Section 9.2
- 43. PSAR, Volume I, Section 9.3
- 44. PSAR, Volume I, Section 9.7
- 45. PSAR, Volume I, Section 9.6
- 46. PSAR, Volume I, Section 9.4
- 47. PSAR, Volume II, Section 14
- 48. PSAR, Volume II, Section 14.1
- 49. PSAR, Volume II, Section 14.2
- 50. PSAR, Volume I, Sec't ion 6.1
- 51. PSAR, Volume I, Section 6.2
- 52. PSAR, Volume II, Section 14.2.2.3
- 53. PSAR, Volume I, Section 8.2.3
- 54. PSAR, Volume I, Section 14.2.2.4
- 55. PSAR, Volume I, Section 1.2.7 and Section 1.5.7 and Supplement 3 Question 1.3a
- 56. PSAR, Volume I, Section 1.5.6 and Supplement 3, Question 1.3b
- 57. PSAR, Volume I, Section 1.5.8 and Supplement 3, Question 1.3c
- 58. PSAR, Volume I, Section 1.5.2 and Supplement 3, Question 1.3d A-3
a .__
- 59. PSAR, Volume I, Section~ 1.5.5 and Supplement 3, Question 1.3e
- 60. PSAR, Volume I, Section 1.5.3 and Supplement 3, Question 1.3(h)
- 61. PSAR, Volume I, Section 1.5.1 and Supplement 3, Question 1.3(i)
- 62. PSAR, Volume I, Section 1.5.10 and Supplement 3, Quess ion 1.3g J
l i A-4 i l
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4S 9 o CUMULATIVE TOTALS i sucTon RADIUS IN o - O l ID EO 30 0 6 0 do S o W ' w N - A C A C A SG. MI. S G. M t. SG.M e A, N 12.5 0 18.4 0 18.4 0 D 9 NNW 15.4 0 56.6 0 55.6 0 9 NW 2.5 0 43.9 0 85.5 0, 4 #g 3 WNW 0 0 17.7 0 71.7 24. W 9.5 0 23.7 0 82.4 9.1 S *s WSW 3.8 0 3.8 0 12.6 0 SW 0.3 0 9.1 0 33.5
- SSW 5.2 0 17.0 0 31.7 4 S 4.3 0 24.9 0 26.9 0 SSE 6.6 0 41.4 0 56.1 0
_ SE 1.5 .2 36.8 4.9 56.4 7. 73,, LAND USE ESE 4.3 0 36.8 14.7 46.6 E 1.5 0 13.3 0 62.3 0 WITHIN A ENE 8.8 0 47.1 3 61.9 3 50 MILE RADIUS NE 8.8 .75 35.3 6.7 35.3 6.: NNE O O 5.9 1.2 5.9 1.! Figure 3
1 I l DAIRY ANIMALS l I WITHIN A 50-MILE RADIUS
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,/ NEWTON 1355 B043 l
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* " pope 1835 PRANKLIN JOHNSON VAN SURE 3454 1EOS 1933 3344
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7 ah,;~ - 9 YELL 37E3 PARRY 42. f I SCOTT T
7" pWLASKI I 1884 I - GARLAND I455 B ALIN E 1434 MONTGOMERY g ess Figure 4
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ESTIMATED TRANSIENT POPULATION C 1967- EDIE 3 TOTAL P O AU L ATIO N 15 CUML Figure 5
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