ML20235B492
| ML20235B492 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Bodega Bay |
| Issue date: | 12/28/1962 |
| From: | PACIFIC GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20234A767 | List:
|
| References | |
| FOIA-85-665 NUDOCS 8709240123 | |
| Download: ML20235B492 (178) | |
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J/ PRELIMINARY HAZARDS
SUMMARY
REPORT
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PACIFIC GAS AND ELECTRIC COMPANY
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, Page I. INTRODUCTION A. Purpose of This Report I-l B. Back6round 1 C. Scope of This Report 1-t II.
SUMMARY
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I A. General II-l B. Unit Size 1 C. Process Flow Diagram 1 D. Unit Layout and Structures 1
. E. Reactor 1 ,
F. Containment 2 G. Safeguard Criteria 2 III. DESCRIPTION OF TEE UNI _T A. General III-l B. Plant Arrangement 1 C. Containment 2 (3 D. Reactor 4 s E. Steam Supply System 8 F. Turbine Plant Systems 11 G. Control and Instrumentation 15 H. Service Syste=s 22 I. Reactor Servicing Equipment 23 J. Shielding and Access Control 24 l K. Electrical Systems 25 IV. REACTOR CHARACTERISTICS A. Introduction IV-1 B. Thermal and Hydraulic Characteristics 1 C. Nuclear Characteristics 5 4 s%g e
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D. E. Analysis of Equipment Malfunction Operator Errors 6 11 F. Fires and Other General Hazards 14 G. Major Accidents 14 H. Radiological Effects of Major Accidents 23 I. Basis for Limitation of MCOA 24 J. Protection Afforded Against Reactor Vessel Failures 26 VIII. OPERATION A. Operating Principles VIII-l B. Organization 2 C. Procedural Safeguards 2 D. Preoperational Testing 2 E. Initial Fuel Loading and Sta.Mup 2 F. Normal Operation 3 - G. Waste Disposal Systems 3 H. Refueling 4 I. Maintenance 4 J. Storage 5
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LIST OF PIGURES FIGURE POLI 4 WING NO. TITLE PACE II-1 Flow Diagram II-1 2 Site Plan II-1 III-1 Preliminary Plan.and Site Arrangement III-1 , 2 Plant Arrangement Section' III-1 3 Plant Arrangement Plan Above and Below Elevation -30' 8" III-1 4 Plant Arrangement Plan Above Elevation -6' 6" and +10' 0" III-1 , 5 Plant Arrangement Plan Above Elevation +13'6" and +25'0" III-1 6 Plant Arrangement Plan Above Elevation +42'0" III-1 7 Plant Arrangement Plan At Elevation +62'3" III-1 8 Containment of Reactor Systems III-2 9 Reactor Pressure Vessel III-4 1 10 Reactor Isometric. III-5
, 11 Puel Assembly III-5 12 control Rod Assembly III-5 13 control Rod Drive III-6 14 Control Rod Drive System, P & I D III-6 15 Reactor, P & I D III-8 16 Peedwater, P & I D III-8 ') 17 18 Clean-Up and Shut-Down Systems, P & I D Turbine, P & I D III-10 III-11 19 Condensate, P & I D III-12 -
20 Reactor Cooling Water, P & I D III-13 l 21 Turbine Cooling Water, P & I D III-13 ! 22 Condenser Cooling Water, P & I D 111-14 23 Legend, P & I D , III-14 24 Functional Diagram, Reactor and Turbine Generator Protection III-18 25 Reactor Protection System III-18 26 Refueling System III-23 27 Puel Handling Plow Sheet III-23 28 Electrical Single Line Diagram III-25 j".- '.) g,. .. ., _ _ .,
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FOLLOWING TITLE PAGE V-1 Map of North Bay Area ' V-1 2 Map of Area Surrounding Site V-1 ( ,3 Topographic Map of Site and Environs V-1
.4' Water Wells in Vicinity of Site V-lO
-5 Aerial Photograph of Site and Environs V-6 VI-l Radwaste Collection System P & I D VI-1 VII-1 Post Accident Containment Pressures -VII-20 l 2 Fuel Temperature Transient VII-20 ,
3 Fission Product Inventory - Maximum Credible Operating . Accident VII-21 4 . Fission ' Product Inventory - Desi6n Basis Refueling Accident , VII-23 5- Stack Discharge Rates VII-23 ' . ~y l l 4 l O e gV ' ', \
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1 APPENDIX I Pressure Suppression Test Program
' APPENDIX II Pressure Suppression Design APPENDIX III Report'on Meteorological Conditions at Bodega Head and Bodega Bay APPENDIX IV Report on Earthquake Hazards at the Bodega Bay Power Plant Site APPENDIX V -
Earthquake Hazards and Earthquake Resistant Design - Bodega Bay Power ' Plant Site ' (. .A i t
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s, V. 1 r-I-1 1 t INTRODUCTION 1 A. PURPOSE OF THIS REPORT This report presents the technical information required by 10 CFR Part 50 to provide reasonable assurance that the proposed nuclear power unit t can be constructed and operated at Bodega Bay Atomic Park without undue ' risk to the health and safety of the public. B. BACKGROUND Bodega Bay Atomic Park is situated on Bodega Head, Sonoma County, on the shore of the Pacific Ocean and approximately 50 miles northwest of the City of San Francisco. The proposed installation will consist of a single-cycle, forced-circulation, boiling water reactor, with a rated - thermal output of 1008 megawatts, supplying steam to a 325 megawatt turbine generator unit. Pacific Gas and Electric Company (the Company) ; will design and supervise construction of the unit. General Electric 1 Company will furnish the nuclear steam supply system and the turbine 8enerator.
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C. SCOPE OF THIS REPORT ' t
'\ l This report describes the site and its environs, gives the principal " - 1 features of the proposed unit, and includes a preliminary analysis of the project from the standpoint of safety, all as required by Section i 50.34 of 10 CFR Part 50.
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~l A. GENERAL The. unit employs a single-cycle, forced-circulation, boiling water reactor .
to produce steam which.is used directly in a steam turbine. q, The' reactor is of the thermal neutron, heterogeneous type, using slightly ' enriched uranium oxide as fuel and ordinary water as moderator and coolant. Steam-leaves the' reactor vessel'at 1075 psia and 553.5'F ,(saturated). It enters the' turbine at a pressure of 1060 psia. After passing through the turbine it is. condensed and demineralized.' The condensate is then returned
.to the reactor. : '~'j B.~ UNIT SIZE The unit has a nominal gross electrical output of 325 Mw and a corresponding -
. reactor thermal-output of 1008 Mwt. (These figures are affected somewhat by the performance of certain auxiliary equipment not yet purchased, and are therefore subject to adjustment.)
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C. PROCESS FLOW DIAGRAM
,, .The process flow diagram is shown in Figure II-1. i F1 D. - UNIT IAYOUT AND STRUCTURES ,'
The Site Plan for the unit is shown in Figure II-2. The major structures are:
- 1. Cooling Water Intake
- 2. Barge Dock
'3. Refueling Building
- 4. Outdoor Turbine Generator
- 5. Ventilating System Stack
- 6. Office and Machine Shop ,
- 7. Warehouse .-l
- 8. Switchyard Control Building *
- 9. Meteorological Tower .
E. REACTOR , The reactor pressure vessel has an inside diameter of 15 ft-1 in, and an j inside length between heads of 50 ft-4 in. It contains the reactor core and l the steam separators and dryer. The core is approximately a right cylinder j with a circumscribed diameter of 147 in. and a height of 125 in. , 4 4'
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l} 1 l' The 145 control rod hydraulic drive mechanisms are mounted on the bottom of the reactor vessel. The cruciform control rods contain boron for neutron 1 absorption. i
'l l The individual fuel rods are 0.421 in. in outside diameter, including a stain- "
less steel clad 11 mile thick. The fuel material is slightly enriched uranium dioxide. Four forced hieculation loops take water from the reactor vessel and return it to the vessel, pumping it through the note. F. CONTAINMENT . Containment for the unit consists of operating containment and refueling con-
'tainment. Ot wating containment is provided by a pressure suppression system; refueling cointcinment by a refueling building.
i The basic elements of the pressure suppression system are (1) a dry well, (2) y a suppression chamber sind pool, and (3) vent pipes. The dry well consists of - a steel hhell, f constructed in accordance wit'n the ASME Boiler and Pressure Vessel Code, watch encloses the reactor vessel, its recirculating pumps and piping, the contrul rod drives, and other associated piping. The suppression chamber is a toroidal steel shell, encased in reinforced concrete, which lies l~'
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' suppression chamber. From this header 112 pipes, 24 in. in diameter, project downward and terminate below the water surface.
G. pgEGUARD CRITERIA 7 I
- 1. Normal Operation l
[' All phases of normal operation of the unit, including disposal of all radioactive materials produced thereby, will be such that no person, whether I on or off the site, will be exposed to radiation in excess of permissible limits.
- 2. Safeguards Against Accidents The design, construction, and operation of the unit will be such that the likelihood of accidental infraction of the normal operation criterion is minimized.
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.g hf DESCRIPTION OF THE UNIT A. GENERAL l
1 Extensive development programs and operating experience with boiling water
; reactors provide a firm basis for the design of the unit. In particular, ,
design and/or operating experience have been obtained from the following '
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.m (1) Vallecitos Boiling Water Reactor, (2) Dresden Nuclear Power Station, (3) The Big Rock Point Plant, and (4) The , Nuclear Unit at Humboldt Bay Power Plant.
The descriptions of structures, equipment, and systems contained in this i report represent the design as it is presently conceived. It is expected that changes will be made during the course of detail design, but that these changes will not represent fundamental departures from the basic design, nor will they substantially alter the hazards analysis or the conclusions drawn therefrom. Such changes will be described in amendments to this application, including the final hazards summarygreport. B. PIANT /JtRANGEMENT l , The plant layout is shown in the following figures: - Figure III-1, Preliminary Plan and Site Arrangement Figure III-2, Plant Arrangement Section , Figure III-3, Plant Arrangement Plan Above and Below Elevation -30' 8" Figure III-4, Plant Arrangement Plan Above Elevation -6' 6" and + 10' 0" l Figure III-5, Plant Arrangement Plan Above Elevation +13' 6" and +25' 0" Figure III-6, Plant Arrangement Plan Above Elevation +42' 0" Figure III-7, Plant Arrangement Plan At Elevation +62', 3" The reactor and its pressure suppression containment system are supported in an unlined reinforced concrete structure, the major portion of which is below ground level (Grade +25-0"). The refueling building above this structure l contains fuel handling equipment and the new and spent fuel storage facilities. The control room is adjacent to the refueling building. The outdoor turbine generator and its auxiliary equipment are served by a gantry crane. An l administration building - which contains offices and laboratories - is l located adjacent to the refueling building. The warehouse, dock, vaste i storage area, and other facilities are located as shown in Figure III-1, l " Preliminary Plan and Site Arrangement." 1 1 l l
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h III-2 '
- C. CONTAINMENT
,1. Introduction-Containment for the unit consists of operating containment and refueling '
containment. Operating containment is provided by a pressure suppression system, which will be sealed and ready to, function whenever the reactor , is in the hot and pressurized condition. Refueling ' ' containment is,provided by a refueling building, which will be effective for fuel handling operations, and for the reactor at times when the pressure suppression containment system is not sealed. The refueling ' building also provides a secondary containment barrier during reactor operation.
- 2. Operating Containment Pressure suppression containment consists of three main elements: (1) a dry well, (2) a suppression chamber partially filled with water, and (3) connecting vent piping. The arrangement of these elements is shown schematically in Figure III-8, " Containment of Reactor Systems." In the event of an operating accident involving a. rupture in the piping within the dry well, steam or a water-steam mixture would flow into the dry well vessel and through the vent pipes into the water pool, where it would be quickly and completely condensed.
q The design of the pressure suppression system will be based on extensive j tests which are being conducted for the specific arrangement proposed for the Bodega Bay unit. The test facilities, program, and results obtained ' to date are described in Appendix I. These results are consistent with
- earlier tests conducted in connection with the design of the similar con-tainment system for the nuclear unit at Humboldt Bay Power Plant.
- a. Dry Well - The steel dry well vessel consists of a spherical lower portion, 60 f t in diameter, and a cylindrical upper portion, 26 f t in diameter. The overall height is approximately 100 ft. The net air volume of the dry well and the connecting vent pipes and vent header is 115,000 cu ft.
The dry well vessel will be designed and constructed in accordance with the ASME Boiler and Pressure Vessel Code, Section VIII, including , app 1*icable code cases for an internal pressure of 62 psig. The shell and heads will be constructed of SA-201 Grade B steel produced to SA-300 specifications. The vessel and its penetrations will be so constructed that the maximum leakage rate will not exceed 0.57. of the volume in 24 hours I
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h LEGEND
" * "' ( " "
CONTAINMENT OF REACTOR R .OmNO,o v^'v5
"#1OR OTEo d'V/
N CHECM VALVE SYSTEMS SAFETY VALVE (SPRING LOADED) i RELiEr VALVE (soLENOio OPEaATEo) H VALVE (NORM. CLOSED, TYP.) FIGURE E-8
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'The dry well' vessel shellIis.ba'cked up by reinforced concrete adequate to resist jet and shock loadings that could accompany the Maximum Credible Operating Accident.
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The' concrete surrounding- the dry well also serves 'as. a biological shield. . Cooling is provided to prevent excessive thermal' stress ,
, concentrations due to gamma heating in the area adjacent to the -{j core. ' ,
The atmosphere in the dry well will be maintained'at an average temperature of 150*F during reactor operation. . The' cooling system consists of blowers and heat exchangers supplied from the plant cooling water system.
- b. Suppression Chamber - The suppression chamber is a. steel pressure vessel in the shape of a torus, with a major diameter of 93 ft and a cross-section diameter of 26'ft. It contains 52,400 cu ft of water and has a net air space volume above the water pool.of 80,200 cu ft. The suppression chamber will.be designed for a pressure of M- ~' "
3! psig in the air space above the pool. It will be designed and constructed in accordance with the ASME Boiler and pressure Vessel Code, Section VIII, including applicable code cases. The vessel and
. its penetrations will be so constructed that the maxt:nu:: leakage rate will not exceed 0.5% of the volume in 24-hours at the design pressure .
.of 35 psig.
- c. Vent Pipe System - Eight 8-ft diaw ter vent pipes connect the dry well to a 5-f t diameter vent header in the form of a torus which encircles the dry well and is contained within the air space of the suppression chamber. From the header 112 pipes, 24-in. in diameter, project downward and terninate 4 f t below the surface of the pool. The design does not nov include baffles between the pipes in the suppression pool. Additional pressure suppression tests are planned to determine whether or not baffles are necessary. l
- d. Core Spray - Core spray is provided to minimize core melt-down and fission product release in event of coolant loss. Water is taken from the suppression pool and delivered by the core spray pumps to spray nozzles above the core.
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- e. Vacuum Breakers - Vacuum breakers.are provided to protect the dry well from possible negative pressure, relative to the suppression pool.
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'> f. Isolation valves -' One isolation valve is provided at the containment V barrier.for each major pipe or duct which penetrates a pressure sup-pression containment barrier.and connects to the nuclear steam supply t i
system or opens into the. containment free space.' These valves.are located as close to the containment wall as practicable without sacrificing accessibility for maintenance. The isolation valves are' j backed up by remotely. operable process valves . (as, for example, the - > l turbine stop and bypass valves).or, if the line.contains no process-valves, by a second isolation valve. Typical valve arrangements are shown schematically in Figure III-8. Where necessary, at least one I valve in each line will close automatically from appropriate isola-tion signals.
- 3. Refueling Containment Refueling containment is.provided by the refueling building end its *
, ventilating system. It is designed to contain, with controlled in- '
leakage, fission products released from the design-basis refueling accident. i
- a. Refueling Building - The refueling. building is'a rectangular reinforced concrete' structure having a maximum permissibic in- l 1eakage rate:of 1007. of its volume per day when maintained at a l negative pressure of 1/4 in. of water. All personnel access and
. equipment openings are provided with interlocked double closures.
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- b. Ventilation System - A controlled-release ventilation system which
- discharges to the stack is provided to maintain the refueling building at negative pressure when containment is required. ]
Fission product cleanup equipment is provided for the removal of halogens and particulate before discharge. Refueling building .; containment provisions, including starting the controlled-release ventilation equipment and closing the normal ventilation system isolation valves, are automatically brought into effect by a " build-ing high radiation" signal from radiation monitoring instruments within the building. D. REACTOR
- 1. Introduction The unit employs a single cycle, forced circulation, boiling water ,
reactor with internal steam separation. The principal reactor design - dataare given in Table III-1. ,
- 2. Reactor Pressure Vessel The reactor vessel, shown in Figure III-9, is a vertical cylindrical I pressure vessel with an inside length between heads of 50 ft 4 in. and an inside diameter of 15 ft 1 in. The base plete material is a high strength alloy carbon steel, SA-302 Grade B. The vessel interior is
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# clad with Type 304 stainless steel applied by veld overlay. The vessel will be designed, built, and tested in accordance with Section VIII
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,"' of the ASHE Boiler'and Pressure Vessel Code, including applicable _ nuclear code cases. ' The design pressure is 1235 psig at 575'F. . Wall thickness ,
is approximately 6 in. of base metal plus k in. of cladding. The vessel will receive a' code stamp. . L The vessel contains and supports the reactor coolant, core structure and fuel, the steam separators and steam dryer, and'other internal components.
.It is supported by a skirt attached to the lower head. Vessel stabilizer brackets are designed to resist earthquake and jet reaction forces, and to accommodate thermal' expansion.
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- 3. Core Geometry and structure-The reactor core, 'shown in Figure III-10, occupies approximately the volume of a right circular cylinder 147 in. in diameter and 125 in. high.
The core contains 592. fuel assemblies, 580 of which are adjacent to control r ods . The 12 assemblies not adjacent to control rods are'at the periphery of the core. The' fuel. assemblies rest on plates supported by the control
. rod guide- tubes.
The 145 control rods enter the Eore from below through the control rod guide tubes. 'These tubes extend from the fuel' support plates to the control rod drive thimbles penetrating the reactor vessel lower head. The _) guide tubes provide guidance and protection for the control rods, and ~ restrict the flow of coolant into the control rod gaps. - The core shroud is supported from blocks attached to the vessel wall and forms the interior wall of the downcomer for the recirculating reactor coolant. The lower end of this downcomer is blocked off and the water is drawn off through four recirculation loops and pumped back into the plenum below the core.
- 4. Fuel-Assembly The fuel assembly, shown in Figure III-11, is made up of 49 fuel rods in a square array (7 x 7), with an active fuel length (cold) of 125 inches. .
- 5. Control Rod .
The control rod design is shown in Figure III-12. The poison section 54 Leh controls the nuclear chain reaction consists of small vertical stainless steel tubes filled with boron carbide powder compacted to 65% to 75% of theoretical density. The tube wall thickness is selected to withstand the maximum internal gas pressure which will develop. These tubes are held in a cruciform arrayby flat stainless steel plates which ,
. extend the full length of the poison section and provide a smooth outside surface.
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1 4 <. III-6 l l e', The boron carbide powder is separated longitudinally into several inde-pendent compartments by stainless steel balls, which are located at approximately 15 in. intervals. A plenum for helium is provided by the 25% to 35% of the tube volume which is not occupied by boron carbide. , The control rods are capable of being positioned in 6 in, steps to con-trol flux distribution in the reactor core. These rods normally are moved one at a time at a rate of approximately 3 in. per second. They all serve a dual function as safety rods.
- 6. Control Rod Drive System
- a. General - The reactor control rod drive system positions the control rods to regulate reactor power level and power distribution within the core. Each of the lh5 control rods is positioned hydraulically by its own drive system.
The hydraulic fluid (reactor feedwater) for normal operation is supplied by one of two positive displacement pumps. These pumps take suction from the condensate system at low pressure and lov temperature and discharge into the hydraulic system filters. The water required for drive cooling and for operating the drives is reduced.through a pressure reducing valve to approximately 200 psi above reactor pressure (Pr + 200). This hydraulic pressure provides the energy for positioning the drives during normal rod movement. - Water for cooling the drives is further reduced to (Pr + 30). Energy for fast insertion of all rods is supplied by pressure ' accumulators, one accumulator per drive, or by reactor pressure. Both of these sources are independent of the hydraulic system for - normal rod movement.
- b. Basic Design - The control rod drives are basically similar to the Humboldt locking piston-type drives. An assembly drawing of the locking piston drive mechanism is shown in Figure III-13 and a P&I Diagram of the system is shown in Figure III-14. The control rod drive mechanism is an integral unit which is installed in a thimble. The thitble is velded to the reactor vessel. The drive mounting flange attaches to a mating flange on the thimble.
The drive mechanisms are capable of raising or lowering the control rods at the normal controlled rate, as well as providing rapid insertion if required. All drives are capable of positioning the control elements in 6 in. increments of the stroke to achieve control and optimum flux distribution in the reactor core. Each drive vill hold its control rod in a fixed position until actuated for movement. , l
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TABLE III-1 III-7 REACTOR DESIGN DATA
- 1. Operating Conditions Reacto'r power, Hwt 1008 Gross electrical power (nominal), Hw 325 Reactor steam pressure, psia 1075 Reactor steam temperature, Si' 553.5 Feedwater temperature, 'F 402
- 2. Reactor Pressure Vessel Design pressure, psig 1235 .
Hydrostatic test pressure, psig 1853 l Design temperature, 'F 575 l Reactor vessel inside diameter, ft-in. 15-1 2eactor vessel inside length, ft-in. 50-4 Base material, shell plate SA-302B Clad material Type 304 Minimum clad thickness, in. E approx
- 3. Fuel and Core Assembly Fuel material -
UO2 Initial enrichment of first core, 7. U-235 2.7 Weight of uranfum in core, Ib 148,000 g.s Fuel pellet diameter, in 0.398 - g') Cladding thickness (stainless steel), in 0.011 Fuel rod outside diameter, in 0.421 Fuel rod active length (cold), in 125 Number of fuel rods per assembly 49 Number of fuel assemblies 592 Moderator to fuel volume ratio 2.7 Overall length of fuel assembly, in. 150
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Weight of fuel assembly, ib. 350 Circumscribed core diameter, in. 147 Channel wall thickness, in. .060 Channel material Zr Channe1 width (inside), it.
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- 4. Control Rods Number 145 Poison material boron carbide Pitch (square array), in. 10.0 Structural material 304 SS Active length, in. 124 Shape cruciform Width, in'. 6.94 Thickness, in. .312 Stroke length, in. 125 Velocity (normal) in./sec 3.0
"~' Fast insertion time, (for 807, travel) see 2'.5 max
- 5. Control Curtains Number in initial loading 316 Material 0.1% Boron SS Length, in. 125 Width, in. 9 Thickness, in. 0.10
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w. g! L III-8 E. STEAM SUPPLY SYSTEM , The steam supply system includes the reactor and its auxiliaries, reactor !- service systems,-and the feedwater system. Figures III-15 and III-16 show the ; E reactor and the feedwater system, respectively. All pressure vessels and that ', portion of the piping which is inside the-dry well comply with applicable requirements of the ASME Boiler and Pressure Vessel Code. Piping outside the dry well complies'with;the. requirements of the ASA Code for Pressure Piping. ,. In general, stainless- steel is used for all surfaces which are in contact with . hot reactor water during operation. The main steam lines and the feedwater system are carbon steel. > 1.. Reactor Flow -l ( The arrangement of the reactor internals is shown in Figure III-10. ] During operation reactor heat is transferred to slightly subcooled water { entering the bottom of the core. Boiling produces a steam-water mixture ' j which ' increases in steam quality and velocity as it flows upward through the core. Above;the core, bolted to the core shroud, is a double tube sheet assembly that supports the axial flow steam separators'. Centrifugal .j force drives the water to the outside walls of the separators where it i drains to the downcomer annulus. From the downcomer annulus water passes to four recirculation loops and is pumped back to the plenum below the
- core. {
g} '2. S, team Separation ', Steam separation takes place entirely within the reactor vessel. In e ! addition to the axial flow centrifugal separators, a steam dryer and ' l dry box assembly are provided. This assembly, located near the top head of the vessel, provides final separation of steam and water and channels the dry steam to the outlet nozzles. The dryer assembly consists of
.I closely packed stainless steel wire mesh. Entrained water impinges on the stainless mesh in the dryer, drains to gutters below the dryer, and ]
passes through pipes to a point below the reactor vessel water level. j Steam leaves the dryer with a moisture content of 0.17. by weight.
- 3. Reactor Recirculation Inops .
The recirculation system consists of four loops,'each containing a high ' capacity motor-driven recirculating pump. These units take suction from , the. bottom of the downcomer annulus, and return to the inlet side of the core. Appropriate baffling in the reactor vessel prevents mixing of inlet and outlet flows. The pumps are vertical centrifugal units with mechanical shaft seals, rated 29,000 gpm at 100 ft TDH. Valves are
.provided in each recirculation loop to permit isolating a pump when necessary. *
- 4. Reactor Safety Valves 3 The reactor safety valves are sized to prevent the reactor pressure from s/ exceeding the allowable over-pressure for the vessel. There are twelve b
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- valves, mounted on the circular main steam header in the dry well and arranged to discharge through pipes which lead to the suppression chamber and terminate below the level of the pool.
- 5. Turbine steam Bypass Valves The turbine steam bypass system is designed to limit reactor pressure rise ,
so that the reactor safety valves will not lift in the event of a full l load turbine trip. It is not designed to prevent a reactor scram in such l case, however. It consists of eight solenoid operated steam unloading ' valves, mounted on the two 20 in, main steam lines ahead of the turbine stop valves. The bypass valves discharge through a desuperheater to the main condenser. The bypass system will accommodate approximately 40% of 1 rated steam flow. The turbine bypass valves open automatically from a high pressure signal. A hydraulically operated bypass valve is similarly arranged for use in
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j regulating pressure during startup and shutdown of the unit.
- 6. Reactor Feedwater System ,
l The reactor feedwater system consists of the reactor feed pump, the reactor ) start-up pump, and the feedwater heaters and their drain coolers, as shown in Figure III-16. Heater No. 4, the lowest pressure heater, and its drain i
'} cooler are located in the main condenser neck. This heater and drain cooler each consist of two shells operating in parallel. The other heaters are located in the heater compartment adjacent to the turbine.
The reactor feed pump is driven by the main turbine shaft through a step-up gear drive at 3633 rpm. A smaller motor-driven start-up pump is also , provided.
- 7. Liquid Poison System The liquid poison system consists of a storage tank, positive displacement I pumps, valves, and piping. It can be used to introduce a liquid poison solution into the reactor core and provides a back-up means of shutting j down the reactor in the event that complete shut down cannot be achieved by use of the control rod system. The liquid poison (a solution of sodium
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- 8. Reactor Cleanup System The reactor cicanup system, shown in Figure III-17, is designed to remove corrosion products and possible fission products to maintain reactor water purity at a solids concentration of 0.5 ppm or lower. Reactor water
. is taken from a recirculation line and pumped through the tube side of a regenerative heat exchanger, a non-regenerative heat exchanger, filters, and the cleanup demineralized. Purified water is returned to the reactor through the shell side of the regenerative heat exchanger. This continuous cleanup of reactor water is accomplished with a minimum of heat loss and with no water loss from the cycle during normal operation.
Spent resins from the cleanup demineralized are not normally regenerated, but are sluiced to the radwaste system resin disposal tank for storage pending ultimate off-site disposal. *
- 9. Reactor Shut-down Cooling System ,
The reactor shut-down cooling system is also shown in Figure III-17. It ] consists of the shut-down pumps and heat exchangers, with a heat removal ! capacity of 60 x 106 Btu per hr. Design pressure is 150 psig. This { system provides for decay heat removal from the reactor water system during l shutdown. The reactor is initially cooled by controlled reactor steam flow l
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to the main condenser through the turbine bypass line. After initial
# cooling, the shut-down system is placed in operation to further reduce reactor water temperature to 125'F. This system is kept in service during l refueling operations and at such.other times as the reactor is shut down with irradiated fuel in the core and with the reactor vessel head removed or in process of removal.
- 10. Core Spray System The core spray system supplies cooling water to the reactor core spray I sparger. Two full-capacity pumps are provided. These take suction from the suppression pool, through the suppression pool coolers, as shown in Figure II-1. These circuits are independent, each consisting of cooler and F2mp. Two spray inlet . nozzles are provided on the reactor vessel.
When reactor pressure is below 150 psig, the core spray system will be started automatically under conditions of reactor low water level, and can also be started manually.
- 11. Isolation Cooling System The isolation cooling system provid.ts the means for removing heat from the reactor in the event of loss of the main condenser as a heat sink.
A dual system is provided. Heat can be vented to the suppression pool and/or to the emergency condenser. Two solenoid operated steam dump valves,one a
- spare, actuated by high reactor pressure or by operator hand control, dis-s charge to the suppression pool. Reactor feedwater will replace the steam and
.) restore normal reactor water level. The emergency condenser consists of a tank of water, vented to atmosphere, and containing two tube bundles connected to the reactor. The inlet to the tube bundles is connected to reactor steam.
The outlet line returns to the reactor water. The emergency condenser is s .,. 3 . , q. . , . , _.- . , . , , _ _ _ , . . _ .
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6A F. TURBINE PMNT SYSTEMS The turbine generator unit is installed on.a concrete pedestal which is supported by a continuous mat foundation. The7 foundation also supports the main / condenser. -Turbine generator auxiliary equipment and systems, the condensate system, and the condensate domineraliser are located in concrete t vaults under or adjacent to the pedestal. Thsre'is no turbine-building in ' the usual sense. The entire turbine plant area is serviced by a 125. ton gentry crane. , 1
- 1. Turbine Generator '
The turbine generator. unit operates at 1800 rpm. The 325 Nw turbine is a tandem compound, double flow machine with 43 inch last stage buckets, designed for steam conditions of 1035 psia, saturated '(0.257, moisture due to throttling in the line from the reactor). It drives a 384,000 kva,- 0.85 power factor. 0.64 short circuit ratio,J.2,000 volt conductor- , cooled generator, rated at 30 psig hydrogen pressure. The generator capability at 45 psig hydrogen pressure is 42't,000 kva, 0.582 short ) circuit ratio.- The exciter is driven through reduction gaars from the i generator. shaft. The turbica also drives the reactor feed pump, ' located D' at the head and of the unit.' ~ ,, n J s f] The guaranteed rated flow to'the turbine is 4,170,000 lb per hr at inlet 'f steam conditions given above'. The design flow is 4,540,000 l'a per hr. 1 The connections to'the turbine are shown in Figure III-18. - t The turbine, including itsiconttols and accessory equipment, is designed ' to operate on steam from a boiling water reactor. .Particular attention has been given in its design to the elimination of pockets or crevices in which radioactive material night lodge. Drains are provided for.each I turbine stage. The turbine is provided with moisture removal buckets ahead of each extraction point. iIn addition, meisture separators are installed in the two cross-arounds between the high pressure and low pressure cylinders. Materials used in the construction of the turbine are selected to minimize the wear caused by wet oxygenated steam.
- 2. Condenser i
The main condenser is designed to perform the following functions: 1 (1) Condense vacuum. steam exhausted from the turbine to obtain the desired (2) Demerate the condensate, heater drains, and other returns.
' (3) Serve as a heat mink for excess reactor steam vMch may be dumped through the bypass system.
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- (4) Holdup condensate in the hotwell for decay of short-lived radioactivity.
The condenser.is a single" pass unit with 162,000 sq ft of surface. It is - constructed of standard materials-for salt water servic,e, with a fabricated steel shell and vertically divided water boxes. ~The condenser is located beneath- the low pressure cylinder of the turbine, and its tubes are per-pendicular to the turbine center line. Feedwater Heater No. 4 and its ,
; drain cooler are mounted in the connecting piece between the turbine -
exhaust and the condenser tubes. - To provide greater assurance that circulating water will not leak into and-contaminate the condensate, the tubes are welded to the tube sheet, in addition to the normal metal-to-metal rolled joint. The condenser also has provision for isolation of any condensate that might be contaminated ' with circulating water in the event of a leak adjacent ba a tube sheet. The tube sheet material is silicon bronze and the tube material is aluminum brass (1" 0.D..x 18 BWG). The hot well capacity is 27,000 gallons. It is constructed with division plates to provide a 3 minute condensate holdup. The condensate leaving the hotwell is guaranteed to contain not more than 0.01 cc of dissolved oxygen.per liter. A twin element, two stage, steam jet air ejector unit with inter- and af ter-condenser is provided. Each element is capable of removing 52.2 73 scfm of dry ' air equivalent. The air ejector discharges to the plant ventilating stack through oversize piping which provides about 30 min l holdup capacity.
- 3. Condensate System The condensate system is shown in Figure III-19. It consists of the condensate pumps and the condensate demineralized. -
Three half-capacity vertical condensate pumps are provided. These are located outdoors, adjacent to the condensate demineralized vault. Normally, two of the pumps will operate in parallel taking suction from the condenser hotwell. Steam for heating feedwater is supplied from extraction openings in the turbine. The turbine is also provided with moisture drains at several stages and the flows from these drains are combined with the main ex-traction flows from a lower stage before admission to the heaters. Turbine extraction non-returnsy'alves are furnished in each heater extraction line. These valves are arranged'rsselose on a turbine overspeed trip and on heater high water level. _)
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,1 F
Six mixed-bed ion exchangers are provided for removal of impurities- '
- in the condensate.- Five of these units will operate while the sixth is being regenerated or is on standby. . -
The resin regeneration system consists of cation and anion resin - regeneration tanks and a regenerated resin storage tank. Spent resins are hydraulically sluiced from the demineralizers to the regeneration system, where-they are separated, individually regenerated, and stored. . Spent resins are monitored for radiation level af ter removal from the ion exchangers, and before classification and regeneration. If ,the resin is found to contain a concentration of radioactive material so high as to lead to uneconomic handling and disposal of the regeneration waste solution, the resin itself is sluiced directly to radwaste. Resin of low activity level is regenerated and reused. Because of the potentially high radiation levels associated with this equipment, ion exchangers and resin regeneration tanks are located within shielded compartments and are arranged for remote operation.
- 4. Cooling Water Systems The cooling water. systems supply treated cooling water to the following :
reactor and turbine plant equipment, as shown in Figure III-20 and Figure III-21, respectively: 4-7 ', Reactor Plant (Figure III-20) ' s Reactor Shutdown Heat Exchangers and Pumps Reactor Shield Coolers Dry Well Air Coolers Fuel Pit Coolers Suppression Pool Coolers l Reactor Cleanup Non-Regenerative Heat Exchanger Waste' Collection Tank Cooler Radwaste Vapor Condenser Reactor Cleanup Pump Reactor Enclosure Drain Tank Cooler Reactor Recirculation Pumps , Lube Oil Coolers l Turbine Plant (Figure III-21) Generator Hydrogen Coolers Generator Arnature Heat Exchanger Generator Seal Oil Coolers { Turbine Lube Oil Coolers
'{
Reactor Feed Pump Coolers ,j Reactor Start-up Pump Coolers '
, Instrument Air Compressors Service Air Compressors
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( Cooling water is pumped through the reactor cooling water heat.exchangers and the turbine c,ooling water heat exchangers, thence to the separate . ' cooling water systems. Circulating water is pumped through the-tube side of these' heat exchangers by the' auxiliary circulating water pumps and is discharged to.the main discharge tunnel. - t
- 5. Circulating Water Syste; 2 Condenser cooling water is drawn from Campbell Cove in Bodega Harbor as
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' .l traveling screens. . The condenser cooling water is pumped by two half' '
capacity, vertical, wet-pit pumps. The circulating water from the condenser and from the cooling water heat exchangers is discharged to the Pacific Ocean.
- 6. Instrument and Service Air Systems Compressed air will be provided for the instrument and service air l systems. These systems have not been' sized, but will consist of air compressors, receivers, and, in the case of the instrument air system, cleaning and drying equipment.
- 7. Makarup Water System. '
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f i: III-15 i; O G.;' CONTROL AND INSTRUMENTATION
'.1. Introduction :
The control and instrumentation systems serve to enable the control operator to start up, operate, and shut down the reactor, turbine gener- .
' ator, and necessary auxiliary equipment, and to protect this equipment.
The instruments related to vital plant functions are located on consoles and panels in the control room.
- 2. Control Room The control room, . located adjacent to the refueling building as shown in Figure III-7, contains all the controls and instrumentation essential for safe operation of the plant. All controls, indicators, and recorders ,
needed by the operator for startup, shutdown, and operation are located on or are readily visible from the console. The reactor board, the turbine generator board, and switchgear. panels contain instrumentation which must frequently be referred to and which therefore must be reasonably accessible to the operat.or. 3 power Control ' The reactor control rod settings are selected by the operator, and these settings determine the reactor power level and power distribution.
.,.s, The operator adjusts the settings as necessary to compensate for fuel U) ,
burnup and changes in xenon poisoning.
- The turbine control includes a conventional speed governor and an initial pressure regulator. During normal operation' the turbine admission valves are controlled by the initial pressure regulator. -
The turbine speed governor is set at a speed higher than system frequency and is not in control of the unit. The steam bypass valves are normally closed, and all reactor steam flow is through the turbine. With the power control system set up in this manner, the turbine follows the reactor output rather than responding to system demand. However, the necessary protection against turbine overspeed is main-tained.
- 4. Reactor Control Reactor power level is regulated by control rods which are positioned by the operator in the control room. By means of switches mounted on.the control console, a control rod is selected for movement, and that rod is inserted or withdrawn one position. Continuous rod move-ment, in or out, can be achieved by simultaneouscperation of the'
- rod control'svitch and a one-position override switch.
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t-f r III-16 ' j l 1 L Only'one rod can be moved at a time for normal control. All rods can be x) inserted simultaneously for a prompt reactor shutdown. , To bring the reactor to power from the cold condition, reactivity is increased by rod withdrawals until the reactor becomes critical. The-rods are then further withdravn unti1~ a period of about 30 seconds.or longer is obtained. This period is maintained until the proper heating rate is established. ' Water level in the reactor vessel is controlled by a three-element level control system. This system uses the measurements of steam flow, feed-water flow, and water level. Normally, the steam flow signal equals that'of water flow. 5 Reactor Instrumentation
- a. Control Rod Position Indicating System - An indication of the '
position of each control rod is provided on the reactor board. In addition, "all-in" and "all-out" pilot lights furnish an indication that travel limits have been reached. These devices are grouped together and arranged on the board to correspond to the rod positions in the core.
- b. Reactor Neutron Monitor System - Instruments are provided to monitor *
-the neutron flux . level of the reactor from startup through full ^
,..s power. - The instr.tmentation covers a range of nine decades in three f
~J- phases: (1) Startup Range, (2) Period or Intermediate Range, and (3) Flux Level or Power Range.
(1) Startup Range - In the initial fuel loading a neutron source is . inserted in the reactor core to assure a count rate of several neutrons per second. The startup instrumentation covers the range upward to about 107 counts per minute. The primary neutron detectors are proportional counters housed in guide tubes in the concrete dry well vall opposite the core. Neutron leakage through the vessel vall interacts vith the boron-lined chamber of the proportional counter. The resulting ionization of the contained gas gives a series of pulsen proportional to the reactor neutron count level. The average rate of this ' series of pulses is measured on a count rate meter with a logarithmic scale encompassing six decades of measurement. The count rate is recorded continuously. An additional circuit differentiates the log count rate, and indicates the reactor period at this low level on a period meter. Short period at this lov level is annunciated only. ,
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III-17 , (2) Period Range - This instrumentation is used to determine the i , rate at which the neutron flux is increasing, and guides the operator in adjusting the control rods as the reactor power l level increases toward fun power. The primary detectors are l gamma compensated ion chambers housed in guide tubes in the dry well vall opposite the reactor, similar to those for the start-up chambers. The output from each chamber is passed to a Log N amplifier, which indicates the logarithm of the input on a scale covering from 10-7 to full power. The output of the Icg i N amplifier is. continuously recorded. A time derivative of the output is also indicated on a period meter. A varning vill be { indicated visually and audibly if any channel should approach a short reactor period. Reactor scram is initiated if a short period is indicated by tvo out of three period monitors. (3) Power Range - When the reactor reaches about 1% of full power, boiling begins and the steam voids cause the period measure-ments to fluctuate. In order to avoid spurious scrams, the period channels are then disabled from the safety circuits. An overlap in coverage of two decades below this point is provided by the power range instruments which begin to function and monitor the reactor to full power. The primary detectors are gamma compensated chambers positioned in guide tubes at the midplane of the reactor and located in the dry well vall. The output is fed into a power range flux 3 ' amplifier of fast response and stable characteristics. The , i power level is read on a meter indicating percent of full power, and is continuously recorded. A varning vill be annunciated if { any of the six flux level indicators should exceed a preset ' percentage of over-flux. A scram signal to insert control rods occurs whenever two or more of the six power level indicators ) show an excess of neutron flux. l I
- c. In-Core Flux Monitors - A number of miniature ion chambers are I installed in thimbles in the core. These in-core flux monitors provide accurate measurement of the axial and radial flux distri-l bution, and thereby serve as a guide to the operator in selecting j control rod patterns for maximum operating efficiency and proper power distribution. The in-core thimbles extend from the core through the bottom head of the reactor vessel. The electrical leads are collected at a terminal board external to the vessel. The out- ,
put from these chambers is amplified and indicated on an integral display on the reactor panel which also includes control rod position ! indication. ' l i These chambers vill be calibrated during operation by means of flux vires which are inserted in the chamber thimbles. After exposure
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for a pre-determined time interval, the vires are removed and the induced activity is counted, providing a power level reference for i setting the in-core indicators. t s-go q e , m esy -sos tery M a A* D et w * **v'q**4*- . ~
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- d. Reactor Temperature - The reactor vessel outside vall metal temper- !
'O ature vill be measured in a number of places to be determined. These temperatures vill be recorded on the process control panel.
- e. Reactor Pressure - Pressure vill be measured at the pressure vessel and transmitted electrically to the control room. At the panel, the ;
transmitted signal vill be recorded and indicated.
- f. Recirculation Flov - Differential pressure vill be measured across each of the recirculating pumps. Also, a flow nozzle is installed in the discharge leg of each recirculation loop. These signals vill be transmitted to indicators in the control room.
- g. Reactor Water Level - Liquid level in the reactor vessel vill be measured continuously by means of externally-mounted differential pressure-type sensing devices. The reactor water level vill be recorded, and indicated in the control room. High and lov level vill be annunciated.
- 6. Reactor Protection System The function of the reactor protection system, which is shown schemati-cally in Figure III-24, is to protect equipment and personnel by scramming the ree.ctor in the event of a potentially unsafe trend or condition. Control signals which initiate scrams originate from a p number of control and detection devices. These signals activate control circuits to cause the reactor to be immediately shut down .
and to initiate operstion of containment penetration closures, emergency cooling, and other measures which may be necessary for safe plant shutdown. The reactor protection system is a two-channel, " fail-safe" design, as shown in Figure III-25 Both channels must be de-energized to ; produce a reactor scram or other safety system function. Whenever j practical, the two channels of each sub-system are physically ' separated and clearly identified to minimize the possibility of maintenance personnel causing an accidental shutdown, l l
- a. Scram Conditions - Preliminary analysis of the reactor system indicates th at the following abnormal conditions should initiate a scram:
(1) Short Reactor Period - This prevents the reactor power from l increasing too rapidly, thus anticipating a high flux condition. This scram signal vill be automatically bypassed when the reactor reaches about 1% of rated power. l
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r III-19' I, Dj (2) -High Neutron Flux - This limits the instantaneous core power to a level well below that which could cause fuel damage. (3) HighReactorPressure-Thislimitsthefriseincorepower,due
.to pressure rise to a safe value. Since the pressure rise
. ingplies an interruption in steam flow, emergency actions for
.,- core cooling are initiated.
i - (4) High Dry Well Pressure ~- Abnormal rise:in pressure could indi- [ cate a rupture of the primary system within the dry well, or: equivalent. valve or pump leakage, severe enough to drain the- .
. taken.-
water from the reactor vessel'if no corrective action were Since this-could expose the core, and the possibility i
.of release'of. radioactivity exists, isolation of the dry well and. emergency cooling are initiated.
E (5) Low condenser vacuum - This assures that the reactor will not be operated without its main heat sink. Continued operation
- could cause escape of radioactive steam from burst condenser diaphragma. = This scram signal is bypassed when reactor pres-
.sure. is below about 300 psig. . ;
(6) Low Reactor Water Level-~- This assures that the reactor will. not be operated without sufficient water above the core. (7) Scram Dump Tank High Level - This assures that the reactor g~ vill not be operated with insufficient free volume in the s drive ' system scram dump tank, which would prevent rapid-insertion.of the rods. , i j 7 Process Radiation Monitoring ' l
- a. Air Ejector Off-Gas' Monitor - Non-condensible gases from the. main condenser are monitored to determine the gross activity being discharged to the stack. If the. gross radioactivity should become too high, the off-gas valve would be closed to prevent environs contamination. Contacts are provided for alarm at lower activity levels. ,
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- c. Liquid Process' Monitors - The following liquid processes are '
monitored continuously and recorded at regular intervals to determine radioactivity icvele'and trends: .,
'(1) ' Radunste Discharge . ' ;
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'(2) Cooling' Water System t ,,
(3) Circulating Water Leaving cooling Water-Heat Exchangers
- d. Other Cas Procccs Monitors - In addition to the air ejector off-gas, i the following are monitored: <-
l 1 (1) The ' gross gamma activity in the vent stack of. the emergency condenser ic continuously monitored to detcet breakthrough of the radioactive. steam to the atmocphere in the event of tube failure during operation. (2) Suppression chamber vent line. (3) The stack gas is continuously monitored to assure that the ' limits set by the facility licence are not creceded.
- 8. Area and Health Monitoring L/ ~7y 'Camma-sensitivo detcetors arc installed throughout the plant, each
. - arranged to monitor a specific area. The gamma flux 1cyc1 at cach, '
detcetor is indicated continuously in the control rocm.'. Adjustabic d
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;,, trips provide annunciation in the event of a high level at any detector.
3 A multi-point recorder provides a permanent record of the gamma flux levels. Area monitors in the refueling building vill initiate closure of refueling . building isolation valves.and starting of the controlled-release venti- , lation system upon indication of a high radiation level in the building. ;
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Two of the fixed position area gamma monitors are situated to provide i monitoring of the fuel stcrage area, in compliance with Paragraph ! 70.24 of 10 CFR 70. Portable radiation monitors are provided for use in area monitoring for personnel protection and contamination control. 9 Primary Containment Dry well pressure and temperature are indicated in the control room. High dry well temperature is annunciated. The suppression chamber pressure and pool vater level are indicated in the control room. High or low chamber pressure, high pool temperature, and high or low pool vater level are annunciated.
- 10. Turb'ine Plant instrumentation ,
- a. Turbine Control System - Control and supervisory equipment for the turbine generator are conventional and are arranged for remote operation from the turbine generator board and adjacent switchgear panels. Turbine lube oil pressure and steam extraction pressures are transmitted to receivers on the board'
- b. Condensate System Control - Main condenser controls include a low vacuum trip arranged to close the main steam line isolation valves.
Level control of demineralized water make-up to the condenser and condensate rejection from the condensate pump discharge downstream of the condensate demineralized- are also provided. The condenser , hotwell level and the condenser vacuum are recorded on the turbine generator board. Abnormally high or low hotvell level and low condenser vacuum are annunciated.
, Conductivity of the condensate gathered from near each tube sheet,
, and conductivity of the ecndensate from each half of the condenser hotvell are meast. red to detect condenser cooling water leakage.
- c. Feedvater Heater Control - Controls for the feedwater heaters con-sist of level controls on each heater shell for normally controlled cascade of 6. rains, and for high level dump to the condenser.
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- d. Make-Up Water Control - The water levels in the makeup water storage tank, the demineralized water storage tank, and the condensate .
storage tank are indicated locally, with abnormal high or n
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(, ,s , , - W ; III-22 Ti . r /N low level' annunciated. 'High level in the demineralized water ' storage tank also trips closed the demineralized water supply. High level'in the makeup water. storage tank trips the sea water evaporator-steam supply valve. Both of these systems are arranged for, manual start. 11.- Turbine: Protection Devices
- If the. turbine should overspeed due to sudden loss of electrical load,.
the speed governor signal vill over-ride the' initial pressure regulator .
. signal which normally controls the turbine admission valves. The'
- admission valves will- then close sufficiently to maintain satisfactory turbine speed. This closure vill cause tlie reactor pressure to rise.- The turbine bypass valves then are opened by means of their pressure controller and dump steam to the condenser.
A turbine emergency overspeed trip is also provided. Other conventional turbine protective' devices, each of which vill trip
~the stop valves, include a low vacuum trip, thrust bearing failure trip, and generator protective relay trip. '
- 12. Meteorological Facility
. A tower is provided with instruments 'to record vind speed, vind direction, and temperature gradient. This facility is described in 9.$y..
b Section V-E. 13 _ Environs Monitoring - During operation of the plant, an environs monitoring program vill be carried out on a regularly scheduled basis. This vill consist of a continuation, in part, of the program described in Section V-H, modified as required to suit the needs of plant operation. H. SERVICE SYSTEMS 4
- 1. . Fire Protection, .
The fire protection system vill be designed and constructed'in accordance with the Company's fire protection practices for generating stations. The following facilities vill be furnished:
- a. Water - Water at 100 psig vill be furnished to all points throughout the plant area where water for fighting fires may be required. The
, normal fire pump supply is well water from a storage tank with salt water from the bay as an emergency source. The following equipment
' vill be used for this system.
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'one '- 750 GPM engine-driven stand by fire pump
. two . ;100 GPM booster pumpo ub.'; CO2 Syst&m - A low pressure carbon dioxide system is provided for protection of the bearings of the turbine and generator, the '
f .,
. interior'of the turbine main lube oil tank and the lube oil filter.
J . . c. Fog System - A fixed fog system is provided for the area under the turbine front standard, the hydrogen seal oil equipment area, the
' lube. oil. reservoir, and the lube oil filter area. If these' areas have time-limited access, a detector controlled automatic fog system vill be provided.'. *
'd.
-Portable Equipment - Portable fire protection equipment v111 be
' located strategically throughout 'the plant. -
, 2. Heating and Ventilation A preliminary arrangement of the heating and. ventilation system is shown in Figure II-1.
Q of I. ' REACTOR SERVICING EQUIPMENT 1. Fuel Handling
. The ma.}or elements of the refueling system are shown in' Figure III-26.
In general, fuel handling is accomplished by manual guidance and visual observation of all operations. Water is used as the shielding material. The fuel remains in the water during its transfer from the reactor to the fuel storage pool. Figure III-27 outlines the various steps in fuel handling.
- 2. Spent Fuel Storage Pool The spent fuel storage pool, located in the refueling building as shown in Figures III-2 and III-7,contains racks to assure storage of spent fuel assemblies in a non-critical configuration. The pool is equipped with a filter, and coolers. It also contains the channel stripping machine for removing irradiated channels from the fuel assemblies.
1 3 New Fuel Storage
. . A vault is provided in the refueling building for storage of new fuel prior to loading it into the reactor. Storage racks are arranged to prevent accidental criticality. ,
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hf6QfurqQaMp;$il@p:$$Qf.digeabhWEup.%MMai46se46:;g,1;g.Jgg,,q.g ; g , - . O; , III-24 J. SHIELDING AND ACCESS CONTROL Radiation shielding design and access control procedures are specified to limit the maximum exposure of plant personnel to 5 rems per year. The target weekly maximum exposure exposure is 100 mrems, with the intention that the normal veekly vill be vell below this f16ure, reserving the major part of the permissible exposure for off-standard conditions. Radiation zones vill be established and access control procedures and design vill be such that, the foregoing exposures are not exceeded. Shielding consists primarily of ordinary concrete of sufficient thick-ness to limit radiation levels to the following values: Zone 1. (Areas where access need not be controlled.) Less than 0 5 mrem per hour. Such areas include the control room, the administrative offices, and the site except in the immediate vicinity of the unit. Zone 2. (Areas where continuous occupancy may be required on occasion.) Less than 1 5 mrem per hour. Such areas include the operating floor of the refueling building, outdoor areas immediately adjacent to the unit, and some equipment compartments. Zone 3 (Areas requiring periodic entry for sampling, inspection, maintenance,etc.) Less than 15 mrem per hour. Zone h. (Infrequently entered areas. ) Greater than 15 mrem per hour. Access control procedures vill be based on the following criteria: 1. Personnel shall enter radiation zones 3 and 4 only with the knowledge of the shift supervisor or the control room operators.
- 2. Personnel entering radiation zones 3 and 4 must be treined concerning the hazards therein, access control procedures, and in the use of radiation survey instruments, or must be accompanied by personnel so trained.
- 3 Special access control procedures involving use of step-off pads and monitors will be established as needed to prevent the spread of radioactivity from contaminated areas.
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i Access control design vill follow normal practice for a nuclear power y-h :
, installation. High radiation zones vill be blocked by chain barricades or locked doors. Security fencing in the plant area vill'be provided.- *
- - K. ELECTRICAL SYSTEMS
- 1. ' General The' electrical systems- for the unit are shown in Figure III-28,
" Electrical Single Line Diagram". :In general they are very similar ,
to those used for conventional steampower plants of comparable size.
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The electrical energy produced in the main generator at 22,000 volts is fed over isolated phasu bus.to the Company system and to the
' plant auxiliary systems .
'2. Transmission The voltage level for the system energy is stepped up from 22,000 volts to 220,000 volts.by means of the main step-up transformer, <
rated 384,000 KVA, and fed through a circuit breaker to the 220,000 volt bus. .Two 220,000 volt circuits deliver the energy to the P.G.& E. system at Ignacio Substation some 32 miles away. These same two ' circuits vill provide energy to the plant when the turbine ,'
- m. generator .is out of service and also serve as standby power source. ;)
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- 3. . Auxiliary Power An isolated phase bus connection at'22,00d volts feeds energy to the
' unit auxiliary transformer which steps down from 22,000 volts to 4,160 volts to supply the indoor metal-clad 4 KV attxiliary bus.
This bus is in two separate sections. Either or both of these bus sections'can be fed from the Unit Auxiliary Transformer mentioned above or the Standby-Startup Transformer which is supplied from the ! 220,000 volt bus previously mentioned. Therefore, the plant auxiliaries system may at any time receive its supply from the generator, I from the system or from both. Under normal operating conditions the plant auxiliaries vill be fed from the generator and vill be - { automatically transferred to the system upon failure of the source. < Startup energy vill be supplied from the system. i l J
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- 4. Standby Power Sources \! s The main Company system, through the 'startup-standby transformer, serves as the main standby power source. If the normal source to the 4 KV auxiliary buses should fail,the startup-standby transformer source would automatically be switched to feed.the station auxiliar-
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auxiliaries required for a safe plant shutdown. In addition to these sources,a substantial station battery will provide ~ standby energy for necessary 'd.c. control and instrumentation , loads and several emergency pumps. An inverting system will supply a.c.'from this battery source for essential a.c. loads. I e 9
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,I , REACTOR CHARACTERISTICS j
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n IV-1 REACTOR CHARACTERISTICS T ,i l A.' INTRODUCTION The Bodega Bay reactor is a light water moderated and cooled, forced circula-tion, direct cycle, boiling water unit with internal mechanical steam separa- l tion. Its nominal thermal output is 1008 Mw at a pressure of 1075 psia and
- with feedwater entering at 402'F. The core will be loaded with 592 fuel .
4 assemblies, each consisting of,49 cylindrical rods containing slightly enriched ~ j UO 2 . The fuel assemblies will be designed for an average irradiation of 15,000 j Mwd / ton of uranium. l l This Section describes the thermal, hydraulic, and nuclear properties of the reactor, including the principal design criteria. i B. THERMAL AND HYDRAULIC CHARACTERISTICS Coolant enters the bottom cf the core and flows vertically upward over the fuel, discharging through the steam separators into the reactor vessel upper , plenum. Provision for core flow radial distribution control will be made by orificing the entrance region to any or all of the fuel assemblies.
- 1. Power Distribution .
The maximum heat flux for which the fuel must be thermally designed is (~) '# defined by the average operating conditions at rated reactor power and the design power peaking factors (see Table IV-1). These peaking factors identify the peak to average power conditions for gross power distribution, local power perturbation, and overpower conditions, and are intended to define the power distribution for design purposes only. The power dis- , tribution achieved during normal operation is expected to be appreciably better.
- 2. ' Fuel Cladding Integrity The pre;ently anticipated fuel cladding material isstainless steel, although the remarks concerning cladding integrity and thermal design are applicable to any cladding material thac might be selected. The fuel will be designed to operate without loss of cladding integrity over the design exposure period at the maximum heat fluxes possible within burn-out limitations.
- 3. Burnout The reactor system will be designed to maintain a minimum burnout ratio of 1.5 or greater. Minimum burnout ratio is defined as the minimum ratio of burnout heat flux to actual operating heat flux at any point in
, the core, wherever it may occur. It is evaluated at the design power . peaking factors noted in Table IV-1.
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3 TABLE IV-1 .
.i REACTOR THERMAL AND HYDRAULIC DESIGN DATA General Operating conditions.
1. Reactor Thermal Output, Mw 1008 " Reactor Pressure, psia 1075 - l Total Steam Flow, ib/hr 4.17 x 10 6 Total Core Coolant Flow, ib/hr 43.5 x 106 Feedwater Temperature, 'F 402
- 2. Core Description Number o:f Fuel Assemblies 592 Number of Movable: Control Blades 145 Number of Fixed Control Curtains- '
316 Moderator to Fuel Ratio 2.7 Active Fuel Length, in. 125 Equivalent Core Diameter, in. 137 Circumscribed Core Diameter, in. 147 Fower Density, Kw/ liter 33
- 3. Fuel Description g '
Number of Fuel Rods per Assembly 49 Fuel Material UO Cladding Material StkinlessSteel Cladding OD, in. 0.421 Cladding Thickness, in. 0.011 Weight of U in core, Ib 148,000
- 4. Control System Number of Movable Control Blades 145 Shape of Movable Control Blades Cruciform Control Blade Fitch, in. 10.0 Control Material in Movable Control Bcc Number of control Curtains 316 Shape of Curtains Flat Sheets Curtain Material 0.1% Boron SS Location of Curtains Flat side between ends of control blades
- 5. Design Power Peaking Factors Gross 2.40 Local 1.30 .
Overpower 1.20
..s Total 3.75
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Heat Transfer Area, sq ft 33,300
. Average Heat Flux, Btu/ hr-ft 2 103,300 .
Minimum Burnout Ratio .-: 1.5
- 7. Hydraulics Total Coolant Flow, Ib/hr 43.5 x 10 6 Steam flow, Ib/hr 4.17 x 10 6 Cora Exit-Quality (average), % 9.6 . 1 s
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L li IV-5 C. NUCLEAR CHARACTERISTICS ' ' The Bodega Bay reactor is a light water moderated, low enrichment, uranium oxide fuel reactor. At operating conditions the moderator is permitted to boil, producing a spatially variable density of steam voids within the core. < The use of light water moderator produces a neutron energy spectrum such that the majority of fissions are produced by thermal neutrons. The presence of U-238 in the low enrichment uraniam oxide fuel leads to the production of significant quantities of plutonium during core operation. This plutonium contributes both to fuel reactivity and power production of the reactor. In addiedon, direct fissioning of U-238 by fast neutrons yields approximatelysevenpercent of the total power and contributes to an increased fraction of delayed neutrons in the cere. The U-238, coupled with the high temperatures in the uranium oxide fuel, also contributes a strong negative Doppler coefficient of reactivity which improves the inherent safety of the reactor. The excess reactivity of the core is controlled by means of 145 bottom entry cruciform control rods containing compacted baron ~ carbide as a nuclear poison. In the initial core loading, control augmentation is provided by means of borated control curtains located between fuel elements. These control curtains will be removed as necessary throughout the lifetime of the initial core to maintain sufficient reactivity for rated power operation. The total control _ system is designed with enough shutdown capacity so that the reactor will i always be suberitical with any one control rod completely withdrawn from the Core. I The fuellattice is heterogeneous, consisting of fuel assemblies, each con- ) taining 49 stainless steel clad, uranium oxide fuel rods. These assemblies are separated by water gaps in which the control rods move and in which the concrc1 curtains are located. Local power peaking caused by these water gaps is corrected in each fuel assembly by means of lowered enrichment in selected fuel rods.
- 1. Reactivity and Control The initial core fuel loading is enriched to approximately 2.7 per cent l U-235. This enrichment provides sufficient excess reactivity to overcome i the effects of temperature and void changes within the core and permit fuel burnup through the first fuel cycle. Control of the initial core is accomplished as shown in Table IV-2. ,
l TABLE IV-2 CONTROL TABLE keff, Clean Core Uncontrolled at 68'F 1.27 Akeff, Control Rods . 0.13 6 k egg Borated Curtains 0.12
^3' ke gg, Fully Controlled 0.97 kerf, Worst Rod Out 0.99
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- 2. Reactivity Coefficients In order that steam formation in a boiling water reactor be self- (
regulating, the core lattice must be so designed that the core becomes under-moderated as moderator is heated and voids are formed. However, if the reactor is greatly under-moderated, the negative reactivity effect of the voids will produce a feedback which may contribute to system instabilities during reactor operation. The design moderator-to-fuel ratio in the Bodega reactor is optimized between these two limits. From the standpoint of core safety, the virtually instantaneous negative fuel temperature or Doppler coefficient terminates or regulates fast core transients. - A summary of approximate reactivity coefficients for the reactor in its initial operating condition is presented in Table IV-3. TABLE IV-3 APPROXIMATE REACTIVITY COEFFICIENTS Coli Hot Operating Moderator Temp. Coeff., (Ak/k/*F) -5 x 10-5 -22 x 10-5 . Moderator Void Coeff. (4 k/k/%, void) -
- 1 x 10-3 -2.3 ;. 10-3
.y Fuel Temp. (Doppler) Coeff. ( A k/k/*F) -1.5x 10-5 -1.0 x 10-5 -1.2 x 10-5
- 3. Irradiation Dependence of Core Characteristics During operation, the composition of the core undergoes a continuous change due to the depletion of U-235, the production of transuranium elements in the fuel, and the burnout of temporary poison in the .
control curtainsy The net reactivity change is such that at any time the reactor can be shut down to the cold condition with any one control rod remaining fully withdrawn from the core. The moderator reactivity coefficients will tend to become slightly less negative with fuel depletion primarily due to the reduced control rod density requirements in the core. The fuel Doppler coefficient, however, will remain virtually unchanged. During fuel burnup, control rods are used to counteract the power dis- I torting effect of steam voids. Taken together, the control rod and void distributions may be used to flatten gross reactor power beyond that possible in a non-boiling core. Sufficient flexibility in gross power distribution is allowed for in the design so that fuel burnup and isotopic composition can be varied spatially in the core to the extent necessary to counteract voids at the end of a fuel cycle when few or no control rods remain in the core. ,
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- 4. Fuel Cycle t L
The reactor is designed to be refueled on a progressive partial batch
- schedule. Although the detailed refueling plan has not yet been specified,
' it is expected that a plan similar to the following will be used: ,
The initial core will'be depleted to approximately 8,000 Hwd/T,at which time .most of.the control curtains will be removed. Core operation will- , then continue until the e~nd of the first cycle'is reached. At the end of the first cycle it is anticipated that approximately 207. of the fuel will be removed in a " scattered""piittern from the core and replaced by fresh fuel. Use of temporary control curtains may be required for the first few such fuel cycles, until equilibrium cycle conditions are reached. ,7. O J. t i 4 e t 4 O
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l r V-1 7- PLANT SITE AND ENVIRONMENT
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A. GENERAL DESCRIPTION
- 1. Location The site is located on the coast in Sonoma County, California, approx-imately 20 miles from the city of Santa Rosa, the county seat, and 50 miles northwest of the city of San Francisco. The site consists of 225 acres at the south end of Bodega Head, a peninsula bounded on the east by Bodega Bay and Bodega Harbor, and on the south and west by the Pacific Ocean. The location is shown in the following maps:
Figure V-1, " Map of North Bay Area" V-2, ' hap of Area Surrounding Site" V-3, " Topographic Map of Site and Environs"
- 2. Access 4
- a. Highway - The site is accessible from the town of Bodega Bay by a road which generally follows the north and west shoreline of the
, harbor. Presently this road is in poor conditio.n and a new access road will be constructed by the Company and deeded to Sonoma County.
The town of Bodega Bay is on State Sign Route 1, a paved two-lane
- I I highway which generally follows the coast of California. The principal north-south route in the vicinity, Highway U.S. 101, passes approx-imately 20 miles to the east of the site.
- b. Barge Deck - A barge dock will be built on the harbor side of the Company 's property. It will be used to receive heavy equipment during 1 the construction period, and subsequently will be available for ship- !
ment of irradiated fuel. The harbor is protected and has adequate depth to accommodate barges suitable for such shipments. ) l
- c. Railroad - The site is not served by rail. The nearest access, the main line of the Northwestern Pacific Railroad, is approximately 20 miles to the east. It is not planned to bring a spur onto the site.
- 3. Site Improvements 1 I
l The location of the plant on the site is shown in Figure II-2, " Site Plan". l This drawing shows the principal plant facilities, including the reactor, turbine generator, transformer, switchyard, administration building, refuel-l ing building, warehouse, condenser cooling water system, dock stack, l meteorological tower, and access and service roads. The transmission line is shown crossing the harbor entrance and proceeding eastward along the outer shore of Doran Park.
- 5. Distance to Nearby Features
~~. a. Property Line - The distance from the reactor to the nearest point on
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- b. Bodega Harbor - The shortest distance from the reactor to Bodega Harbor will be 450 ft; the shortest distance to the centerline of the entrance channel is 750 ft. -
- c. Doran Park - The distance from the reactor to the nearest point on '
Doran Park is 1300 ft.
- d. Nearest Residence - The distance from the reactor to the nearest ;
occupied house is approximately 1% miles. This house is located in a small community known locally as Keesport, north of the site on the harbor side of Bodega Head. '
- e. Nearest Public Road - The access road which will be built generally d in the tidelands adjacent to Bodega Harbor will terminate in a parking area approximately 300 ft from the reactor. This will be a public county road.
- 5. Principal Activities on the Site The principal activities on the site will be those related to the generation of electric power.
B. POPULATION AND ACTIVITIES NEAR THE SITE
--)
- 1. Within 1% Miles of the Site At the present time there are no permanent residences within 1 miles of -
the site. The northerly portion of the company's innd on Bodega Head is presently used for grazing sheep and cattle. Approximately 320 acres of
- 1and adjacent to the site on the north portion of Bodega Head is being acquired by the University of California for use as a field station for marine biology and other scientific studies.
4 The southern tip of Salmon Creek Beach - which forms part of the Sonoma Coast State Park - lies within 1 miles. Doran Park (known also as Doran Beach) is a low-lying sand spit presently used for camping and picnicing only. Nearly all of Doran Park is owned by Sonoma County, and the county has an option to acquire the remainder. It is then planned to designate this area as a County Park, restricted to picnicing and other daytime uses. A Coast Guard Rescue Station has been authorized for construction on * ' Doran Park approximately one-half mile from the unit. County-owned tidelands surround Bodega Head, these lands having been ' conveyed to the county by the state subject to the former making sub-stantial improvements thereto. The county has agreed to lease to the Company for a term of 50 years approximately 4 acres of tidelands lying in { Campbell Cove. The barge dock and certain other plant auxiliary structures j will be located on this land. The lease will also cover approximately 6.66 j acres crossing the harbor entrance, which will provide an easement for the i overhead transmission lines I
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' area and public' access will be permitted, subject only to such restrictions ,
as may be necessary to safeguard public health and safety. jl In addition, it is planned that certain areas on Bodega Head which are
' owned by the Company and which have scenic or historic value will be made accessible to the public at some future date following initial plant ,
operation when it is determined by the Company and by the various agencies concerned that it is safe.and proper to do so.
- 2. .Within 5 Miles of the Site Approximately 500 persons reside within 5 miles of the site. About 350 of these are in the town of Bodega Bay and environs. Commercial and sport fishing are principal activities. Cattle and sheep are grazed on the grasslands in the hills inland and along the coast. Sonoma coast State ' <
. Park includes most of the beaches north of the site.
^
' 3..-Within 25 Miles of the Site Much of the population of Sonoma County resides in Cotati Valley. The
, princi'al cities are Santa Rosa (21 miles east northeast) Petaluma (24 miles east), Healdsburg (23. miles north northeast), and Sebastopol (15 miles east northeast). This valley produces diversified agricultural products, including eggs, poultry, beef cattle, dairy products, grapes, f 'h . prunes, apples, hops, and pears. There is relatively little industrial-
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activity in the area. The Russian River resort area is about 10 to 15 miles north and northeast of the site. Table V-1 lists the population density for various distances from the site, and. Table V-2 shows the population and location of the communities within 25 miles. I I
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' Site - Miles Square Miles 1960 Census Persons /Souare Mile 0 - 1 1.4 0 0 1' - 5 23.6 500 21 ,
O' - 5 25.0 500 20 5 - 10 102.0 1,600 16 0 - 10 12!.0 2,100 17 10 - 15 194.0 15,700 81 0 -
- 15. 321.0 17,800 55 15 -
20 309.0 30,100 97 0' - 20 630.0 47,900 76
'?3 20- - 25 370.0 66,700 180 0 -
25 1000.0 114,600 115 NOTES: (1) The area shown is land area only. The population density within a radius of 25 miles from the site, including ocean area as well as land, is 58 persons per square mile. (2) Population figures are based on the 1960 Census. For Sonoma County these were furnished by' Enumeration District by the Sonoma County Planning Department. ' Parts of Marin County also lie within 25 miles. The Marin County popula-tion. distribution by Census County Division is included in the table. i s ~~,, S k
i 1 0-1 j t V-5 TABLE V-2 POPULATION AND LOCATION OF COMMUNITIES NEAR SITE - Direction Distance Community (1) From Site From Site Population (2) (0 to 1.5 miles) t (None) (1.5 to 5 miles) Bodega Bay NNE 2 350 (5 to 10 miles) Bodega ENE 5.5 100 Occidental NE 9 200 Camp Meeker NE 10 100 (10 to 15 miles) Jenner NNW 10 150
,_., Monte Rio NNE 11 900 1
Guernewood Park N 13 350 Guerneville NNE 14 900 Graton NE 14 1055 Forestville' NE 15 300 Sebastopol ENE 15 2694 (15 to 25 miles) Rio Nido NNE 15 100 Cazadero N 16 230 Cunningham ENE 16 100 Mirabel Park NE 16 150 Cotati E 19 1852 Fulton NE 20 250 Santa Rosa ENE 21 :131,027 Penngrove E 22- 530 Healdsburg NNE 23 4816 Petaluma E 24 14,035 (1) Communities having a population of 100 or more. . (2) Based on 1960 Census i '. d
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- c. i s-l V-6 I C. TOPOGRAPHY Bodega Head is a rocky peninsula projecting southward from the coast into the Pacific Ocean. To the east it partially forms and protects Bodega Harbor. To the north it drope away to a dunes area which leads to the mainland. The Head has three principal hills, 'the most northerly being the highest, with an elevation of 266 feet. The plant is situated in a saddle between the other two hills. Except for the hills, the Head some-what resembles a mesa or tableland, bordered by steep cliffs which drop away to the ocean or harbor from an average elevation of approximately 100 feet. The cliffs are broken by a few coves, some containing small beaches. Figure V-5 is an aerial photograph of the site and' environs.
. The coastal regione of Sonoma County consist of barren rolling hills, rising to a maximum elevation of about 1,000 feet within 10 miles of the site.
None of these offsite hills attains a height of more than 200 feet within 2 miles of the site. To the north and further inland the terrain becomes more mountainous. reachinga maximum elevation of 4,3% reet at the summit of Mt. St. Hel' : >, 35 miles away. D. GEOLOGY AND SEISh0 LOGY Much of California is subject to seismic activity. Therefore the geology and seismology of the site and bordering areas were thoroughly in- * ,,, vestigated before' the Companydecided tolocate a power plant on Bodega Head. T The Company retained a number of consultants to study and evaluate the area for possible sites for power plant purposes, and to establish design criteria. A preliminary geological reconnaissance of Bodega Head was conducted by Clark E. McHuron, Consulting Engineering Geologist, in 1958, to recommend suitable power plant sites on Bodega Head. When the Company had acquired the property at the south end of the Head, it retained Dames and Moore, Soil Mechanics Engineers, to conduct a geophysical seismic survey of the selected site and a preliminary subsurface exploration. In 1960, Drs. Don Tocher, Seismologist, and William Quaide, Geologist, both at' that - :.. time with the University of California, were retained to make a detailed study of the selected site from the standpoint of seismology and geology. Professor George Housner, of the California Institute of Technology, was retained to interpret the studies of Tocher and Quaide and to reco= send structural design criteria for the plant.. The report by Tocher and Quaide, " Report on Earthquake Hazards at the Bodega Bay Power Plant Site, Pacific Gas and Electric Company", is included in Appendix IV, and a summary of their findings as given in the conclusions of this report is as follows:
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'[) . l. . The proposed site lies a short distance vest of the vestern !
margin of the San Andreas fault zone.
- 2. No ev.1dence was found of any major faults under the site.
Although the evidence is not conclusive, it suggests strongly that the trend of the saddle vestward across the Head from Campbell Cove is not controlled by faulting. (This opinion
- has been further strengthened by subsequent field investi- -
gations by Tocher and Quaide in the company of Company engineers.) 3 The quartz-diorite underlying the site vill provide a much better foundation than any other geologic formation on Bodega Head. (The presence of quartz-diorite has been positively identified by. detailed exploratory borings. )
- 4. The quartz-diorite shows evidence of old minor faults; however, l there have been'no movements on these minor faults in the I past few thousand years. Lack of recent movements strongly implies, but does not g'tarantee that there vill be no movements throughout the life expectancy of a power plant.
5 The region near Bodega Bay is remarkably free of small to ) moderate size earthquakes. The region is subject to infrequent major earthquakes generated by large fault movements from the San Andreas fault zone.
- 6. All power plant buildings and appurtenant structures should be designed to resist an ea:.thquake of a modified Mercalli ' f l
intensity of VIII or, to provide a margin of safety, of l intensity IX. The quartz-diorite should provide a foundation ; as good as any rock formation that can be found in the Bodega l Bay region with respect to the hazards of ground shaking in 1 earthquakes. ' 7 Slopes cut into the Pleistocene sands and silts around the site should be kept lov and well-drained to eliminate any danger of j mass movements in the event of a large earthquake. (Provision I for this requirement is being made in specifications for site ! preparation in consultation with the Company's Soils Mechanics l Consultants, Dames and Moore.)
~
A study of th'e earthquake hazards and recommendations covering design procedures have been made by Professor Housner. He concludes that the j Bodega Bay power plant can be satisfactorily designed to resist earthquake j ground motion and recommendsthat the design be based on a Magnitude 8.2 i Richter seismic disturbance. His report, " Earthquake Hazards and Earth-quake Resistant Design, Bodega Bay Power Plant Site, Pacific Gas and Electric Company", is included in Appendix V. 1 i e __-.pyn,
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] Based on Professor Housner's recommended design criteria, and with his approval, the Company has established design criteria for the three classes of structures he has defined. In general, the usual methods of earthquake resistant design will be followed except that the lateral force factor vill vary with the natural period of the structure under consideration, based on the earthquake spectrum of the El Centro May 18, 1940 earthquake. "
The specific seismic design factors for the three classes of structures are as follows: Class 1 Structures - (Buildings and equipment vital to the operation of the plant, damage to which during an earthquake might lead to the uncontrolled release of radioactive material.) Included in this class are the reactor, its internals, its supports, the reactor recirculation ~1 oops, certain instrumentation, certain safety devices, the emergencyc661fhgsystems, the containment system, and the fuel storage system. For Class 1 structures having a period of 1/10 second, or less, a lateral seismic factor of 33% g vill be used with nor=al allevable working stresses. (Class 1 structures having a period greater than 1/10secondvillbesubjecttodynamicanalysisasdescribedabove and in Appendix V.) ** Class 2 Structures - (Structures essential to operation of the plant, O' but whose failure vould not entail any radioactivity hazard. ) Included . in this class are the turbine generator, the feedvater system, etc. For Class 2 structures having a period of 1/10 second, or less, a lateral seismic factor of 20% g vill be used with normal allevable working stresses. Iftheperiodisgreaterthan1/10seconda dynamic analysis vill be made. Class 3 Structures-(Strt$ctureswhicharenotessentialtothe operation of the plant and whose damage or failure vould be an inconvenience but vould not necessitate a shutdown of the plant.) ' Included in this class are the office building, varehouse, cranes, etc. Class 3 structures vill be designed with a lateral seismic factor of 20% g, and with a 1/3 increase in allovable working stresses for the seismic loading.
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,n, E. !ETEOROLOGY ,
The climate at Bodega Bay is typical of the northern and central coastal . area of California and ig. satisfactorily described by meteorological data _' which are available for nearby locations. Such data are not yet available f, for the site itself. A preliminary meteorological study based on limited available data is included in Appendix III, and can be summarized as follows:
- 1. Precipitation Bodega Bay, like the rest of Cal'1fornia's co'ast,has two distinct climatic regimes: a vet season extending from November through March, and a dry season from May through September. The average ,
annual precipitation at Bodega Bay is estimated to be 35 inches. Annual precipitation is expected to range from a lov of 15 inches , for dry years to a high of 80 inches in the vettest years. Approximately 80% of the total precipitation normally falls during the vet season.
- 2. Temperature o
The proximity of the Pacific Ocean gives Bodega Bay a uniform temperature with an estimated annual range of monthly average ', _ , temperature of about 7*F. The record high temperature at Point ,,
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is 98'F; the record lov is 27'F. 3 Wind During the dry season the vind movement is from sea to land approximately 52% of the time and from land to sea 23%. During the vet season the movement is from sea to land 36% of the time and from land to sea 35%. The remaining percentages in both - seasons account for general movements when the vind velocity is 3 mph or less, and no direction has been recorded. North-vesterly vinds (including NNW and WNW) occur 29% of the time , during the wet season and 50% during the dry season. 2+ . Diffusion Climatology An initial estimate of the diffusion pattern at Bodega Bay is provided by a special tabulation (Table 13, Appendix III). The data to be developed by the meteorological program vill provide additional information..
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', 5 Meteorological Program ,
1 A meteorological facility is being installed at the site to i provide necessary data for diffusion studies. This facility l consists of a 250 foot tover with instruments mounted at the 7 foot, 50 foot and 250 foot levels. The location of the meteorological tower is shown in Figure II-2, " Site Plan". The instruments vill measure temperature and vind speed and j direction. All readings vill be digitized and recorded on l paper tape. i F. HYDROLOGY '
- l. Surface Waters k l
Surface drainage of those plant areas which are subject to possible radioactive contamination vill be to catch basins, which may; be pumped to radvaste. Drainage from other plant areas and from the remainder of the site vill be to Bodega Harbor or to the Pacific Ocean. The site is on a peninsula with the ocean on one side and the harbor channel on the other. -
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The stream nearest to the site is Salmon Creek, which discharges to the Pacific Ocean about 3 miles north of the site. The mouth of the Russian River is about 10 miles north of the site. [} 2. Ground Waters - { Ground water supplies all local needs. The presence of subsurface fresh water on Bodega Head (as described under Paragraph 3, '
- 4 Water Supply, following) indicates that subsurface flov from the hills to the north and east of Bodega Harbor is suff!.cient to -
prevent salt water intrusion, and would also be sufficient to ; { prevent any water originating in the plant area from contaminating l these ground waters. 1 Tests vill be conducted during the construction period to determine the ground water gradients and flow rates in the plant area. Present indications are that this water flows either to Bodega Bay or to the Pacific Ocean. 3 Water Supply The Bodega Bay Public Utility District was formed to supply 4 vater to the town of Bodega Bay. Water is obtained from wells ! located on Salmon Creek, as shown in Figure V-4, about four i miles north of the site. This water is filtered and treated, pumped to two 75,000 gallon tanks located about one half mile
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# north ,of town, then distributed through a system of mains to the i community and along the highway to about one mile south of town.
The remainder of the premises around the harbor and bay are supplied by individual vells and small community systems. One of the latter serves a group of residences located on. Bodega Head. at a place known locally as Keesport, just over l{ miles north , j of the site. This water is obtained from wells adjacent to these residences. Fresh water for the plant systems vill be obtained by distillation I 1 of seawater or demineralization of domestic' vater. . Fresh water f for domestic use (showers, toilets,.etc.) vill be obtained from ' the B.B.P.U.D. or from vells which may be developed on Bodega Head.
- 4. Flooding The finish grade elevation for the plant was established at plus 25 0 feet and.is well above the highest tide or storm vave level estimated by the U. S. Corps of Engineers.
! G. OCEANOGRAPHY AND MARINE BIOLOGY Condenser cooling vater vill be drawn from Bodega Harbor and discharged to the Pacific Ocean as shown in Figure II-2. The flow rate vill be approxi-
^t mately 250,000 gallons per minute, and the water temperature vill be l
increased about 18*F vhen the unit is operated at rated load. This heating ' conforms to the Company's standards for conventional steam power plants and has been found to be not injurious to marine life passing through or exposed to the cooling water system. Treated liquid vastes vill be released under controlled conditions to the condenser cooling vater from time to time. Some of these vastes vill contain radioactive materials, with radioactivity levels consistent with applicable state and federal regulations covering offsite discharge.
- 1. Oceanographic Studies The capacity of the ocean to diffuse the condenser cooling vater and minimize the effects of temperature and radioactivity on the '
marine biota has been investigated by a series of tests conducted at the site under the direction of Dr. Ernest Salo, of Humboldt State College, the Company's consultant on Oceanography and Marine Biology. Uranine dye and driftpole tests show that diffusion takes place very rapidly at the proposed location of the discharge, due to its being in an area of great turbulence and mixing. These tests vill continue through at least one annual cycle of oceanographic and meteorological conditions.
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- 2. Marine Biology Survey A preliminary survey of the ecology of Bodega Head, Bodega Harbor, and vicinity has been conducted under the direction of Dr. Salo.
The area was found to be characterized by three ecological zones or types.
- a. Bodega Harbor - Bodega Harbor is a rather typical protected shallow-vater harbor with its characteristic flora and fauna.
It has extensive tide flats containing rich invertebrate fauna. ' Of particular interest are the horseneck or gaper clams,. , the Washington clams, the soft shell clam, and to a lesser extent the cockle beds'. .
- b. Outer Coast - The outer coast from the tip of Bodega Head to
'] Mussel Point is typical of the outer California coast with its steep cliffs, rocky shoreline, and a few semi-protected coves or inlets. The flora and fauna of the outer coast is that typically associated with surf action and its accompanying sea spray.
- c. Nearchore Waters - The nearshore veters of the Head can be considered as a nursery and spawning ground for such fishes as ling cod, greenling, red and black rock fishes, flounders and soles. In season, migratory fish such as the silver and chinook salmon inhabit these waters. Some abalone fishing
, takes place in and about the rocky coves.
Bodega Harbor is a port of landing of various marine products of commerce, such as salmon, the edible crab, shrimp, ling cod, rock fish, sole, and other bottom fishes. In addition, sport fisher- ' men use private boats and chartered boats for ocean fishing. The ecological survey is continuing. This includes preparation of check lists for two stations in the Harbor and four stations on the outer coast.
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. H. RADIOLOGICAL SURVEY A' preoperational radiological monitoring survey of the site and 'its environs i vill' be initiated.tvo years before commencement 'of operation of the 4 reactor. The purpose of this program is to obtain information concerning naturally occurring radioactivity in the vicinity of the site to aid in ' l ascertaining the effectiveness of.vaste discharge controls and procedures '
in protecting the public from any possible radioactive contamination l resulting from operation' of the reactor. '* The details of this sampling program have not been completed for Bodega i Bay. However, it is anticipated that it vill be similar to that ' conducted for the Company's Humboldt Bay nuclear unit. This includes quarterly sampling of the marine waters, plankton, bottom sediments, invertebrates, l shell fish, resident. fishes, and of the intertidal algae and eel grass; l quarterly sampling of the soil, vegetation, local agricultural products, j
.well water, stream vater, and stream mud; veekly sampling of air particu-lates and air background; and continuous sampling of atmospheric fallout. l l
The samples vill be taken from representative locations in the environs, j I q ...
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, VI-1 VASTE DISPOSAL
('S , A. RADIOACTIVE LIQUID WASTES
- 1. General The radioactive liquid waste system (radwaste system) collects, treats, stores,'and disposes of radioactive liquid wastes of.ift .a. .:. .
plant. origin. Radwaste is collected in sumps and drain tanks and is then pumped to the radwaste facility on the north side of the refueling building for treatment, storage, and disposal.' Wastes are handled on a batch basis following sampling and analysis. Disposition'may consist of return to the condensate system, long term storage awaiting disposal off-site, or disposal - to the discharge tunnel which flows into the Pacific Ocean. There is no disposal to the ground.
- 2. Collection of Radwaste '
Radwaste is collected at the following sumps and tanks as shown in Figure VI-1: (1) Turbine Area Drain Tank This 6000 gallon tank is located in the refueling building at Grade -3'- 3". It receives floor and equipment drains from the turbine generator i and feedwater heater areas, and from the upper levels of the refueling () ' building. The tank is divided internally to permit collecting floor and equipment drains in separate compartments. Each compartment is pumped to the radwaste facility. (2) Dry Well Sump This is a sump in the floor of the dry well. It collects drains from equipment contained in the dry well and is pumped to the reactor equipment drain tank. (3) Reactor Equipment Drain Tank i This 6000 gallon tank is located beneath the dry well at Grade -57. It receives drains pumped from the dry well sump as well as floor drains, l equipment drains, and drains from the main steam line. This tank is pumped to radwaste. i (4) Laundry Drain Tank i This 500 gallon tank is located beneath and receives wastes from the ' laundry in the administration building. This tank is pumped to radwaste.
- 3. Radwaste Facility
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The radwaste facility is located outside of and adjacent to the north wall : i of the refueling buildbag. It includes the following tanks:
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VI-2 5 s (1) One concentrated waste storage tank - 50,000 gal (2) Two resin storage tanks - 15,000 gal each (3) Three waste receiving tanks (4) Two waste sampling tants , (5) One waste neutralizing tank The radwaste demineralized is a single mixed bed unit sized for a flow of ' 200 gpm. This system also includes a pump and filter. No provision is made for regenerating the demineralized resins; spent resins will be sluiced to one of the resin storage tanks. The radwaste concentrator is sized to concentrate 5 gpm. Steam for the unit is used to evaporate the liquid waste. The concentrator vapor is con-densed and sent to one of the waste receiving tanks. The concentrated radwaste is discharged to the concentrated waste storage tank. The radwaste filter is of the recycle type, consisting of a vessel filled with diatomaceous earth or other suitable material. It is arranged for back flushing to the concentrated waste storage tank. Operation of the radwaste facility will be based on manually initiated I batch control. Instrumentation for the facility is located on a control board adjacent to and suitably shielded from the process vessels. Annunciation of "radwaste trouble" is previded in the control room.
, 4. Wastes to be Handled During normal operation of the unit only a small part of the capacity of the radwaste system will be required. Most of the capacity is provided for abnormal operation, such as maintenance requiring extensive decontamination.
- 5. Liquid Waste Discharge Limits j
Radioactive vaste discharges of process origin to the ocean will conform to limits established in 10 CFR 20 and to the waste discharge permit which will be obtained from the North Coastal Water Pollution Control Board (a state ' agency regulating waste discharges into state waters). l The radioactive concentration of waste discharges will be measured in the discharge tunnel after dilution with approximately 250,000 gpm normal con-denser cooling water flow. Monitoring of the environs at regular intervals will provide data to assure the adequacy of waste disposal controls and to indicate any need for modification of waste disposal methods. B. RADIOACTIVE SOLID WASTES l 4
- 1. W
_astes to be Handled i koutinely occurring sources are spent demineralized resins, filter material, and miscellaneous contaminated trash. Scrapped equipment and materials will
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VI-3 require storage' occasionally. } l 2. Solid Waste Storagg _ l Demineralized resins will-be sluiced tS one of the two resin storage tanks. Filter material from the radwaste filter and other sources will also be stored in these tanks. } Other solid wastes will be stored in a vault located across the road from the north side of the refueling building. This vault is provided with a carbon dioxide system for fire protection. Storage of solid wastes at the site will be temporary. Final storage. when required will be off-site. This will be done by a licensed contractor, using approved means and location of ' storage.. - C. RADI0 ACTIVE CASEOUS VASTES l
- 1. General The gaseous vaste disposal system handles air and other gas exhausts which could contain radioactive constituents. The system incorporates monitoring, appropriate holdup for radioactive decay, and provision for isolatiotr if necessary. Release to the atmosphere is through the ventilating stack.
The height of the stack, the discharge velocity, and the permissible release q rates 6111 be" determined later.' .
- 2. Sources of Caseous Wastes Process area ventilation exhausts will contain trace quantities of dusts, mists, and vapors of process origin. Similar substances will be collected at the laundry ventilation hood and the hoods in the radiochemical laboratory.
The dry well air is exposed to neutron activation around the reactor vessel, resulting in some trace cetivation' products, such as argon-41. This air is in a closed system and will be purged if n2cessary when access to the dry well is required. Non-condensible radioactive gases are removed from the main condenser by the air ejector. The turbine gland seal system also exhausts these gases, but this is a minor source compared to the air ejector. The air ejector exhaust system includes a 20-30 minute holdup (in buried piping) between the condenser and the stack. This allows for radioactive decay of short-lived isotopes to acceptable levels, and provides an ample period to detect any serious release of fission product gases from the reactor fuel. When required in such an event the holdup line can be isolated by valving. i s !
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-s The gland seni exhaust system includes a shorter holdup to reduce the activity of the short-lived radioactive gases. Activity rates of gases leaving the reactor steam outlet at 1008 Mut are given in Table VI-1.
TABLE VI-1 PRINCIPAL NORMAL OPERATION ACTIVATION OFF-GASES AT 1008 MRT Gas Half-Life Activity at Reactor Steam Outler (uc per see) N-13 10.0 min 925 N-16 7.4 sec 3.7 x 108 ' N-17 4.1 sec 4 x 103 0-19 29.0 sec 6.6 x 103
- 3. Permissible Stack Emission Rates The plant will be operated in accordance with the permissible stack emission rate to be established. This rate will be based on the allowable maximum annual dose of 0.5 rems per year at any off-site location as given in 10 CFR 20. The calculations will take into account stack height, discharge
' velocity, the topography of the site and environs, and meteorological con-ditions at the site.
The stack releases during normal operation are expected to be only a small
'S fraction of the permissible release rate. Essentially all of the permissible rate is reserved for conditions involving possible fission product release -
from defective fuel.
- 4. Off-Cas Mon:torine Continuously operated monitoring instruments are installed on continuous flow sample streams of both total stack flow and air ejector flow. These provide both a continuous record and control information on stack gas release. In addition, a collection sampler is installed in a total stack flow sample line to provide a record of any particulate matter or halogens released.
- 5. Environs Monitoring The permissible emission rate will be based on a combination of theoretical and prcctical considerations. The proof of the acceptability of these rates will be obtained from the site and environs monitoring program.
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A. SAFEGUARDS CONSIDERATIONS The various safety features of the unit have been described in preceding sections of this report. The general safeguards criteria that guide the design are in three broad categories: (1) Raiiation hazards control during normal operation, (2) Design features for preventing accidents, and (3) Design features for mitigating the effects of an accident in the unlikely event of its occurrence (Containment).
- 1. Radiation Hazards Control During Normal Operation All phases of normal operation of the plant, including disposal of radio-active waste materials, will be such that they do not result in exposure of any person on or off the plant premises to radiation in excess of per-missible limits. Radiation hazards control vill be provided by shielding, monitoring devices, and by equipment and systems for handling, storing, and disposing of radioactive materials and wastes.
- 2. Design Features for Preventing Accidents The design features that are most significant preventing a nuclear accident are: (1) the inherent safety of the reactor and (2) the systems which shutdown and cool the reactor.
The design of the reactor is such that it tends to shut itself down upon f
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a potentially dangerous increase in power. An increase in fuel temperature l or an increase in steam void volume tends to reduce reactivity and limit a i power excursion, f
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Two separate and independent neutron poison mechanisms are provided to I assure shutdown of the reactor. The control rods provide automatic fast shutdown in response to indications of potentially dangerous trends. The manually-actuated liquid poison system provides an additional source of negative reactivity for reactor shutdtwn in the event that complete shut-down cannot be obtained with the control rod system. Three separate and essentially independent reactor cooling systems are pro-vided to avoid or minimize melting of fuel. The primary heat removal system uses the main condenser by way of the turbine or the turbine bypass. The ~~ second system is the isolation cooling system, which consists of a feed-biced arrangement in combination with an emergency condenser. The third system introduces water from the suppression pool to the reactor core spray nozzles to minimize fuel cladding failure in the event the first two systems fail to keep water over the core.
- 3. Containment Should all other safety provisions fail, radioactive material which might be released in the event of an accident would be contained.
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B. SPECIFIC SAFETY FEATURES
- 1. Nuclear Excursion Safeguards l
The reactor safety system protects against unchecked nuclear excursions by anticipation of runaway conditions or by fast response to excessive power or pressure. Also, the negative reactivity effect from fuel heating (the Doppler effect) as well as the negative moderator temperature and void coefficients act to limit potential excursions. J The principal features of the design of the Bodega Bay unit which assure stable operation are: (1) high system pressure, (2) forced circulation, (3) a conservative void reactivity coefficient, (4) a long fuel heat transfer time constant, and (5) use of a pressure regulator to control steam pressure. Analytical studies of the reactor are being made as the design progresses to assure that its operation will be stabic. System analyses are also being made ro assure satisfactory dynamic response. Basic computer models used for these studies are similar to those proven by Dresden and VBUR tests. Further verification of or improvement in the analytical models will be obtained when operating data become available from several other eg boiling water reactors scheduled to begin operation before Bodega. J There has been sufficient agreement betwecn analytical predictions and tests on boiling water reactors to assure that stability and transient behavior can be predicted by the methods used for these analyses. Proof of the analytical results will come with the start-up and operational testing of the Bodega reactor. As the design develops, reactor systems are analyzed to determine the effect system operation or malfunction would have on reactivity. From the analyses of specific accidents it is concluded that no credible nuclear excursion large enough to cause physical damage external to the core could occur in a carefully designed reactor of this type. The results of some specific analyses, bases on preliminary design data, are included in the following paragraphs.
- 2. Isolation Cooling System Isolation cooling is provided by a dual sytem which includes both feed-bleed and emergency condenser functions. Following are some advantages of 3 i
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hick W M A u s M l a .:A d d 2::52;Guuczka:1hls & D O-hiliM1%hndaic M & M L a:aa N A[ z ,' F' J VII-3 / - s l< 1 i the combined system compared to either a feed-bleed arrangement or an 2($; emergency condenser, alone: I (1) Feed-bleed can readily dispose of the significant transient steam ( flow immediately following scram and loss of the main condenser. This permits a short closure time for the main steam isolation valves, independent of emergency condenser size. (2) If the emergency condenser shoald be isolated by a refueling building high pressure signal, the feed-bleed portion of the system alone can dissipate decay heat generation without operation of the main safety valves. f 1 (3) The combined system can protect the reactor during a feed failure or a total power blackout for at least two hours. (4) The combined system can function independently of the core spray . reactor cooling system. A feed-bleed system alone would involve the same systems for ultimate disposal of heat from the suppression . 1 pool as does the core spray system.
- 3. Pressure Suppression Containment Pressure suppression containment has been selected for this unit because it offers a number of advantages over conventional dry typs containment.
With regard to safety, it is believed that the following advantages are offered by pressure suppression containment: (1) In the event of a primary system rupture within the dry well and a subsequent release of fission products, radioactive material would largely remain in the dry well, which is shielded to a large extent by the reactor biological shield and surrounding ground and structures. (2) Accident energy would be absorbed by rapid and complete condancation of steam in the suppression pool. (3) The small physical size of the dry well results in an enclosure with fewer welds, lower probability of imperfections,
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O. 3 . and less. chance of escape of fission products if they should leak from the steel vessel into the concrete enveloping it.
. (4) The small physical- size of the pressure suppression oscem makes it economically feasible to locate much of it below grade. The system ',
is so. located that leakage from it would pass icto the refueling. - i building, where system leakage could be collected, cleaned up, and ,
-discharged through the ventilating stack.
(5) The water pool would retain solid and soluble fission products washed down or discharged into it and thus these would be unavailable for leakage.. It is unlikely.that fission products would accompany the 4 steam discharged to the pool. ] (- Co ANALYSES OF OFF-STANDARD CONDITIONS Off-standard conditions are those abnormal conditions which will arise in , the regular operation of the unit; such as valve testing, loss of electrical j load, and adjusting the pressure regulator. The following events are representative of conditions in this category which
.might be expected to occur. Except where noted, analyses are based on a reactor power level of 1008 Mwt and a reactor pressure of 1060 psig'as the atarting condition.
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- 1. Testing Valves in the Main Steam Line
- a. Bypass Valves - The turbine bypass valves will be tested periodically to assure that they will open in event of an electrical load rejec-tion. Each of these valves will be fully opened and closed, one at a time. Analytical studies for this test have not been completed for Bodega; however, results of studies for similar bypass arrangements, such'as at Humboldt, show that the test will not result in any significant disturbance in the reactor system. It will cause but a momentary change in the output of the turbine, and the output will return to normal after the valve is closed,
- b. Isolation Valves - The main steam isolation valves will be tested periodically by running them through full travel one at a time at rated reactor power or lower.
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w turning the centrol switch to the test position and then pushing the test pushbutton. The valve willithen close to the 25% shut position (about 10% reduction in steam flow per valve) and stop. When the . pushbutton is released, the valve will return to the open position. Analytical studies for this test for Bodega have not been completed. dowever, it is not expected that a significant flux or pressure transient will result.
- c. Turbine Stop Valvea - The main steam stop valves will be tested in the same general manner as the steam isolation valves. However, the pressure sensor of the regulator is ahead of the stop valves so the pressure drop caused by closing a stop valve will auto-matica11y be compensated for by opening the admission valves. There are limit switches on the stop valves which preven,t closing beyond the point where the admission valves could not compensate for the increased pressure drop. This test, properly conducted, will not result in a system disturbance.
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- 2. Chanking Pressure Regulater Handwheel Setting A rapid change in the pressure regulator handwheel setting could cause some overshoot of reactor pressure. A sufficiently rapid change in set point could result in a flux transient large enough to scram the reactor.
The rate of change will be limited by operating procedures to prevent the occurrence of a significant flux transient. .
,T
/
- 3. Continuous control Rod Withdrawal or Insertion .
Continuous withdrawal of control rods at power would cause flux peaking in the area from which the rods were withdrawn. The rate of power increase would be a function of the rate of rod removal, the particular rods removed, and the control rod and void configuration. The local power and over-all steam formation rate would increase smoothly. The increase in steam formation rate would tend to increase reactor pressure, which in turn would cause the initial pressure regulator to try to increase steam flow in the turbine. When the steaming rate exceeded the capacity of the turbine, pressure would rise and cause the reactor to scram on reactor high pressure. During a similar continuous control rod insertion, the pressure regulator would hold reactor pressure constant by closing the turbine admission valves, thus reducing steam flow to the turbine. The reduction in power would be smooth, and if rod run-in were to be continued indefinitely, the reactor would be shut down.
- 4. Loss of Electrical Load The turbine bypass for' this unit is designed to accommodate the full steam generation following a turbine trip-out and scram without having the I
L_______
W%tmadsh:Li.%LXSw n' "W" - 'W'5 UNA n *b &%2:SONSA?W$OU '# S=e , VII-6 safety valves operate. Analytical studies for a turbine trip-out at rated
,, load have not been completed. ~#
D. ANALYSES OF EQUIPMENT MALFUNCTION protection against the consequences of equipment malfunction is an essential objective in the design of the Bodega reactor. The main principles employed are: (1) use of equipment with the highest practical reliability against failure; (2) fail-safe design, i.e., any failure will cause actica to go in a safe direction; and (3) safety through redundant devices. The designs incorporating these principles have been summarized in the earlier Sections of this report. The present Subsection outlines the results of analyses of plant safety should any important item of equipment or system fail to function as intended.
- 1. Control Rod Drive Malfunction Safety in the event of most types of control rod drive failure is inherently provided by the number of rods available for control, and by the fact <
that in the hot operating condition many rods at random could fail without impairing the ability to shut down. Even in the cold clean condition, the reactor may be shut down if one rod of maximum worth is withheld from the core. The backup liquid poison system would permit shutdown in the event complete shutdown could not be obtained with the control rod system. The various contingencies that might arise as a result of malfunction of
~i the control rod drive system are discussed in the following paragraphs.
- a. Rod Drive Hydraulic System Failure - Loss of normal control rod drive power to a single rod would not create a serious condition and the reactor could be safely shut down. The rod drive power is obtained from one of two full capacity pumps. The loss of a pump would be annunciated and the remaining pump would be actuated. Complete loss of r.ormal drive power would be expected only with loss of station power, which would scram the reactor. Such a loss of normal drive power would not affect the power for scram insertion of the rods, which is obtained from the stored energy of the accumulators or from reactor pressure.
Accumulator failures would not interfere with the ability to scram. Only one rod would be affected by the loss of one accumulator. Scram j of this rod would be accomplished by reactor pressure if it is above about 500 psig. j e L______---_-__
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- b. Control Rod fechanical Failure - The design of the control rod system is such that a separation of a drive from its poison section should not I_\ occur. In comparison with the original Dresden design, the Bodega control rod system incorporates design improvements, including eliminat-ing the need for a drive shear pin, provisions for a new minimum clearance coupling between rod and drive, overtravel position provision for checking the integrity of the blade t'o drive coupling, and a more gradual decelerating buffer,
, Separation of the drive from the poison section would not interfere j with the insertion'of the poison section into the core. Accidents J resulting from separation of a drive from the poison section and sticking of the poison section in the core with subsequent drop out af ter the drive has been lowered are considered in Paragraph G. . If a rod were to become stuck and could not be driven into the core, the remaining rods provide adequate margin for reactor shutdown.
Protection against failure of the rod latch is provided by the multiplicity of collet fingers; all of which would have to fail for latching not to occur. Failure of the dnlocking piston spring would not prevent latching because the spring force of the fingers could return the unlocking piston to its locking position. Failure of the spring and several of the fingers could prevent rod locking after the fingers had been unlocked during a rod lowering operation. This situation would not interfere with the ' cram s function of the rod since hydraulic pressure resulting from a ' scram signal would move the rod into the core and force the unlocking piston own, permitting the ,
~}_ remaining fingers to lock. )
- 2. Recirculation Pump Failures The reactor design will be such that no fuel burnout will occur on loss of all pumps or seizure of one pump.
- 3. Main Steam Valve Closures
- a. Turbine Stop Valves - Tripping of the turbine stop valves can be caused by a manual trip or any one of several automatic trips as shown in l j
Figure III-23. The bypass system is sized to handle the transient steam flow for stop valve closure initiated at any load up to rated 1 { without safety' valve operation. ' Analytical studies of stop valve closure have not been completed but it is probable that a flux transient grear
- enough to cause the reactor to scram would result.
- b. Main Steam Isolation Valves - Accidental closure of the main steam l isolation valves would cause a pressure rise and flux transient and it is expected that a high flux or high' pressure scram would result.
Analytical studies for' isolation' valve closure have not been completed. If these studies indicate that the flux rise is great enough to endanger the fuel, contacts to initiate scram when the valves are still partially open will be added to the reactor protection system.
')-
l
lE$.~4M;$i$.U$U hoi!.hiB&Sf.il:Ah!51.$tNSn%-nd5S.chh.W12did 1.$$$.hk.&n.2M:::bi.17 I , VII-8 I After the valves close and pressure rises, high pressure indication brings the isolation cooling system into operation.
- c. Coincident Steam Shutoff With Failure to Scram - This coincidence of equipment malfunctions is assumed arbitrarily in order to establish i the design of the reactor safety valves. The improbability of such '
an occurrence can be realized by considering the number of coincident events necessary. (1) Main steam flow would have to shut off completely. This would require either that a full load turbine trip occur simultaneously with failure of the bypass valves to open, or an abnormal condition
]
cause the main steam isolation valves to shut. ; I (2) Failure to scram. The abnormal condition which caused the isolation valves to shut would also initiate a scram. In addition, both the high flux and high pressure scrams would have to fail in order
, for an automatic scram not to occur. Furthermore, the operator would be alerted by multiple alarms and loss of load indications and would initiate manual scram.
Failure of the bypass valve during loss of load coincident with failure to scram is an unlikely combination of events. Failure to scram with closure of the isolation valves coincident with failure of flux, pressure, and manual scrams is also an unlikely combination of events. ' The safety valves will be sized to provide sufficient capacity to limit , the peak pressure below that allowable by Code for these assumed com-binations of events.
- 4. Failure of a Reactor Safety Valve to Resent If one of the twelve reactor safety valves should fail to reseat following scram, pressure could fall below normal and the reactor would cool down.
The maximum rate of cool down for the case of a fully open safety valve would be approximately 10*F per minute which could be tolerated by the vessel as a very infrequent event. S. Failure of Reacter Safety System The reactor safety system is designed en fail-safe principles. Significant malfunction will cause an immediate insertion of control rods and reactor shutdown. Also, because of the number of sensors nrovided and variables monitored, failure of a sensor will not impair t a 'bility of other sensors to initiate a scram.
- 6. Fuel Cladding Failure In the event of a fuel cladding failure, protection is provided by the off-gas holdup system, its radiation monitors and isolation valve provisions.
@hnunllis.fis'2LE;3ihUiENO lbS.J&hF%h:$2EEDNdn;d5u2di-Mus ' Udu-g, l r VII-9 The system is designed to alarm at a noble gas emission rate that could s reasonably be expected to deliver the maximum permissible offsite dose if the emission were to continue at that rate for a year. Automatic valve closure will occur at a ten times higher gas emission rate.
- 7. Loss of Feedwater l
If the reactor water level should fall due to failure of the feedwater l system to supply the needed quantity of water, the reactor would be i scrammed and the isolation valves on the main steam line and cleanup system would be closed as a result of the low water level signal. The isolation cooling system would be placed in operation by the subsequent high reactor pressure signal. The isolation cooling system, which is the primary protection against the consequences of a loss of feedwater accident, could independently cool the reactor on a continuous basis.
- 8. Loss of Condenser Vacuum Loss of condenser vacuum will scram the reactor at approximately 24" Hg vacuum. At approximately 23" Hg vacuum the turbine bypass valves open.
If condenser pressure continues to rise, the bypass valves close at approx-imately atmospheric pressure to prevent rupturing the exhaust hood diaphragms. The isolation cooling system will cool the core after separation.
- 9. Loss of Auxiliary Power In the event of loss of all auxiliary power, the reactor will scram. Stored hydraulic energy would assure a scram even if the safety system lost its power supply. Reactor heat would be removed by the isolation cooling system if power to the condenser circulating water pumps were lost.
Emergency power for the operation of the reactor safety system, isolation valves, core spray pumps, refueling building controlled-release ventilation system and other critical systems would be available from the diesel generator or station battery.
- 10. Instrument Air Failure Should the instrument air compressors fail, a scram would occur in several minutes when the compressed air storage tank pressure falls. This scram could be caused by " fail-safe" failure of one of several air operated devices.
One group of devices which would cause such a scram is the control driye system scram valves.
- 11. Pressure Regulator Failure '
Failure of the initial pressure regulator could cause the turbine admission valves to clcse. The greatest flux and pressure transient 4 i
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' due to such failure would result from rapid closure of the admission
' 'l valves while the unit was carrying full load. This transient would be equivalent to the turbine stop valve closure discussed in' paragraph 3a.
-If regulator . failure were to permit pressure to drift upward significantly, a high flux or pressure scram would result ultimately. ,
- 12. Emergency Condenser Tube Failure
. Failure of a tube in one of the emergency condenser tube bundles would be detected by the radiation monitors on the emergency condenser exhaust vent.
A failed tube bundle could be isolated by manual actuation of the' motor , operated valves from the control room. The emergency condenser could continue to operate at half capacity on the remaining bundle. 'The feed-bleed system would take over the heat removal duty of the inoperative tube bundle.
- 13. Reactor System Ruptures Inside the Dry Well '
The consequences of a reactor system rupture would depend on its size. A very small rupture or system leak (one releasina slightly more energy than the heat removal capacity of the dry well cooler).would eventually. result in a high dry well pressure trip, but an earlier shutdown could be initiated manually on observation of a steady gradual rise in dry well pressure and temperature.
T A somewhat larger rupture, such as a break in an instrument line, could i
/ lead to venting to the pressure suppression chamber. Dry well high pressure would scram the reactor. Feedwater flow, isolation feed, or core spray would be available to keep the ' core covered and prevent core overheating.
Vacuum breakers on the dry well would prevent pool water from being drawn up into the bottom of the dry well after reactor pressure had been reduced and venting stopped. Major system breaks are considered under the heading of cajor accidents, Paragraph G.
- 14. Failure to Replenish Cooling Water in Emergency Condenser Gradual evaporation of the shell-side water in the emergency condenser will .
occur during its operation. Protection against failure to replenish cooling water in the emergency condenser is afforded by an initial water supply < sufficient to last for at least two hours and indefinitely with makeup water supplied automatically from the fire water storage trnk by gravity feed. D-
hiMh und$dLk&cd Min $Nih925NM3ln1&ihSd.bbh555Skkd%1. SOv ' Y i' i ,. VII-11 q In addition, a low water level alarm is provided in the control room to alert the operator. Makeup for the fire water storage tank is supplied by i automatic level control from well water or the sea water evaporator. Normal J water level in this tank is equivalent to several days supply for the emergency condenser. 1!. Primary System Ruptures Outside the Dry Well Isolation valves are used to protect against primary system ruptures outside the dry well. Low tuctor. water level initiates closure of all primary system isolation valves except those in the emergency condenser lines and those in inlet lines that are protected against reverse flow by pheck valves. Also, sensors in the refueling building and pipe tunnel are used to initiate valve closure. The area radiation monitors are available to alarm radio-activity release. , i Pressure sensors in the refueling building are used to initiate closure of j the isolation valves in (1) the supply line to the emergency condenser and (2) the supply line to the coolant clean-up equipment. The pipe tunnel has four temperature sensors located near the turbine stop valve, which are used to initiate steam.line isolation valve closure. The feedwater line in the pipe tunnel is pr6theted by the isolation check valve which could prevent any significant loss of reactor coolant through a break l in the feedwater line. I I
.) '
s
- Doors and penetrations into the pipe tunnel are sealed to prevent uncontrolled steam discharge into other building areas. The tunnel is vented directly to atmosphere alongside the condenser, and also to the turbine area by way of a removable wall section at the head end of the turbine which would be pushed out under pressure. The turbine area is open to atmosphere. The venting of i the pipe tunnel is such that rupture of the main steam line could not cause a pressure greater than the design pressure of the tunnel.
A rupture of the main steam line in the pipe tunnel would be the most severe cccident of this type because ruptures in smaller lines would release less coolant'from the reactor before isolation. Rupture of the turbine exhaust diaphragm would cause closure of the turbine stop valve and/or bypass valves from loss of condenser vacuum. A main steam line rupture is discussed in paragraph G on Major Accidents. ) E. OPERATOR ERRORS Safety features incorporated in the unit are designed to protect against accidents in the event of operator errors. The measures designed for protection in the event of transient conditions or equipment malfunctions, discussed earlier in this Section, would be available whether these contingencies arose as a result i s 1 i'
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of operator error or otheruise. The following paragraphs discuss safety in the f~) ' svent of possible operator errors not covered earlier. j
- 1. Startup Accident The control rod drive systems cnd the rod withdrawal procedures are designed to provide adequate control of the reactor during control rod withdrawal to criticality, and on to low and rated power operation. If the operator should deviate from the startup procedure and withdraw an off-standard sequence of rods leading to a rate of power rise too rapid for operator action, the '
period or neutron flux monitors would scram the reactor before power could increase significantly. A startup accident, believed to be the worst accident of this type possible - for this reactor, has been tentatively analyzed, based on preliminary design data, assuming that all of the following operator errors and equipm;nt failures occur: (1) Through error, the neutron flux level instrument range switches are set on the power range scale instead of down-scale as required by operating procedures. Thus, the reactor would not scram until the power level reached 120% of rated power and short period scram pro-tection would not be available. The range switches would normally be set down so that a scram would occur at a power level at least 1000 times below rated power. Down-scaling'these switches also' connects the intermediate range neutron instrumentation to'the scram circuits so .D1 that the reactor would also scram on short period. (2) The withdrawal interlock circuits, which' prevent-rod withdrawal unless the flux level monitors are set downscale, fail to function. - (3) The operator proceeds to withdraw the control rods as fast as they can be withdrawn and persists in_this error until the reactor is shut down ( by a high-flux scram initiated at.120% of rated power. l ( (4) It is also assumed that the operator errors include failure to follow the prescribed control rod withdrawal pattern. A criterion used in design- l l ing the reactor is that under the most adverse operating condt ions and ' rod pattern the maximum worth of a control rod is not to exceed 3.6%ak. The highest possible rod worths occur in the hot standby condition with rods withdrawn in an abnormal pattern contrary to operating procedures. Rod worths will be significantly lower than 3.6% a k for planned rod patterns or with the reactor cold. The maximum reactivity insertion rate corresponding i to 3.6% a k would be 0.4% A k per second at the 3 inches per second with-drawal rate of the locking piston drives. . Other assumptions basic to the analysis of this accident are as follows: 1 (1) The reactor level is 10-10 times rated power at the time when the delayed critical point is reached. l c ') 1 0
w
- en VII-13 (2) Fuel and moderator temperatures are initially at 550*F.
(3) The maximun available control rod reactivity addition rate of 0.47. A k per second applies continuously throughout the course of the accident. (4) The negative Doppler reactivity effect from fuel heating is the only shutdown mechanism acting to turn the resulting nuclear excursion. No credit is taken for any negative reactivity effect from moderator heating and void formation. (5) Control rod motion does not start until 0.5 seconds af ter the scram signal at 120% of rated power. The results of the analysis indicate that the power transient would be turned i by the Doppler reactivity effect before control rod insertion could ! significantly affect the power generated in the burst. The minimum reactor period would be about 15 milliseconds and reactor power would reach a peak level of 5,000 Mw. The energy generated in the uncontrolled fuel zone ! during the nuclear burst would total about 300 Mw-sec. This energy , release would increase the average fuel ten.perature in the uncontrolled I zone to 1000*F and the temperature at the hottest point to 2100*F. Thus l it is not expected that fuel melting or cladding damage would occur as a result of this accident.
- 2. Fuel Loading and Handling Accidents
'T , / Protection against criticality accidents during fuel loading operations is ,
assured by the loading procedures. These require that fuel be loaded only { when all control rods are fully inserted. Also, withdrawal and reinsertion l of a single rod before and af ter each loading increment is required to { verify that tne reactor is safely suberitical with one adjacent or l encompassed rod fully out. In addition, the refueling operator is in I communication with the control room operator who is observing the nuclear instrumentation for evidence of any approach to criticality. The safety circuits are activated to scram the control rods if an excursion should occur on rod withdrawal, and interlocks are provided to prevent lowering fuel into the reactor unless rods are inserted. ! The " Fuel Handling" and " Cask Drop" accidents which were an important con- l sideration for Humboldt, are not possible at Bodega because fuel transport ) from the reactor to the fuel storage pool is accomplished ender water. 1 The design basis refueling accident described in Section G assumes a number of coinctdent independent refueling errors, including dropping an assembly of maximum worth into a near-critical zone in the center of the core.
- 3. Cold Water Accident I
h A boiling wat.er reactor holds considerable reactivity in steam voids i
/ during boiling operation, therefore rapid reduction in inlet water tem- f perature could rapidly reduce these voids and initiate a nuclear excursion. I
-I l
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-/ l Possible causes of a cold water accident are (1) increased feedwater flow, (2) loss of feedwater heating, and (3) startup of a cold recirculation loop.
The feedwater and recirculation systems will be so designed that none of these occurrences will result in a transient severe enough to damage the fuel. F. FIRE AND OTHER CENERAL HAZARDS
- 1. Fire ,
The fire protection measures follow conventional practice as described in Section III-H.
- 2. , Earthquake The plant site is located in a region which is subject to seismic activity.
For this reason, conservative earthquake design criteria have been adopted for the unit. From the information presented in Section V, it can be con- q cluded that the structures will not fail under seismic forces and that the ' equipment can be designed to operate sa:isfactorily during an earthquake. .
- 3. Flood
<y . - There is no danger of flooding at the plant site. (See Section V)
- 4. Weather and Miscellaneous Loads All structures are designed to withstand maximum weather and other potential loadings in accordance with standard codes and normal engineering practice. ,
G. MAJOR ACCIDENTS Major accidents are those accidents which have potential for the release of a significant quantity of radioactive material from the reactor fuel. The safety < analyses presented in preceding sections of this report show that the unit design provides high assurance against accidents that could create a radioactivity hazard. Design safeguards features for Bodega would minimize the radiological consequences of a major accident. The following accidents are repr'esentative, . although it is considered highly unlikely that any of them would occur.
- 1. Control Rod Drop Accident Protection against a control rod drop accident is provided by design and operating procedures. An accident involving the drop out of a control rod .
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i 'l r Y VII-15 ! blade wh'en the reactor is critical, vill be considered in the course of l A de:ign. The analysis vill be based on the assumption that a control rod ' blade tick: in the fully inserted position, that the drive is then with- I drawn, and that the blade works loose and drops out when the reactor ,is " i at or near critical. ! The maxima::: vorth of a control rod in any possible rod pattern under any reactor operating condition vill not exceed 3 6% Ali . A vorth this high is possible only when the reactor is in the hot standby condition. Normal ' rod patterns will give significantly lower rod worths. This highest , possible rod vorth can occur only if rods are withdrawn in an abnormal , pattern causing the flux to peak around one inserted rod. Preliminary calculations have been made assuming the drop out of a 3.6% o k roc st an initial power at criticality of 10-5 times rated. Conservativecalculationsgndicatethataminimumperiodof3 milliseconds and a peak power of 2 x 10 Mw would result. Average fuel temperature in ' the uncontrolled fuel zone would reach 5500*F. The total energy release of approximately 2700 Mw-seconds in the uncontrolled fuel zone would not be great enough to endanger the reactor vessel. Fuel damage could occur in the uncontrolled zone. It is esticated that ' 33%ofthefuelrodsinthecorecouldhaveatemperatureexceedingthe esticated burnout threshold of 4000*F and that about 1% of the total fuel volume would exceed the estimated 8000 to 9000*F temperature threshold for fuel rupture from UO 2 vaP r pressure. Released fission products y essentially would be contained in the primary system and there would be little' chance of a hazardous release to the environs. f" Following are some of the design features and operation procedures which minimize the possibility of a rod drop accident. (a) The control blades are carefully designed to minimize the possi-bility of sticking in the core. The blades travel in a gap between fuel channels with approximately one quarter inch clearance and the blades are equipped with rollers which ride on the channel valls when contact is made. (b) The coupling and other control rod drive improvements significantly reduce the probability of an accidental separation of the rod from the drive. Couplings of this design have undergone extensive tests , under simulated reactor conditions and at conditions more extreme than those expected to be encountered in service in the reactor. (c) The design provision for testing coupling integrity by bringing the drive to the overtravel position provides an effective method for - verifying rod coupling prior to reactor startup, when rod following
_ _ . _ .g. ty . ; VII-16 (~S,
2' can not be checked by observing the response of nuclear ,
instrumentation when a drive is moved. (d) Operating procedures require that control rod movem~ents follow pre-planned patterns designed to minimize the reactivity worth of . individual rods. Thus, extensive fuel damage would not be expected if this accident were to occur uith normal rod patterns. .
- 2. Main Steam Line Rupturc Outside the Dry Well The potential consequences of a main steam line rupture outside the dry well will be minimiced by the use of an automatic closing isolation valve which limits discharge of primary coolant. The valvo closing time will ,
be such that there is no danger of damaging fuel due to loss of unter from i the core. It is anticipated that the fuel would be cooled to a temperature i closely approaching that of the coolant during the blowdoun period and that coolant would be restored to the core by turbine coast down inertin driving the main feed pump and by emergency feed. t A two seconds delay in closure initiation and a 30 second closure time { have been issumed for this preliminary analysis. The total coolant discharge through a t reak of one of the tuo 20 inch main steam lince has been calcu- ) isted based on a linear rate of valve closure and assuming critical flow at l the end of the pipe. On this basis approximately 116,000 pounds or 43% of I the initial water in the reactor and recirculation piping would be discharged ( through the break. The activity in the discharged coolant would be predominately N-16, which would be effectively dissipated by decay, dif fusion, and initial expansion s e a I - , . . . . . . , - , , . . -c., -
- l. . _ _ _ _ _ . _ _ - _ _ _ _ - - - _ -
r [' i VII-17 j 1 l of the steam cloud before the cloud reached the site boundary. , j If the reactor should contain fuel with cladding leaks, the reactor water i released through the break would contain.some halogen fission products. If I the reactor were operating prior to the break with a concentration of halogens , in the reactor water of 50 ue/gm (which corresponds to approximately a 5 curie - i per second noble gas stack release rate, based on the relative fuel to water release rates for noble gases and halogens observed at Dresden), the coolant discharged to the pipe tunnel could contain approximately 2600 curies of halogens. Normally the halogen concentration in the reactor water will be far less than this. Approximately 607. of the discharged coolant, carrying at least a proportional amount of the released halogens, would be expected to remain as water and rain out in the pipe tunnel or on the plant site. The other 407. could be discharged
- to the atmosphere as steam. This analysis assumes a proportional amount (407.)
of the halogens remains with the steam. On contact with the atmosphere the outer edges of the cloud would mix with the air, but the inner portion would not readily mix. The inner portion would be buoyant, with a density roughly equivalent to air heated to 600*F. This portion , of the steam cloud with any contained halogens would rise to a high elevation -'I and diffuse. Ground concentrations of halogens from this inner portion of the cloud would be small. s Highest ground concentrations would result from halogens contained in the outer i edges of the steam cloud which mix well with air near the ground. Following " mixing, the resulting equilibrium saturated air would be only slightly buoyant ' (equivalent to air at 0-7'F above ambient, depending on air conditions at the time). Under very unstable meteorological conditions, it might be possible for . a large fraction of the cloud to be well mixed near the ground. The volume of an equilibrium saturated air cloud containing all the released steam would be q approximately 1000 cu ft/lb of steam or about 50 million cubic foot total. , Neglecting buoyancy and halogen fall out and assuming a hemispherical cloud and unstable atmospheric dispersion, it is estimated that inhalation of this cloud ; at the nearest site boundary during its period of passing (10 mph wind) could i cause a thyroid dose of about 2 rems, well below levels generally considered acceptable in unlikely accident situations such as this.
- 3. Reactor System Rupture in the Dry Well If a medium size primary system rupture (up to about one foot diameter) wern to occur in the dry well at operating conditions, high dry well pressure or low reactor water level would shut down the reactor and close the containment j system isolation valves. The steam or water loss plus continued feedvater l flow or isolation feed flow would rapidly reduce reactor pressure to 150 psig I and permit the core spray system to operate.
l i 5 l 1
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k i d e b M _ % l E m x.r2 -i.f d -3 h !,4 W R W h 0l.,iL:$ m th E W m & h .i . d h niu h c. n l : m .. a i ;- y - VII-18 m Consequences of a large system rupture are described below as the
% ximum Credible 0perating Accident".
- 4. Maximum credible Operating Accident (MCOA)
The maximum credible operating accident for the unit is defined as the worst coolant loss accident which could result from the near instantaneous rupture of any pipe connected to the reactor vessel. The pressure suppression containment system is designed to withstand this accident. A major rupture of one of the reactor water recirculation lines would result in the highest rate of coolant loss. The postulated MCOA is based on the following assumptions: (1) There is a near instantaneous rupture of a reactor water recirculation line. All reactor water is discharged through a flow area equal to the sum of the cross-sectional areas of an inlet line plus an outlet line (6.4 sq f t). (2) At the une of the accident the reactor is at an overpressure , condition (1250 psig), with normal water level. (3) The reactor has been operating at 1008 Mwt. (4) The accident defined by assumptions (1) through (3) should not have significant radiological consequences because the core spray system would minimize fuel melting and limit fission ' product release to a small fraction of the total fission ' T product inventory in the core. However, in order to illustrate , the effectivences of the pressure suppression system in contain- ' ing fission products, the MCOA analysis is based on the further assumption that half of the core melts from decay heat. ' There is no intention to represent the MCOA selected as being particularly
" credible". Rather, the conditions chosen for analysis are judged to be at least severe enough to constitute an upper limit for the severity of credible operating accidents, and accordingly are suitable for demonstration of the effectiveness of containment.
The following events are not considered credible for the unit: (1) A reactor vessel rupture. However, the design is such that a rupture equivalent in area to a 28" pipe can be accommodated at any location 1 l I i l
. (
_ - _ l
Imh& ens ndddregninw.idGM5:eieskinMMn "G kE5SJs4 d* 'dIWSbbb y - p VII-19 on_the reactor below the refueling seal ring, and a break equivalent in area to a 10 inch pipe can be accommodated above the seal ring. (See Sections VII-I and -J.) (2) A nuclear excursion coincider.t with the MCOA or large enough to initiate the MCOA, (See Section VII-B.) . (3) An explosivecchemical reaction during the MCOA. (See Section VII-I.)
- a. Pressure and Dynamic Effects - Certain dynamic effects could occur at the time of a pipe rupture. The dry well structure will
' be designed to withstand jet impact and shock waves associated with the MCOA. The dry well structure will be evaluated for its resistance to assumed missiles, such as broken pipe accelerated by jet forces.
Also, the reactorcsupports will be designed to withstand jet reaction forces. . The pressure suppression tests described in Appendix I have demon-strated the rapid and' complete condensation of steam under conditions representing the maximum credible operating accident. The test
- results show that the dry well pressure would not reach its design value of 62 psig.
After a few secondstthe dry well pressure would fall off as the available steam and water was expelled from the reactor.
- b. Post Accident Pressures and 1.eakage Rates - Following the pressure
, ). suppression transient, flow of feedwater, isolation feed, or spray water would rapidly quench the hot (275'F) steam in the dry well. As dry well pressure fell below chamber presstre, vacuum breakers would operate to equalize pressures in the dry wall and chamber. - Calculated post-accident containment pressures are shown in , Figure VII-1. This Figure is based on the assumptions that the ] suppression chamber vent line is closed at the start of the accident, j that the post-accident coolingt system (core spray pumps and suppres-sion pool coolers) begins operating in one minute, and that at fifteen minutes a changeover is made to sea water for the cooling water side of the coolers. The upper curves in this Figure show the corresponding leakage rates based on an initial leakge rate . of 0.57 per day for the dry well at its design pressure. l 3
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VII-20
,3 c. Release of Fission Products - The assumptions and basis of i; / the calculated fission product release follow:
(1) A fully effective core spray would preclude significant core
' damage and result in little or no off-aite radiological. effect.
However, this analysis assumes a highiy improbable independent failure, partial ineffectiveness of the core spray, which would permit half the core to overheat, from decay heat.
~
'(2) The core heat-up curves are shown in Figure VII-2. These curves were estimated assuming that coolant loss occurs in about 8 sec, as shown by test results, and that no decay heat '
would be lost from the uncooled fuel. The pressure decay curve (Figure VII-1) is on the same basis except that the heat stored in the uncooled fuel is assumed to become available to cause pressure when fuel melts. The fuel melting curves are conservative because thermal radiation and other heat loss possibilities which have been neglected would be expected to limit temperatures and the extent of meltdown in the overheated fuel. (3) Fission products are assumed to be released from overheated , fuel at a rate proportional to the fraction of the core which is at cladding perforation or fuel melting temperature. The s following fractions of the fission products in the overheated ,j portion of the fuel are assumed to be released: Fuel at Cladding Fuel at Fuel , Perforation Temperature Melting Temperature (1600'F) (5000'F) Noble gases (Kr, Ke) 20% 100% Halogers (Br, I) 20% 100% Volatile solids (Te,Se.Cs,Ru) 0 50% Other solids 0 1% O a -
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- d. Fission Product Inventories - Based on the above assumptions, cal-culated fission product inventories as a function of time after the maximum credible operating accident are shown in Figure VII-3. This Figure is based on an estimation of an effective fall out and plate out half life in the refueling building of 12 hours for all fission ,
i products except noble gases. Also, credit is taken for removal by the ventilation clean up equipment of 957. of halogens and solids , discharged from the refueling building. The Figure does not show directly the fission product inventory in l- the post incident cooling water. This water could contain nearly , all the fission products represented by the difference between those released from the reactor and those in the dry well air space. l The total inventory in the ventilation system clean up equipment would be approximately nineteen times the values shown on the curves of fission products discharged to atmosphere, excluding the noble gases. Total refueling building inventory would be the sum of the quantities in the air space, on the surfaces, and concentrated in the clean up equipment. Release rates and radiological effects of this accident are described l in Section VII-H. l
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- 5. Design Basis Refueling Accident h Protection against' fuel loading accidents is described in Section VII-E-2.
The design basis refueling accident is a fuel loading accident compounded by a number of independent refueling errors. It has been used to provide a design basis for the refueling building controlled-release containment system.
- The following events would have to occur in order to produce this accident: j (1) Control rods would have to be withdrawn in a unique pattern around the vacant fuel position to be loaded to bring the array to near-critical. -
(Loading procedures and interlocks both prevent fuel loading if all control rods are not inserted. Also, the loading procedure requires , verification that the reactor is safely suberitical by withdrawing and reinserting a control rod before and after each loading step.) (2) The control room operator would have to fail to notice the indications from his instruments that the control rods were out and that the reactor was near critical, prior to loading,, (Procedures require him to observe this instrumentation end to be in communication with the refueling
- operators during all fuel loading operations.)
(3) The fuel element handle, the fuel grapple, or the grapple cable would have to fail. (4) The fuel bundle would have to be dropped from a position that would allow it to fall into the vacant position and meet no resistance to its fall. (5) The fuel would have to have a relatively low irrad.iation exposure. (The , j poteni;1al reactivity worth of a fuel bundle loaded into a near-critical ; lattice of fuel after burnup has progressed far enough to require refuel-ing is too low to result in an excursion that could cause fuel burnout I or melting even if the fuel element were dropped.) l The maximum potential reactivity addition occurs with fuel irradiated to about i 2000 Hwd/ ton. At this level, withdrawal of four adjacent rods will bring a zone { containing 15 fuel assemblies and a central unoccupied fuel position to near { critical and the reactivity worth of a fuel assembly added to the unoccupied ! I fuel position will approach 3% a k. This analysis assumes that the assembly is worth 3% o k and that it is dropped from a position two feet above the core. This gives a maximum reactivity addition rate of 15% Ak per second. The effects of the resulting power transient have been calculated assuming an initial fuel and moderator temperature of'68'F and an initial power level , of 10-10 times rated. The calculations indicate that the transient would have a minimum period of 2.3 milliseconds and that the energy generation in the uncontrolled zone would total 800 MW-sec. These calculations assumed that the reactor is not shut down by scram or fuel destruction until af ter the power burst. s The average temperature in the uncontrolled fuel zone would rise 3100*F as a i result of this burst. Calculations which neglect the heat of fusion and heat of vaporization of the UO2 indicate that the peak fuel temperature would reach j 9000*F. About five pounds of UO2 would. reach temperatures in excess of the estimated 8000*F temperature threshold for the rupture of l 4 e *
*iLS.lin w w & L a .%,& s w i n W G i a;utE &lr B WA M h % k 63 M L M . % hn ? .'
VII-23 the fuel rods from UO2 vapor pressure. About 400 pounds of UO 0"Id reach temperatures in excess of its 5000'F melting temperature.2 "About 700 ' O of the 784 fuel rods in the uncontrolled zone would contain fuel at tem-peratures in excess of the estimated 4000*F threshold for fuel rod burnout. It is estimated that the fission product release fractions for the maximum I credible operating accident, given on page VII-20, at 1600*F and 5000'F also { apply to the design basis refueling accident at" fuel temperatures of 4000 and 5000*F, respectively. l The fission product release has been calculated based on an estimated de-contamination factor of 105 in the reactor water and an effective fallout and plate out half life in the refueling building of 12 hours for all l fission products except noble gases. These removal mechanisms are assumed to have no effect on noble gases. Also, credit is taken for the efficiency of the ventilation cican up equipment, which will be designed to remove at I 1 cast 95% of the halogens and solids discharged from the refueling building. Based on the above, calculated fission product inventories are shown as a 1 function of time after the design basis refueling accident in Figure VII-
- 4. The total fission product inventory includes the fission products generated in the burst. The inventory of fission products from an 800 MW-see burst has been calculated from data in USNRDL-456, " Calculated Activities and Abundances of U235 Fission Products", R. C. Bolles and N.
E. Ballove, August 30, 1956. Approximately 25% of these fission products would be generated in fuel that reached 5000*F and above and about 95% would ( be in fuel rods that reached 4000*F at some point. Release rates and radio- t logical effects of this accident are described in Section VII-H, following. H. RADIOLOGICAL EFFECTS OF MAJOR ACCIDENTS , 6 Fc,11owing is a summary of the radiological effects of the maximum credible operating accident and the design basis refueling accident, based on assumptions ' given in Section VII-G. Credit is taken for the efficiency of the ventilation cleanup equipment which is brought into operation in the event of a major accident. The ventilation cleanup equipment is specified to retain not less than 95% of the halogen and solid fission products. Figure VII-5 shows the calculated rates of release of fission products from the stack as a function of time after accident. j The highest discharge rates would produce measurable doses in the environs of the unit as indicated in the following:
- 1. In an unstable 10 mile per hour wind, the maximum exposure rates on the ground downwind of the site would be expected to occur at a distance of approximately 0.6 miles, assuming a 300 ft plume center height over the affected terrain. For this condition, the maximum exposure rates and integrated exposures ere calculated to be as follows:
- a. A noble gas stack discharge rate of 5 curies per second would result in
- a maximum exposure rate of approximately 0.002 rem per hour.
- b. A halogen stack discharge rate of 0.002 curies per second would result
; in a maximum exposure rate to the thycold 61 0.0011 M per hour of
'~ exposure.
- c. The discharge shown in Figure VII-5 for the design basis refueling accident would result in a maximum integrated exposure of approximately I
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Nibh3OdfMA$h13OI /N5NMdN4 EMN O$e'+b^.E M 'Ss1 18 3d d " M k hb ['r . VII-24 () .. 0.034 rem to the whole body and approximately 0.00004 rem to the thyroid, taking no credit for changes in wind direction.
- d. The discharge shown in Figure VII-5 for the MCOA would result in a '
maximum integrated exposure of approximately 0.024 rem to the whole body ,- and approximately 0.019 rem to the thyroid.
- 2. Calculations for the stable case require that an assumption be made relative to the behavior of the stack plume in the vicinity of elevated ground. This behavior will be investigated by tests soon to be conducted at Humboldt Bay.
For the present it will be assumed that the plume would pass over elevated ground, and an arbitrary height of 200 ft. above ground has been used in calculating the following exposure rates and integrated exposures; which, for a moderately stable 5 mile per hour wind, would occur at a distance of approximately 3 miles downwind of the site:
- a. A noble gas discharge rate of 5 curies per second would result in a maximum exposure rate of approximately 0.01 rem per hour,
- b. A, halogen stack discharge rete of 0.002 curies per second would result in a maximum exposure rate to the thyroid of 0.005 rem per hour of exposure.
'] c. The discharge shown in Figure VII-5 for the design basis refueling accident would result in a maximum integrated c::posure of approximately 0.16 rem to the whole body and approximately 0.0002 rem to the thyroid, taking no credit for changes.in wind direction. .
- d. The discharge shown in Figure VII-5 for the MCOA would result in a maximum integrated exposure of approximately 0.11 rem to the whole body and approximately 0.09 rem to the thyroid.
In the unlikely event of one of these accidents occurring coincident with a high wind velocity, the ventilation cleanup equipment and the stack could be bypassed by exfiltration from the refueling building at areas of low s external pressure. Assuming 100% volume per day exfiltration at ground i level during a 40 mph unstable wind and assuming a 50% plate out factor in tha leakage path for fission products other than noble gases, maximum halogen release rate to the atmosphere for the MCOA would be approximately 0.01 curies per second which could cause a maximum dose rate of approximately 0.01 rem to the thyroid per hour of exposure at one-half mile, the nearest point off-site i which might be continuously occupied. ! The direct radiation from the refueling building has been estimated based on f preliminary design. For a refueling building inventory of 400,000 curies the j estimated dose rate is 0.0004 rem /hr at 1/2 mile. The maximum continuous occupancy dose due to direct radiation would be approximately 0.01 rem at l 1/2 mile. ' s
,) The halogens and solids discharge would be such that the resultant maximum fall-out levels at any point and for any atmospheric stability would have i no significant effect on the use of land and agricultural products.
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<s 1. Incredibility of Large Reactor Vessel Breaks j
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- a. Power Boiler Statistics - Statistics on reactor vessels are not available l because of the relative newness of this application. However, good statistics are available on power boilers, and these are generally l 0
applicable to reactor vessels. It is estimated that there are 400-500 - boilers in the United States designed to operate at a pressure over 600 li psia, and that they represent not less than 4000 boiler-years of operating experience. The first such boiler was designed over 30 years ago. An j examination of the Hartford Steam Boiler Inspection and Insurance Company power boiler statistics shows no failure of steam drums in power boilers designed to operate over 600 psia. The materials used in the fabrication of reactor vessels are equal or superior to those used in these drums. Statistics on all types of boilers and unfired pressure vessels show the following relevant points- - l (1) A number of boiler explosions have occurred in small . low pressure l heating boilers. The absence of such failures in power boilers operating over 600 psia is attributed to additional care in fabrication and operation of these units. 1 4 (2) Those failures that did occur were of ten attributable to inadequate j operator training and supervision. ,3 (3) The boiler explosion statistics indicate a large reduction in the incidence of failure over the years.
- b. Brittle Failure - The following considerations with respect to the Bodega reactor vessel assure that it will not fail in a brittle manner: -
1 (1) The vessel steel in the high flux region opposite the core is j specified to have a nil ductility transition (NDT) temperature no {' higher than 10*F. This low NDT temperature provides considerable margin for radiation effects. Over the lifetime of the unit, parts ; of the vessel may receive integrated neutron exposures of up to 3 x l 1018 neutrons per sq cm, which at room temperature would not be l expected to raise the NDT over 100*F.* (2) The vessel steel is specified to be low in notch sensitivity. j
- Vessel notches are kept to a minimum by avoiding notch-like surface !
configurations and by requiring fabrication methods which avoid j sharp notch flaws. ; (3) The vessel will be initially tested at 1875 psig, which is considerably higher than operating and possible transient pressures. l i (4) The vessel will be built in accordance with the ASME Boiler and
- Pressure Vessel Code as modified for nuclear reactor vessels. !
- c. Ductile Failure - The allowable design stress for this ASME Code vessel )
) is less than one fourth the ultimate strength of the material. A ductile *
*J.
. V. Alger and L. H. Skupien, " Neutron Embrittlement at 500 and 600 Degress F of Reactor Pressure Vessel Steels," ASIM Special Technical Publication 276, pages 1164 136, American Society for Testing Materials, 1960. .
l
- - _ - - _ _ - - -_ __.___________Q
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' shear type failure resulting from loadings which exceed the ultimate g~, strength of the material over a significant section of the reactor
'i j vessel could occur only from extreme overpressure or shock waves from a ,
rapid nuclear excursion or chemical reaction. ' The safety'va'1ves are sized to prevent reactor pressures in excess of ' the ASME code allowed overpressure. The nuclear safety system and reactor - characteristics preclude any large nuclear excursion or chemical reaction. , Therefore, a ductile failure is not considered credible.
- 2. Nuclear Excursion This is considered in Section VII-B. '
L
- 3. . Chemical Reaction !
( i (It is possible that zirconium cladding may be used on future cores. A chemical reaction is much less likely when stainless steel is used. The , possibility of a zirconium-water reaction is considered here only to point up the fact that a chemical reaction involving the core is extremely unlikely.. ] Ihe maximum credible operating accident is not assumed to involve a significant coincident energy contribution from a zirconium-water reaction for the follow-ing reasons: '
- a. It has been demonstrated conclusively that a violent reaction of zirconium 4 with water cannot occur when the metal temperature is below its melting
' t, point. Since the first fuel cladding would not reach melting temperature for several minutes after system blowdown, it is improbable that a reaction would occur during the blowdown period.
- b. Experimental studies further indicate that a chemical reaction could not proceed rapidly unless the molten metal were dispersed into particles with a diameter of one millimeter or smaller. If cladding were melted by decay heat as a result of the postulated rupture accident, the process would be relatively slow, and it is not expected that an appreciable amount of the zirconium would be dispersed as particles this small.
Thus,it is believed improbable that a significant fraction of zirconium would react in the course of the accident. Even if a reaction occurred, the energy generated would be small in comparison to the total MCOA , energy release and would be absorbed in the pressure suppression system
~
withoutssignificantly increasing pool temperature and system pressure'. l l J. PROTECTION AFFORDED AGAINST REACTOR VESSEL FAILURES , The pressure suppression system will withstand a break in the reactor vessel i itself equal in area to a 28" pipe at cnylocation below the refueling seal ring, and a break, equal in area to a 10 inch pipe at any location above the seal ring. This is not to suggest that a reactor vessel break of such an area, or any other specific area, is credible. But this design basis provides am'ple margin over any possible limited damage in the form of cracks at seals, nozr.les, or elsewhere.
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! 1 VIII-1 OPERATION y !
A. ' OPERATING PRINCIPLES The basic operating principles which underlie the design of the plant are the following: ,
- 1. Operation and control of the reactor and most of the process equipment is' '
centralized in the control room, which is located adjacent to the refuel-ing building. The control room contains the control console and control boards for the reactor, reactor auxiliary systems, nuclear steam supply system, turbine, generator, and the electrical systems.
- 2. The control room is shielded so that it is tenable with regard to direct radiation in event of any credible operating or refueling accident.
3 control of the reactor win be manual, with the reactor operator seleet- ! ing ,and positioning control rods to achieve desired power level and flux distribution.
- 4. Operators may perform certain operatin6 functions at control panels and .
valve racks outside the control room but only at the direction of or - with the prior knowledge of the control room operators. 5 Maintenance of most facilities outside the dry well may be undertaken by contact methods and without over-all plant shutdown when such work can 1,. be accomplished without exceeding permissible radiation exposure limits. I
- 6. Minimum equipment requirements for operation and operating limits will be
- established for each system and these requirements vill be stated in the .
Technical Specifications, to be submitted latter. 7 . Except for some areas, the refueling building is habitable durin6 normal operation. Radiation monitoring by fixed or portable instruments win be provided for entry into all radiation zones.
- 8. Irradiated fuel and other core components win be handled under water by semi-remote mothods using lon8 Brappling poles and motor operated hoists.
Fuel vill be transferred under water from the reactor to the spent fuel storage pool. 9 Dry ven and suppression chamber containment provisions vill be in effect durin6 all periods of reactor operation when the reactor head is in place. This provision vill apply during startup and shutdown operations. i
- 10. Refueling building containment provisions win be in effect during all periods of refueling operation and during all other irradiated fuel
- handling operations, includin6 those involved in movin6 the fuel shipping cask out of the refuelin~g building.
s I
YA454%$MM:&*:ZMla*LWUch.:W O ? U:n.G% 5 sis.M RdELL C&Qi k u:O s.x n ~ v 3 VIII-2 . 3 10. Operation of the radioactive waste handling system will be such as to give reasonable assurance that the disposal ~6f radioactive materials vill not , result in the exposure of any person on or off the plant site to radiation in excess of permissibl'e limits.
/
- 11. The plant is so protected by automatic safety devices that no single -
operator error or reasonably conceivable combination of operator errors could cause a severe accident. , B. ORGANIZATION The Plant Superintendent and key supervisory personnel of the plant organization will be responsible for on-site activities in connection with operation of the plant. The plant organization will be under the direction of, and will receive technical support from, the Company's General Office in San Francisco. In addition, consulting services from General Electric Company or others will be used as required. l The plant organization chart for startup and initial operation, and the chart of the anticipated organization for routine operation will be submitted later. C. PROCEDURAL SAFEGUARDS Procedural safeguards will be established to assure the safe operation of the ' l plant. These procedural safeguards will include the following: 1 1. Detailed operating and emergency procedures
- 2. Radiation control standards
- 3. Equipment clearance procedure
- 4. Measures to prevent operating error
- 5. Measures to be taken following unusual or unexpected incidents
- 6. Anti-sabotage measures D. PREOPERATIONAL TESTING .
A program of preoperational testing of equipment and systems will be carried j out prior to initial operation of the plant. The purpose of this program is I to demonstrate that the plant has been built according to specifications and l that it is ready for fuel loading and startup. I i The test program will include those checks, e.djustments, calibrations, and l operations required to assure that fuel loading and plant operation can safely be undertaken. Each system will be functionally tested, with special emphasis l placed on those systems which are important to the safety and operability of ' the plant. The preoperational testing program will be sub=1tted later. E. INITIAL FUEL LOADING AND STARTUP A . pro 6 ram for activating the plant vill be carried out. . This program will cover the period of initial fuel loading and progressive
)
fElMN$M?MMOkit25$MdMENNSC$*& ~ CLQ '4hLE * "C 20 Y.fC.L N i k., ,. . V \ 5 VIII-3 i l steps leading to full power operation. Emphasis vill be given to matters which potentially could have an effect on nuclear safety. Written procedures vill be prepared for carrying out this startup program. ' F. NORMAL OPERATION 1 Detailed written operating procedures for all modes of plant operation vill i be prepared prior to startup. Results of tests conducted during startup , l Vill be used to modif y these procedures where necessary. The following is an outline of the principal normal operating procedures having a potential - effect on the safe operation of the plant. These procedures vill be outlined , in more detail in the Final Hazards Summary Report and Technical 3 j Specifications.
- 1. Cold Startup
\
A cold startup vill occur each time the reactor is returned to service followingrefuelingand/oramaintenanceshutdown. This vill be done by following a startup check list to demonstrate that all applicable equip-ment and systems are functioning properly.
- 2. Hot Startup '
When the reactor has been shut down for d short period of time and the reactor vessel and auxiliaries have remained at or near operating te=perature, a hot startup procedure may be followed to return the plant ,
,_N to service. This procedure vill be essentially independent of the cause
'of shutdown provided the cause is known and any non-standard conditions have been corrected.
3 Normal Power Operation During normal power operation the turbine initial pressure regulator vill maintain the reactor pressure at its proper value by adjusting the turbine control valves. The plant load vill be established by the control rod positions. Changes in load vill be made by adjusting the control rod pattern. G. WASTE DISPOSAL SYSTEMS
- 1. Liquid Wastes The plant is designed so that contaminated or potentially contaminated liquid vastes are collected in vaste collection tanks. The contents of , j thece tanks are then pumped to the radvaste facility for sampling and analycis, treating, and disposal.
1
.s i
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e VIII-4 Each batch of vastes vill be sampled and analyzed before release to
- n the discharge tunnel. Records vill be maintained of the volume and ;
concentration of all vastes released. In addition to this control, < a liquid process monitor is provided to monitor liquid vaste, effluent, and a continuously operated sampler is installed in the discharge tunnel to provide a composite sample of plant effluent for analysis
)
as a final audit of the amount of radioactivity discharged to the < Pacific Ocean. +
- 2. Solid Wastes Spent resins vill be sluiced to the resin storage tank for long' term storage and ultimate off-site disposal. ,Other solid vastes will be
.monitcred and, appropriate contamination control measures followed while storing such vastes in underground vaults. Such vastes vill '
ultimately be disposed of off-site by an AEC-licensed contractor. , ) { 3 Gaseous Wastes The gaseous radioactive vastes are' discharged to atmosphere through , the ventilating stack. A stack gas monitoring system is provided , which will be set to alarm at the permissible average release rate, -
- to be established. In the event the release rate should exceed '
this level, appropriate action vill be taken to control releases l vithin license limits.
' i.y
- 4. Environs Monitoring '
A preoperational monitoring program vill be initiated at least two years prior to initial operation. Portions of this program vill be continued after operation begins so that the adequacy of control procedures and release rates can be demonstrated. This program vill be submitted later. H. IEFUELING 1 Prior to refueling, the reactor must be shut down and the dry well and reactor vessel heads removed. Refueling bui1Mng containment must be in effect. Spent fuel and irradiated core cogonents vill be removed from the core by means of long handled grapples and the refueling vinch. Personnel shielding vill be provided by a minimum of 12 feet of water. . I. M_ ADTfENANCE l Ma,jor everhauls requiring a plant shutdown vill be scheduled consistent with ' system power requirements and plant refueling schedules. All maint'enance work 1 l lJ l l
. .-,m . . - ,
d-8&.=.L~15dRzudnMisdA?%::D$&&L dd.E&li&$.AbihElit%;4$Lli2:5k* ' F VIII-5 i 3_ will be performed in accordance with the limits and procedures established in the Radiation Control Standards. The basic principles of maintenance are:
- 1. Maintenance will be done in accordance with schedules and following procedures established to minimize the possibility of error or damage to equipment.
- 2. Records will be kept of equipment and system maintenance to facilitate !
scheduling. I
- 3. A reactor shutdown will be required before permitting maintenance on the reactor or its auxiliaries located inside the dry well vessel. 1
\
- 4. The maintenance of most auxiliary equipment outside the dry well vessel will be undertaken by normal contact methods where radiation levels permit.
i
- 5. Radiation monitoring will be provided as required during the disassembly and maintenance of any equipment subject to contamination or activation.
- 6. Protective clothing will be worn and other radiation protection procedures will be observed as required to avoid the spread of radioactive contamina-tion.
- 7. Equipment will be decontaminated as necessary prior to the performance of I maintenance.
- 8. Work on contaminated equipment will generally be performed in the hot machine shop.
J. STORAGE ,
- 1. New Fuel New fuel will be stored vertically in the fuel storage vault. The fuel will arrive at the plant in shipping containers designed to preclude criticality. Similarly, the racks for storage of fuel in the vault are spaced to be critically safe even if the vault should become flooded. q
- 2. Irradiated Fuel "
1 Irradiated fuel removed from the reactor will be placed in the fuel storage pool. Storage racks in this pool are spaced to be critically ) safe. After a suitable decay period, probably 90 days or more, the fuel will be placed in the shipping cask and removed from the plant.
- 3. Contaminated Material l \
Material which becomes radioactively contaminated will be surveyed as ! near as possible to the point of origin of the contamination. If the
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7 t 'i e VIII-6 4 material in question' carries significant smearable contamination, it will
; ,be bagged and sealed. In some circumstances, temporary storage may be l l
arranged in the refueling' building. Most such items will normally be
. packaged and stored in the solid waste vault.
- 4. Explosive Cases Hydrogen, acetylene, and propane will be stored in cylinders or tanks on l the site. Company safety standards are followed in the storage and use of.these gases.
- 5. Acids and Caustics Company safety standards will be observed in the storage and handling of these corrosive materials. Pumps, tanks, valves, and piping are selected for the safe handling of acids and caustics.
- 6. . Flammable Material All flammable material will be stored in accordance with Company safety ,
standards. Lube oil' systems and the waste storage vault are protected by , CO, systems. All buildings on the site are covered by.the plant fire water l lo8p. - l t - ) l i i
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APPENDIX l , , PRESSURE SUPPRESSION TEST PROGRAM - i i i' 4 s 4 e o I e s
. 1 Appendix I )
w - j
- 7 8 TABLE OF CONTENTS i
e. A. . INTRODUCTION AND
SUMMARY
1 l B.' PURPOSE OF TEST PROGRAM 1 l C. DESCRIPTION OF TEST FACILITY AND RELATIONSHIP TO l BODEGA BAY DESIGN 2
]
i D.- METHOD OF PERFORMING TESTS 5 { E. TEST PROGRAM 6 F. TEST RESULTS 8 G. CONCLUSIONS 11
- .. LIST OF FIGURES Figure No. Follows Page No, i
.I
- 1. PRESSURE SUPPRESSION TEST FACILITY 2 i
- 2. ARRANGEMENT OF TEST FACILITY 2
- 3. . SUPPRESSION CHAMBER ARRANGEMENT 3
'4. RUPTURE DISK ASSEMBLY INSTALLED 4
- 5. R'UPTURE DISKS AND ORIFICE PLATE 4
- 6. , PIPING AND INSTRUMENT DIAGRAM S
- 7. INSTRUMENTATION IN CONTROL SHACK 5
- 8. TYPICAL PRESSURE TRACES 8 -
- 9. TYPICAL PRESSURE TRACES 8 l 10. . TYPICAL PRESSURE TRACES 8 l l .
11'. PRESSURE TRACES WITH DRY WELL PRE-PURGE 8 s.
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(75; PRESSURE SUPPRESSION TEST PROGRAM '
^;s A. Introduction and Summary
- A' pressure suppression test facility was constructed and operated at the Company's Moss Landing Power Plant in the summer of 1962 to proof test' the Bodega Bay containment' design. In general, the facility was similar'to that
- for Humboldt Bay and used some of the same hardware. The facility consisted . ,
- of a- full scale 1/112th segment of the Bodega suppression chamber with one .
full scale vent pipe and one.model reactor and dry well vessels with about 1/112th the volumes of the Bodega design. The results of the test program confirmed the adequacy of the Bodega pressure suppression containment design. Principal results were: ,
- 1. Condensation of steam in the suppression pool was rapid and complete under conditions far more severe than those associated with the maximum credi-ble operating accident (MCOA); the suppression chamber pressure did not exceed 30 psig on any test.
- 2. The highest dry well pressure obtained on any test simulating the design basis accident for the dry well was 52.psig. (The design basis accident is described in Section C of Appendix II.) ,
3.- Variations in suppression pool water level, pool temperature, dry l
- g. well temperature; subcooling of the reactor vessel water; and use of nozzles ' j
~
) with rounded entrances as well as shar'p-edged orifices'had only minor effects !
on the operation of pressure suppression. ,
- 3. Purpose of Test Program i
Although basically the design of the Bodega pressure suppression containment is quite similar to that for Humboldt Bay, there are some significant differences.
~
- 1. The vent pipe size is 24" diameter rather than 14" diameter. .
- 2. There are some appreciable differences in the geometry of the two systems,
- 3. In Bodega the suppression pool is being " work'ed harder", that is, i I
the energy input rate to the pool per unit water volume is higher by a sub-stantial factor than at Humboldt. Test results showed this fac;or to be about , 2.4 under MCOA condition. ^ 1 l I l
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i Because of these differences,_ testing of the Bodega containment design to
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demonstrate satisfactory performance was considered to be' desirable. ' C. Description of Test Facility and Relationship to Bodega Bay Design
- 1. General Arrangement Figures.1 and 2 show the general arrangement of the tese facility.
The Bodega reactor vessel was simulated by a model reactor vessel
.which was mounted on the side of the partially buried tank containing -
the suppression chamber. The other large tank shown simulates the .! Bodega dry well vessel and is connected to the suppression chamber by a vent line iimulating one of the Bodega vent lines. . A rupture accident is simulated by breaking rupture disks mounted
- on the flange:at the bottom of the reactor vessel and' letting the -
water and steam in the reactor vessel discharge into the dry well , through an orifice'or nozzle. . ! l
- 2. Reactor Vessel ,
94 The. test reactor vessel was 27" I.D., 21 ft. long over heads and
- was designed for 1,250.psig internal pressure in accordance with '
the'ASME, code. 'It was made from carbon steel. It was equipped with -
'I a 10" discharge nozzle and other nozzles for instrumentation,' venting, )
draining, and admission of heating steam. Its contained volume was 80 ft.3 and most tests were run with about 54 ft.3 of water in it. This'latter figure is about 1/112th of the water.in the Bodega primary system.
- 3. Dry Well The test dry well was 85" I.D., and 29 ft. long over heads. It was l l
designed for'an internal pressure of 150 psig in accordance with the _l ASME code and was made from carbon steel. It has a 20" inlet nozzle l for the steam-water mixture from the reactor vessel and a l
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Appendix I 2k" discharge nozzle connected to a vent line leading to the suppression chamber. Other nozzles were provided for instrument-ation, venting and draining, and over-pressure niief. i The dry well had a contained volume of 1,100 ft.3, somewhat I larger than 1/112th of the present Bodega design. This was done ,2 t in case future design changes increase the size of the Bodega dry F vell. Actually, dry well size does not affect dry well pressure unless the vent piping is exceedingly small. Dry well size does effect suppression chamber pressure because the air in the dry well is largely or entirely forced into the suppression chamber. After the first few tests a deflector plate was installed in front of the discharge opening in the dry well to cause greater , dispersion of the steam-water jet coming through the inlet nozzle and increase air carry over to the suppression chamber.
- h. Suppression Chamber I l
The suppression chamber was contained within a vessel 12 ft. I.D. and h9 ft long over heads. The vessel was partly buried in the , ground and was the same vessel used for this purpose during the Humboldt tests. 3 The suppression chamber itself, see fig. 3, was a section of the vessel described above, extending across the diameter of the vessel from.the top down to 28 ft. below the top. The sides of the section vere parallel and 3'-8" apart. The bottom was curved with a 13' radius. The sides and bottom of the suppression chamber were steel plates. The rest of the vessel was filled with concrete. i i l The suppression chamber contained 670 ft.3 of air and 339 ft.3 of water. The air volume was 1/112th of the total air volume of the Bodega design and the water volume was 1/112th of the effective water volume of the Bode 6a design at the time the test facility was designed. At present the Bodega suppression chamber is somewhat larger than it was when the test facility was designed, providing more water and air ( space. Therefore the test suppression chamber was being worked harder { than the Bodega chamber vould be. The water volume in the test facil-ity was that amount of water associated with one vent pipe in the Bodega design. Part of the water in the Bodega suppression chamber lies under the eight veut pipes coming from the dry well and for test p.trposes was not considered available for condensing st.esm. However, all of the air in the suppression chenber was considered available for absorbing the pressure rise following an accident. \ 1
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5 Vent Pipe i
. 1
,, The vent pipe'vas a 24.in. Pipe connecting the dry well to the
-suppres'sion pool. Its length was 45 ft. and it contained two tees
.and one ell. The normal submergence of the end of the vent pipe below the pool surface was four feet.
The diameter, length, and flow resistance induced by fittings were the same as that of an average vent pipe in the Bodega design.
- 6. Rupture Disk Assembly A 10 in, double rupture disk assembly together with an orifice or nozzle was used to simulate a pipe break. The assembly was located ir.cediately downstream from the test orifice or nozzle which in turn was located on the discharge flange of the test reactor vessel. The assembly contained two rupture disks, each rated to break at 900 psig at 575' F. With test reactor . pressure at 1,250 .
psig and a gas pressure of about 650 psig between the two disks, the disks vould hold. When the gas pressure was vented off, the upstream disk would break, followed closely by the downstream disk and the test vould be under way.
,,, Figure 4 is a photograph shoving the rupture disk assembly i
mounted in place. An orifice plate may be observed on the upstream , side of the assembly. Figure 5 shows rupture disks before and after testing and also the orifice pinte' simulating the E0A rupture size. 7 crifices and Norzles The simulated break size for each test was established by an orifice or nozzle which was machined in a steel plate mounted on the discharge flange of the test reactor vessel. Orifices vere machined with a sharp, square upstream edge, a 1/16 inch flat section, and the downstream face beveled at an angle of 60' from the axis. An orifice with a diameter of 3 24 in, diameter simulated the break area associated with the EOA, that is, it had an area 1/ll2th of that of a double ended break of 28 in. 0.D. Pipe. Other orifice sizes tested - had diameters of.905 in.,1.66 in., 2.kB in., 3 74 in., 4.50 in., and 3 12 in. The nozzles were made with an entrance radius and straight throat length each equal to 1/4 of the throat dicmeter, which approxi-mates the shape of the entrance to the Bodega reactor vessel recircul-ation outlet lines. Nozzle diameters tested ven .905 in., 2.00 in., 3 24 in., and 4 50 in. Additionally, the 3 24 in. nozzle was tested with the straight throat section mmoved, leaving the entrance radius
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- 8. Instrumentation l i
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- Figure 6 shows schematically the arrangement of the instrument- I ation and figure 7 shows some of the instrumentation in the control ,
shack. The pressure transducers were of the strain gage type and their output was recorded by a light beam oscillograph. The temper- , ature transducers were thermocouple and'their' outputs were recorded j by a multi-point strip chart recorder. Certain principal pressures ' were also indicated by gages in the control shack. D. Method of Performing Tests The following were the principal steps in performing a test on this facility: (
)
- 1. The desired test orifice or nozzle together with the double rupture !
disk assembly would be installed on the discharge flange of the reactor vessel.
- 2. The suppression chamber would be filled with water to the desired
- 1cvel. The desired level would be established by overflow from a nozzle in the side of the suppression chamber.
- 3. The reactor vessel would be filled with water until the desired level was established in the gage glass on the reactor vessel. ,
- 4. Saturated steam from a 1,400 psig source would be admitted through a
~s nozzle in the bottom of the reactor vessel to heat the water. During this time
~~
I the reactor vessel would le vented to remove any non-condensible gases and { drained to maintain the desired water level. Heating up to a pressure of 1,250 ) psic would take about one to one and a half hours. 4a. Several tests were run with the water in the reactor vessel initially subcooled. For these tests (Tests Nos. 39 to 45) subcooling was accomplished l by heating up to a pressure less than 1250 psig corresponding to the water j temperature desired. The reactor was then pressurized to 1250 psig by admitting ' steam from the 1400 psig steam source to a nozzle in the top of the reactor vessel. '
- 5. When reactor pressure reached about 600 psig, nitrogen would be admit-ted to the space between the two rupture disks to maintain a pressure in this 1 space equal to about one half of reactor pressure. As reactor pressure increased, I pressure in this space would be increased.
- 6. When the desired test pressure in the reactor vessel was reached, all vents, drains, and other openings in the reactor vessel, dry well vessel, and suppression chamber would be closed or checked to be closed.
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- 7. Final instrument checks would be made and the oscillograph started l (the temperature recorder would already have been running during the reactor l vessel heating part of the test.) ;
7a. Several tests were run to determine the effect of pre-purging the dry well. For these tests (Tests Nos. 39 to 45) steam was admitted to the dry "J well by opening a valve connected to the 1400 psig steam source. This steam was permitted to flow until the suppression chamber pressure stopped rising, indicating air purging had ceased. This valve was then closed and the test
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- 8. The nitrogen prc.ssure between the rupture disks vould be vented off and the disks would break. During the blevdown from the reactor vessel, ga6es in the control shack indicating dry well and suppression chamber '
pressures would be observed as a check on the results recorded by the oscill-ograph. 9 The various test vessels vould then be vented and preparations for the next test would begin. Frequently these would include recalibration of the pressure transducers which proved to be quite stable throu6hout the test program. 1 I E. Test program ( (
- l. Start-up Tests i
Table I is a listing of principal test parameters and results for cach test. The first six tests were made with the suppression
~c henber open and with increasing orifice size and reactor water volume j to check out the adequacy of the test facility itself. During this series of tests two failures cccurred which caused the loss of most test results. During test No.1 a gasket on the upstream face of the orifice plate broke when the blevdown was initiated; this was probably due to improper tightening of the bolts on this joint. During test No. -l n' 4 the rupture disks broke unexpectedly for unknown reasons. These were '
. ,. the only incidents causing serious loss of test data during the entire test program.
- 2. Tests Without Darlector Plate Tests Hoc. 7 through 13 vere run with the suppression chenber closed and with increasing orifice sizes up to 5 12 in. diameter.
Water levels in the reactor vessel and in the suppression chamber vere at normal levels corresponding to Bodega design. Operation was satis-factory for all tests, however suppression chamber pressure was lower than predicted. This was caused by the failure to expel all the air frcm the dry well during the test. To increase the air carry over, a defl.. tor plate was installed in front of the outlet nozzle in the ! dry well so that the jet of steam-vater mixtures coming from the inlet { nozzle vould be dispersed, causing more air to be purged from the dry well. 3 Tests With Deflector Plate Tests Nos, lh and subsequent were run with the deflector plate installed. Test results indicated most (about 907o) of the dry ; vell air was being forced into the suppression chamber. 1 These tests were run with a variety of orifices and nozzles.
,'i Most tests were run with all reactor water at or very near saturation te=perature ~- the suppression chenber filled to a depth of 11 feet, and the dry well
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A. Test 22 was run with the water in the bottom of the reactor vesse'l subcooled 16'F below saturation, in order to simulate subcooled water in thp bottom of the Bodega reactor vessel. ! B. Tests 23 and 25 were run with the suppression pool level respectively raised and lowered one foot to simulate possible variations ; in the Bodega pool level. ' C. Test 24 was run with the dry well preheated to 150'F by inject-ing steam into it before running the test. This was to simulate the maximum expected dry well temperature at Bodega during plant operation. (Note: During the Humboldt testing, the dry well was heated to 150'F for several tests. No significant difference was observed between tests with a heated and an unheated dry well.) During all the tests run with the deflector plate, suppression chamber pressure never exceeded 30 psig and the dry well pressure never exceeded 38 psig with an orifice or nozzle of MCOA size except for the special tests described in the next paragraph. The highest dry well pressure achieved was 63 psig with an orifice 250% of MCOA break size. The special initial conditions for the four special tests described above had no significant effect except for the test with a preheated dry well , fg where pressures were lower due to loss of air in the dry well because of _ preheating it with steam.-
'. Teste With Dry Well Pre-Purging and Subcooled Water in the Reactor Vessel Tests No. 39 through 45 were run with an initial dry well pre-purge and with reactor vessel water subcooling ranging from 25 to 110'F. Results of these tests are tabulated on Page 2 of Table I (page 7a). The maximum dry well pressure does not follow any particular trend. However, the test-with 90*F subcooling gave the highest pressure drop in the vent piping.
The variations in dry well pressure appear to result principally from differing amounts of air in the suppression chamber resulting from variations in the initial temperature.
- 5. Representative Pressure Traces '~
Figures 8, 9,10, and 11 show the pressure traces for six tests. Figures 8, 9 and 11 are principal pressures for two tests replotted for clarity. Figure 10 is a reproduction of the oscillograph records for three other tests. Some characteristics of these traces are discussed in the next section. Test Results
- 1. Reactor Vessel Pressure j For all tests, reactor vessel pressure was at or near 1,250 psig initially. After the rupture disks broke there was a sharp drop in 3 pressure for all tests, the amount of the drop increasing with orifice I size and amount of initial reactor vessel water subcooling. This was 3
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I O followed by a short period of fairly steady pressure and then a graduil pressure decrease which was in turn followed by a more rapid rate of decrease. The initial drop in pressure is assumed to be caused by:, i (1) A brief delay between initiation of flow and start of flashing of the water in the vessel, and (2) The need for sufficient pressure loss to establish the rate of flashing that corresponds to the flow rate from the vessel. The change from a gradual pressure decrease to a more rapid rate of l decrease is assumed to occur at the time that all water has been expelled '
,from the vessel; the more rapid rate of decrease. resulting from the fact that there is no more water to flash and help maintain pressure. The time after start that this change occurs is assumed to be the duration of 3 j
water flow for purpose of calculating flow rates. I All these characteristics were also observed during the Humb61dt tests.
- 2. Suppression Chamber Pressure The' suppression chamber pressure trace for Tests Nos. 1 through 27 showed an !
initial sharp rise, generally to about 12 or 13 psis, followed by a slight ! drop in pressure; then pressure would rise gradually reaching a maximum at about i the time all water and steam was expelled from the reactor vessel. This gradual { q pressure rise was ' smooth for the smaller orifice sizes and was accompanied by -l
, some pulsations for the larger orifice sizes.
The initial sharp rise in pressure followed by a slight drop was observed during the Humboldt testing. Visual observations of suppression pool action were also made during the Humboldt testing and this pressure characteristic was attributed to the observed violent but brief upward surge of pool water caused by a lar8e blast of air coming over from the dry well at the start of the test and resulting in momentary compression of the air in the
' suppression chamber air space. Visual observation was not considered practical during the Bodega testing but it is considered that this pressure characteristic is due to the cause described above.
Following the initial sharp rise in pressure, the gradual rise is caused by continued purging of the remaining air from the dry well and heat up. Before the deflector plate was installed in the dry well, this purging w}}