ML19317F557
| ML19317F557 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/12/1970 |
| From: | Sampson G TOLEDO EDISON CO. |
| To: | |
| References | |
| NUDOCS 8001150878 | |
| Download: ML19317F557 (57) | |
Text
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r ?,jisamt, .,& DAVIS-BESSE NUCLEAR POWER STATION PRELIMINARY SAFETY ANALYSIS REPORT . Amendment Number 5 , l ( l
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THE z
)^\\pp TOLEDO
' \'D EDISON COMPANY ,
"0011s0 8 78 gl4
May 12, 1970 (, APPLICATION FOR LICENSES FOR DAVIS-BESSE NUCLEAR POWER STATION Docket No. 50-3h6 Amendment No. 5 Enclosed herewith, amending and supplementing the above entitled application, are responses to 15 questions which vere not fully answered in Amendment No. 3, additional infor=ation requested informally and minor corrections and additions as , listed on the attached Index Sheet for insertion in the Preliminary Safety Analysis Report, as previously filed. THE TOLEDO EDISON COMPANY By m , m_ y_ _ Vice President, Power Sworn to and subscribed before me, this 12th day of May,1970. t
$%-u d t /! >_
Notary Public "\ GENEVAI.LEAKE Notary Peb'ic. b:n County, Ohio My Commission Expires Sept. 2.1974
p D-B Section a P,aje, 3.4.2.4 Procurenv.nt 1B-21 3.4.2.4.1 Equipment Manufactured by Babcock & Wilcox 1B-21 3.4.2.4.2 Equipment Purchased by Babcock & Wilcox 1B-21 3.4.2.5 Erection 1B-22 3.4.2.6 Plant Test Program 1B-22i
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3.5 DESIGN CONTROL 1B-22 3.6 PROCURENENT DOCUMENT CONTROL 1B-22 3.7 INSTRUCTIONS PROCEDURES & DRAWINGS 1B-22 3.8 DOCUMENT CONTROL 13-23 3.9 CONTROL OF PURCHASED MATERIAL, EQUIPMENT, AND SERVICES 1B-23 3 3.10 IDENTIFICATION AND CONTROL OF MATERIALS, PARTS AND COMPONENTS 1B-23 ( 3.11 CONTROL OF SPECIAL PROCESSES 1B-23 3.12 INSPECTION 1B-23 3.13 TEST CONTROL 1B-24 3.14 CALIBRATI: 10F MEASUREMENT AND TEST EQUIPMENT 1B-24 3.15 HANDLING, STORAGE, SLIPPING, AND PRESERVATION 1B-24 3.16 INSPECTION, TEST, AND OPERATING STATUS 1B-24 3.17 NON-CONFORMING MATERIAL, PARTS OR COMPONENTS 1B-24 3.18 CORRECTIVE ACTION 1B-25 3.19 QUALITY ASSURANCE RECORDS IB-25 3.20 AUDITS 1B-25 IB-iii Amendment No. 3 l
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g, r THE TOL EDO E DaSON CO E NGINE E R4 NG e CONSTRUCTION MAN AGE ME N T $ M AN AGE ME NT PROCURE ME NT n i I
...............................(.
3......... ......... CHIEF STORT UP E NGINE E R CHIE F E NGINE E RS PROJE CT ,,f MGR.OA ... METALLURGY ... PROJECT CONSTdUCTION INSP. EXPO SPE CI ALISTS E NGINE E R Q A COORDINATOR e QUA LITY CONTROL MANAGER SUPV. SUPv. I I I l !* I I : QA !
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STGRT tJP DESIGN ENGINE E R g l l ~ E NGIP5EE RS l E NG. SUPV. ! e e I f I" a ~l. I----- ARCHi f E CTUR A L
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E NGINE E R F eE LD ENGINE E R ENGR. ENGR. SUPTS. SUPV. EXPD.
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1----. F LE CTRICAL e-SE NIOR l l FIE LD E NGINEE RS I I 1 l8 l-_ ME CHANICAL f l [ ~ t I----- l FIE LD ENGINEE RS . . . . TftO ASST i .TECO uUCtE AR l . COuTRAcTORS . . . . - SUPiS I SuPPuE RS I l Fif LD E NGINE E RS l---. LAYOUT SUPPUE RS -----------! I D RW S,SPE CS, TEST LASS O t IST. E TC LEGEND RESPONSIBILITY
- -- Q A SURVEILLANCE & AUDIT
- - -- INSPECTION SURVEILLANCE
---------- COORDIN ATION DAVIS-BESSE NUCLEAR POWER STATION BECHTEL QUALITY ASSURANCE ORGANIZAIl0N CHART FIGURE 18-2 allen 0 MENT NO. 2
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movement across the site is negligible. If the existing groundwater regime is not altered during construction and operation of the plant, significant solution of gypsum Cannot occur. The quantities of sulfates in the groundwater (1000 to 1700 ppm) and in the bedrock at the site have caused deterioration of ordinary concrete. The design of concrete which vill be in contact with groundwater and bedrock must take this into account. 3 Solution cavities deuression area
- a. Solution cavities in the surface ~
Solution cavities were explored in the surface depression area approx h00 ft south of the location of unit No. 1. The report of our detailed study of the surface de-pression area " Conclusions and Findings, Test Excavation Pro-gram" dated 1 October 1969 is on file with the Toledo Edison Company. y The elevation of the ground surface in the depres-sion area is 1 ft to 2 ft lover than the average elevation of the ground surface in area A shown in Fig. IV-2. Typically, the surface conditions in the surface depression area consist ' of the following soil deposits and rock units. From el 57h (ground surface) down to an elevation between 561 and 553, soil deposits consisting of a glacio- ' lacustrine deposit and a till deposit were encountered. The thickness of the glaciolacustrine deposit varies from 8 ft to 12 ft; its thickness was found to be approximately four feet greater over the cavities than beyond the cavities. The thickness of the till deposit varies from 5 ft to T ft; its thickness was found to be nearly uniform over and beyond the cavities. In general, the bedrock surface was encountered at el 561, approx 13 ft below ground surface. In some loca-tions, where the upper portion of the bedrock collapsed into underlying cavities, the surface of the broken bedrock was encountered several feet lover; e.g. at el 553 in sections 2 and 3 of the test excavation. In the shaft of section 2 of the test excuration, the bedrock was excavated to el 534 Three main rcek units were encountered in the excavation. These rock units are described in the following table. ( 2C-57
Unit Approx El of Thickness ' No. Top of Unit
- ft Description I 561 7 Massive Dolomite. Hard and brittle. Highly jointed, es-pecially over the underlying cavities where.the rock col-lapsed. Joints are open, approx 1 in. max. Rock is veathered near joints.
II 55h 6 Dolomitic Shale and Argilla, ceous Dolomite (Bedded Dolo-mite). The top of the unit is - veathered to silty clay and rock chips. The bottom of the unit is harder and contains vugs less than 0.25 in. in diam-eter. In some locations, the beds are distorted over the underlying cavities. 1l III 548 1h Gypsiferous Dolomite (Bedded Dolomite). Typically contains 20% of gypsum. Solution oc- ) curred in this unit forming ' fissures and cavities; some of them causing collapse of over-lying rock units and soil deposits. Significant fissures and cavities were observed in the 1 gypsiferous dolomite down to e? 53h. These fissures and cavities result from the solution of gypsum and dolomite. Solution appears to have started along joints, forming fissures from a few inches to a few feet vide and several feet high; e.g., two feet vide and ten feet high. At joint intersections, some fissures have enlarged forming cavities several feet in diameter; e.g., ten feet in diameter. In some cases, units II and I, and the till deposit collapsed into the under-lying cavities; e.g., at sections 2 and 3 of the test exesva-tion. In sections 2 and 3 of the test excavation, approx 30% of the volume of the fissures and cavities was empty; the remaining 70% contained soil and rock fragments.
- Elevation is that of the portion of the unit unaffected by the cavities j 2c-58
4 D-B r- APPENDIX 2D
\ LIMHOIOGY PROGRAM FOR DAVIS-BESSE NUCLEAR POWER STATION 7
The lake study program as outlined in this appendix is being conducted under
- the direction of The Toledo Edison Company and the Great Lakes Research Division, Institute of Science and Technology of The University of Michigan was retained by Toledo Edison in 1968 to conduct this study. Dr. John C.
Ayers, Research Oceanographer, of the Great Lakes Research Division is in charge of this study. At our request, a meeting was held in the fall of 1968 with representatives of the Division of Wildlife, Ohio Department of Natural Resources to review the study outline so that all areas of concern to this State Agency could be taken into account and for us to gain the benefit of any State sponsored studies that had been done or that were projected. Since this meeting, other - State and Federal agencies have expressed an interest in tne study and as a result there will be participation by some of these agencies. While the full degree of partleipation of each agency has not been finalized, the results of several meetings have indicated that a joint program will follow a resulting in a very satisfactory arrangement. This_ lake study program is well underway and Dr. Ayers began collecting field data in July, 1968. It is expected that the report covering parts A, B, C, D and E of the attached program will be submitted in October,1969 At that time a sumwy of the report will be transmitted to the Atomic Energy Commis-sion as Appendix 2D of this Preliminary Safety Analysis Report. Sampling for parts F and G (Ecology and Radionuclide Reconcentration) vas begun in June, 1969 Further sampling will be done in October 1969, May 1970 and will be continued prior to station operation to identify any trends in radioactivity levels or biological populations in the lake environment. In summary, The Toledo Edison Cocpany has initiated a comprehensive lake survey program to provide background data for the design and operation of the Davis-Besse Nuclear Power Station.' The program is being coordinated with State and Federal agencies ani will first be reported on in October,1969 All reports prior to and fol]uing operation will be distributed to inter-ested parties. An outline of the pregram follows. LAKE STUDY PROGRAM FOR THE DAVIS-BESSF -IF' CLEAR F0WER STATION I PURPOSE The purpose of the study is to gather lake data. It will be used to establish certain station design criteria and as a design aid to control 1 or minimize any possible adverse affects upon Lake Erie from construc- l tion and operation of the proposed Davis-Besse Nuclear Power Station. II SCOPE The program is to include a description and evaluation of past, present, and projected future lake use, a field investigation to determine physical, 2D-1
D-B chemical, and biological characteristics of the offshore lake regime and to evaluate the effect of station effluent on aquatic life. %. 3 III SCHEDUIE The gathering of data began in the summer of 1968 and, where necessary to establish trends or background, will continue until operation of the station. IV OUTLINE OF STUDY The following is a brief description of each portion of the Lake Study which we feel is required to satisfy the requirements of the AEC and other interested State and Federal agencies. A. General Background ,
- 1. A description of Lake Erie, both regional and local area around plant site. This will include size, average depth, normal height of lake surface above sea level, general current and temperature data, and a description of nearby streamsand/orriversdischargingintotheLake.
B. Lake Use
- 1. A description of how the lake has been utilized in the past and is presently being used for fisheries, resources and 3 recreation with.2 the general vicinity of the site. Recrea- j tion in terns of boating, sportfishing, bathing, and esthetic value vill also be described.
C. Lake Survey Program
- 1. Initial cruise surveys will determine lake bottom topography for an area opposite the plant site and extending one-half mile up and down shore from-the property boundaries. The survey vill extend into the lake to a minimum water depth of 12 feet.
- 2. Current patterns during the particular site wind regimes will be determined. Water conductivity and temperattre profile measurements will be conducted. Limited dye studies and release of drift cartis may be necessary to gs.ther dif-fusion data. Consideration will be given to monitoring the influence of water discharging into the lake from the Toussaint, and in particular the Portage River.
D. Diffusion Studies The diffusion studies will:
- 1. Describe factors which influence station site diffusion such as:
2D-2
f- D-B a Water Currents - direction and speed. b Upvellings c Sinkings - possibly negligible in shallow water. d Wind speed
- 2. Determine Dilution Rates by:
a Dye studies b Drift cards c Other 3 Evaluate Plant Effluent:
~
a) General considerations - temperature and controlled radioactivity of plant water. b) Discussion of basic mechanisms of circulating water plu=e. E. Wind Tides, Waves, Storm, and Ice Conditions Describe and evaluate the effects of each as they relate to the proposed circulating water system, shore protection, and plant site flooding. , F. Lake Ecology Describe aquatic life taken from the area for human consumption. Determine total and yearly commercial take and project increase or decline of com=ercial fishing industry. Conduct a sampling pro-gram to determine species, population, at d distribution of bottom organisms, plankton, and fish. Backgrouxa levels of radioactivity and stable elements will be obtained fo tottom sediments, bottom organisms, plankton, fish, and the water. G. F g onuclide Reconcentration Lac, oratory culture of local aquatic organisms in vater containing typical radionuclides that could be released in the lake during operation of the station, vill be done to detemine the degree to which these organisms concentrate the radionuclides. \ 2D-3
5.9.8 CIPCULATING WATER CANALS f Two canals, namely, the intake and discharge canal, will be constructed. 1 A built-up earthen dike on-shore and an off-shore dike constructed to protect 2 against wave erosion as well as ice build-up will be constructed to establish flow and alignment of the canals. Approximately 700 feet of the intake canal's excavated portion and built-up , earthen dike in front of the intake structure vill be designed and constructed as a Class I structure. The Class I dikes will be constructed of soil excavated in the dry. The dike I I slopes are designed to be 3 (hor.): 1. Analysis of the stability of the dike slopes have been made under static and dynamic cenditions. Even under the 2 worst, hypothetical conditions, that is assuming that a rapid draw down of the canal water to elevation 561' (I.G.L.D)* simultaneously occurs with the maximum 3 possible (larger) earthquake, the factor of safety of the dike slopes was cal- ~ culated to be 150 for a shear strength of the dike soil equal to about 500 lb/ft .2 Based on an analysis of the site soils, the representative shear 3 strength of the soils in the completed dikes vill exceed this value. For a canal water level at elevation of 560 (I.G.L.D), and an effective depth of 12 feet frcm elevatien 560 (I.G.L.D) to the service water pump inlet, apprcximate volume of 10 million gallons will be available within the Class I canal area limits. For circulating water pumps the low level alanns eldvation is 562' (I.G.L.D) and lov level autsatic shut-off elevation is 560' (I.G.L.D). 2 During an emergency shutdown, the decay heat will be removed at a centrolled 1 rate by use of the cuxiliary feed pumps. The station vill be kept at " hot stand-by" for a minimum of 13 hours followed by a 6 hours cooldown period to Icver reactor coolant system temperature to 280 F by using 250,000 gallons of 3 stored cendensate storage vnter. After this initial 19 hours, the decay heat removal system operatien vill be 1 started utilizing the service water pumps located in the intake structure and 2 l pumping from the intake canal. 1 The 10 million gallons of stored water in the intake canal forebay will provide l 2 sufficient ecoling surface to continue cooling the statien by evaporation for at least lh days. Thic should provide sufficient time to re-establish direct water flow cccmunicaticns between the lake and the station intake structure via the 1 l intake canal. The design incorporates a Class I return line frem the service water system to 3 I the intake canal Class I area forebay. Accident Analysis for Loss of Intake Canal is discussed in detail in Section I lb.2.2.6. 3
- International Great Lakes Datum (I.G.L.D)
( I j 5 h2A Amendment No. 3 i
i l D-B APPENDIX 5A DESIGN BASES FOR STRUCTURES, SYSTEMS AND EQUIPMENT 1 GENERAL The design bases for structures for normal operating conditions are governed by the applicable building codes. The design bases for specific systems and equipment are stated in the appropriate PSAR Section. The basic design cri-terion for the design basis accident and seissf c conditions is that there be no loss of function if that function is related to public safety. The design of structures and facilities conforms but is not limited to the applicable codes and specifications listed below.
- 1. Uniform Building Code (UBC), 1967 Edition. -
- 2. American Institute of Steel Construction (AISC) " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings" - Sixth Edition.
- 3. American Iron and Steel Institute (AISI) " Specification for the Design of Light Gage Cold-formed Steel Structural Members" -
1961 Edition.
- 4. American Concrete Institute (ACI) " Building Code Requirements for
( Reinforced Concrete" - (ACI-318-63).
- 5. American Welding Society ( AWS) " Standard Code for Arc and Gas Welding in Building Construction" - (AWS D1.0-66 and AWS D12.1-61)
- 6. API Specification No. 620 & 650 for Welded Steel Storage Tanks.
T. American National Standards Institute, B 96.1 - 1967 " Specification 3 for Aluminum Alloy Tanks". ~
- 8. ASME Boiler and Pressure Vessel Code, Section III, Nuclear Vessels, Class B, governs the design and fabrication of the containment vessel and penetrations.
9 AEC Publication TID 702h " Nuclear Reactor and Earthquakes" - governs the seismic design of all Class I structures.
- 10. American Society of Civil Engineers (ASCE) Paper No. 3269,
" Wind Forces on Structures" - governs vind design requireme'nts.
- 11. American Association of State Highway Officials (AASHO)
" Standard Specifications for Highway Bridges" - Ninth Edition 1965
- 12. Nuclear Energy Property Insurance Association - Mutual Atomic Energy Reinsurance Pool (NEPIA - MAERP), Basic Fire Protection for Nuclear Power Plants.
i
- 13. American Concrete Institute (ACI) " Specification For the Design and Construction of Reinforced Concrete Chimneys" - (ACI-307-69). 3 5A-1 Amendment No. 3
f . D-B 2 CLASSES OF STRUCTURES, SYSTEMS AND EQUIPMENT CLASS I Class I structures, systems and equipment are those whose failure could cause uncontrolled release of excessive amounts of radioactivity or those essential for immediate and long-term operation following a design basis accident. When a system as a whole is referred to as Class I, portions not associated with loss of function of the system are designated as Class II. The following are typical Class I structures: Shield building Containment vessel and penetrations Containment interior structures Station control room Spent fuel storage pool, including transfer tube Portions of auxiliary building housing the engineered safety features or other Class I equipment Supports for Class I equipment ~
~
Cable spreading room ~
')
Switchgear rooms Reactor vessel primary shield and support structures Diesel generator room and fuel storage and supply facilities Part of intake structure containing Class I service water pumps l Station battery room Borated water storage tank i Structures for emergency water supply (if required due to loss of cooling water) The following are typical Class I equipment and. systems: Reactor vessel and internals including control rods and control rod drives Reactor coolant system components (steam-generators, pressurizer, pumps, etc.) and interconnecting piping and supports All piping and connections to the primary system to and including s the second isolation valve SA-2 L.
D-B Containment isolation valves and penetrations [ Main steam and feedvater piping from generator to and including the _ isolation valve 3 Atmospheric dump and main steam safety valves and associated piping Auxiliary feedvater pumps and system Puel storage facilities including spent fuel and new fuel storage equipment la Diesel generator system Containment vessel crane (unloaded condition) . Station battery system Emeldency buses and other electrical gear to and including power equipment required for safe shutdown Service water system, parts which service Class I systems Component cooling water system serving vital NSS heat exchange equipment . k Containment spray system Containment air recirculation and cooling units - Lov pressure injection and residual heat removal system 1 I Core flooding system 2 Makeup system 2
, High-pressure injection system Condensate storage tank (except tornado and turbine missiles) 1 1
CLASS II Class II structures, systems and equipment are those whose failure would not result in the uncontrolled release of radioactivity and would not prevent 3 reactor shutdown. The failure of Class II structures, systems and equipment may interrupt power generation. Class I equipment and systems located in Class II structures have reinforced concrete enclosures designed to withstand the loads for Class I structures. _ 3 DESIGN BASES All steel structures are designed by the Working Stress Method. All major ) reinforced concrete structures are designed or checked by the Ultimate Strength 3 , Method. 5A-3 'k Amendment No. 3 j 1
D-B n 4 NOTATIONS I U = Required ultimate load capacity. D = Dead load of structure and equipment plus any other permanent loads contributing stresses, such as soil or hydrostatic loads. An allowance is also made for future permanent loads. L = Live load and piping load. R = Force or pressure on structure due to rupture of any one pipe. T = Thermal loads due to temperature gradient through wall under operating conditions. H = Force'on structure due to thermal expansion of pipes under operating' conditions. TA= Thermal loads due to temperature gradient through wall under accident conditions. HA = Force on structure due to thermal expansion of pipes under accident conditions. E = " Maximum Probable (%11er) Earthquake" resulting from ground " -s 2j surface acceleration of 0.08 g. E'= " Maximum Possible (Larger) Earthquake" resulting from ground surface acceleration of 0.15g. W = Wind load. (Wind velocity 90 mph at 30 ft, a'uove ground. ) See ASCE 3269 for increase due to gusts and height. W'= Tornado load including differential pressure.
@ = Capacity reduction factor. (Defined in ACI-318-63 Code, Section 1504.) See Appendix 5D for discussion.
f s= Allowable stress for structural steel. (Defined in AISC, Section 1 5) Fy= Yield strength for steel. 5 SEISMIC ANALYSIS For seismic loads, Class I structures and components are designed using the " Response Spectrum Curves" shown in Figures III-5 and III-6 of Appendix
- 20. Seismic accelerations are determined on the basis of dynamic analysis.
Wherever it is considered impractical to make a dynamic analysis , the ==vinnna value of the acceleration response curve for the appropriate damping factor is used. Damping factors to be used in determining response acceleration are shown y in Table SA-1. Amendment No. 2 5A-4
u-w i l D-B TABLE OF CONTENTS Section Pm 6 ENGINEERED SAFETY FEATURES 6-1 6.1 EMERGENCY CORE COOLING SYSTEM 6-2 6.1.1 DESIGN BASES 6-2 6.
1.2 DESCRIPTION
6-2 6.1.3 DESIGN EVALUATION 6-4 l J 6.1.3.1 Failure Analysis 6-5 6.1.3.2 Emergency Injection Response 6-6 6.1.3.3 Special Features 6-10 6.1.3.h Check. Valve Leakage - Core Flooding System 6-10 6.1.4 TEST AND INSPECTIONS 6-11 6.2 CONTAINMENT ATMOSPHERE COOLING SYSTEMS 6-13 6.2.1 DESIGN BASIS 6-13 6.2.2 SPRAY SYSTEM 6-13 6.2.2.1 Description 6-13 6.2.2.2 Design Evaluation 6-14 6.2.2.3 Tests and Inspections 6-lk 6.2.3 CONTAINMENT AIR RECIRCULATION COOLERS 6-15 6.2.3.1 Description 6-15 6.2.3.2 Design Evaluation 6-16 6.2.3.3 Tests and Inspection 6-lT 6.2.h FAILURE ANALYSIS 6-lT 6.3 SHIELD BUILDING AND PENETRATION ROOM VENTILATION AND FILTRATION SYSTEM 6-20 6.3.1 DESIGN BASIS 6-20 6.3.2 SYSTEM DESCRIPTION 6-20 6-1
D-B Section Page
-')
6.3.3 DESIGN EVALUATION 6-20 6.3.k TESTS AND INSPECTION 6-21 6.3.5 FAILURE ANALYSIS 6-21 G
\
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_j i l i 6-11 1
D-B 6.2 ' CONTAINMENT ATMOSPHERE COOLING SYSTH4S 6.2.1 DESIGN BASIS The containment cooling system is designed to limit and subsequently reduce the containment vessel pressure in the event of a loss-of-coolant accident, and hence reduce the leakage of airborne activity from the containment vessel. The air recirculation cooging system has the same cooling capacity as the spray system, which is 150 x 10 Btu /h. Any of the following combinations of equipment will provide the same heat re= oval capability:
- a. Two containment spray pumps.
~
- b. Two air cooling units.
- c. One air cooling unit and one containment spray pu=p.
Two spray pumps and three air recirculation coolers are provided. 6.2.2 SPRAY SYSTEM 6.2.2.1 Description Removal of heat is accomplished by directing borated water spray into the I containment atmosphere. The system is sized to furnish the required cooling capacity if the air recirculating and cooling system is inoperative. The system consists of two half-capacity pumps , two half-capacity spray headers , isolation valves, and the necessary piping, instrumentation and controls as shown on Figure 6-5 The pumps and valves can be operated from the control room. Upon high containment pressure and emergency injection-actuation signal, the two pumps start, taking suction initially from the borated water storage tank. The containment spray system shares the borated water storage tank and the suction lines from the tank with the high and low-pressure injection aystems. After the water in the borated water storage tank reaches a low level, the spray pump suction is transferred to the containment sump. The containment sump water is cooled by the decay heat removal system as described in 6.1. Pump motor power is supplied from normal and standby sources with back-up supplied from the emergency diesel generators. Design data fer the spray system components are given in Table 6-3. Design data for components of the decay heat removal systems used in the recirculation phase are given in l3 Section 9 and supplemented by Figure 6-1. 6-13 Amendment No. 3
D-B 3l Table 6-5 Containment Spray System Performance and Equipment Data Component Containment Spray Pumps 2 Number 2 T' N , gpm 1,300 Motor Horsepower, hp Approx 250 Material SS Design Temperature, F 300 Spray Headers Number 2 2 Spray Droplet Size h,000 Micron Mean Piping Suction and Discharge SS Spray Headers (Downstream of Isolation Valves) -CS 3 6.2.2.2 Design Evaluation The system heat removal capacity is based on the spray water reaching thermal equilibrium with the steam-air mixture within the containment vessel. The containment spray system is designed to deliver full flow to the spray nozzles within 35 seconds after the LOCA. The spray pumps are located in separate rooms in the lowest level of the auxiliary building. This insures the availability of 50 percent of the system and also ensures sufficient NPSH to the pumps. 6.2.2.3 Tests and Inspections Components of the containment spray system are tested on a regular schedule as follows: Containment These pumps are tested singly by closing the Spray Pumps spray headers valves and opening the valves in the test line. Each pump in turn vill be started by operator action and checked for flow to each of the spray headers. This test will also verify flow through each of the borated water storage tank outlet valves. Amendment No. 3 6-14
D-B Borated Water These normally closed valves are cycled by remote 1 Storage Tank operator action to insure isolation closure Outlet Valves Containment With the pumps shutdown and the normally locked Spray Isolation open block valves upstream closed, these valves
' Valves are each opened and closed by operator action.
Spray Nozzles When the unit is shutdown, air or smoke is blown through the test connections with visual o~oser-vations of the nozzles. Containment Sump These valves will each by cycled opened and Line Valves closed by operator action. 6.2.3 CONTAINMENT AIR RECIRCULATION COOLERS 6.2.3.1 Description The LOCA air recirculation cooling system is composed of two out of the three air recirculation units located within the containment vessel. These units are used for both normal and erergency cooling. The normal cooling function is discussed in Section 5.h. Each unit consists of a finned tube cooling coil and a direct driven fan. Cooling water for the air recirculation units is supplied by the service water
' system. The service water supply line for each cooler has a normally oper isolation valve. The discharge line for each cooler has a normally closed stop valve and a =odulating control valve in parallel with the stop valve.
The modulating control valve is operated by a signal frem a temperature con-troller in the unit air discharge to provide automatic temperature control for , the containment during normal station operation. In the event of a LOCA, the emergency injection actuation signal will open the stop valve in the discharge service water line for full water flow. The containment cooling water system is shown in Figure 6-6. The air recirculation coolers are shown in Figure 6-6 and Figure 6-7 The rated capacity of each cooling unit is 75 x 106 Btu /h at design post-accident conditions. The design data for the air recirculation units are shown on Table 6-6. l3 1 6-15 Amendment No. 3
D-B Table 6-6 3l ] Containment Air Recirculation Unit Performance and Equipnent Data (Capacities Are for Single Components) Duty Equipment Data LOCA Operation Normal Operation No. of Units Installed 3 3 No. Required 2 2 Peak Heat Load, Btu /h 6 75 x 10 1.80 x 106 Fan Capacity, cfm (Total per Unit) 58,000 117,000 Containment Atmosphere Inlet Temperature, F 265 10k Cooling Water Flow, gpm 1600 540 Fan Speed Half Full 6.2.3.2 Design Evaluation In establishing the design of the cooling system, the following factors have been considered:
- a. The air recirculation units are located outside the secondary g shield at an elevation above the water level in the bottom of ~)
the contain=ent vessel at post-accident conditions. In this location the units are protected against credible missiles and from being flooded.
- b. All equipment, piping, valves and instrumentation associated with the air recirculation and cooling system is designed to withstand the temperature and pressure transient conditions resulting from a LOCA and the seismic forces resulting from the applicable earthquake.
- c. The failure of the normal and auxiliary electrical power supply will automatically connect the air recirculation units to the emergency electric power from the diesel generators.
- d. The fans and motors are designed to operate under normal con-ditions at full speed, and at half speed during LOCA conditions.
The motors are directly connected to the fan shaft.
- e. Excessive service water flow due to leakage from the coil is annunciated in the control room by high water level in the cooler condensate sump.
- f. The components of the system are used during normal operation and hence are continuously monitored for availability for post-accident operation.
1 J Amendment "o. 3 6-16
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A B B R EVI ATIO N S CIS - CONTAINMENT ISOLATION SIGN AL MOD- MODULATlHG DAMPER 9 bC N C- NORM AL LY CLOSED PE N ET R ATION ROOM g Nm_ _ Q SHIELD BLOC 2. CONT AINM ENT VESSEL r__. _________________________ ____ e 4 mo , A ,1
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LA _ _ _ F_ __ _ __ _ _ _ ._ J AMENDNENT NO. I
r 1 0-B i 9 AUXIL! Atty AD FyEpcENCY SY3T49 f 6 The auxiliary systems required to support the reactor coolant system during normal Jperation of the Cavis-Pesse Station are described in the following sections. Some of these systems are described in detail in Section 6 since they serve as ensi-neered safaty features. The information in this section deals primiarily with the fanetions served darits nermal cperatien. Most of the components within these systems are located within the auxiliary building.
- hose systems with connecting piping betvaen the containment and the auxiliary building are equipped with containment isolation valves as described in 3.1.5.
Tne codes and standards used, as apprcpriate, in tne design, fabrication, and testing of ecmponents, valves, and piping are as follcws: ,
- a. AJME Boiler and Fressure Vessel Code, Secticn !!, Material Specifi-cations.
- b. A2re. Boiler and Pressura Vessel Code, Section III, Juclear Vessels,
- c. CG Sciler and Fressure Vessel Code, Secticn VIII, Unfirel Pressure Vessels and CG Iiaclear Case Interpretations.
- 1. A3ME Boiler and Fressure Vessel Code, Section IX, *delding qualifica-tions.
- e. Standards cf the American Society for Testing Materials,
- f. ANSI, B31.1 - 1967 3
- g. ANSI, B31.7 - Febrwry, l'e'M Draft
- h. Standards cf the American Institute of Electrical and Electrctics Engineers.
i Standards of the National Electrical Manufacturers Associati:n. J. Fyiraalic Institute Standards.
- k. 2tandards of T.balar Exchanger Manufactaers Association.
- 1. Air Moving and Conditioning Association Codes.
=, ANSI, B96.1, Aluminum Tanks. 3
- n. American Oear Manufmeturers Association Standards, i
! o. American iister Wcrks Association Standards.
*! To aid in reviewing the system aravings, a standard set of symbols and attrevia-tions has teen used and is sue:marized in Figure 9-1.
i k ' t 9-1 Amendment No. 3 i O O' O O g-n v t L. _a w-k
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?.1 mKI"5 A."3 PS f?ICATION SYSTE'4
).1.1 fESI2 3MES
).1.1.1 Prersi fystem Nnetien Tha system shewn on fif2re 9 2 supplies the reactor coolant system with fi'.1 and crerstiensl .aseup water; circulates seal water for the reactor coolant puns; receives, purifies, and recirculates reactor coolant system letdown to l provide water Taality and resetor coolant boric scid concentration control; and l 3I seco rs>iates temporary chsnges in the reTaired reactor coolant inventery. The l makeup and purifiestion system performs no emergency function. -
?. l.1. 2 letiown coolers The letdown coolers cool the letdown flow from reactor coolant temperature to a temperature suitable for demineralization and injection to the reactor cool.
snt pu-rp sesis. The maximum letdown flow is ret 2 ired for startup at 100 F/hr 1ste h core life 1.herein the reactor coolant borou concentration is reduced by sn smount ecrresponding to the change due to moderator te perature resetiv-ity deficit. Mest in the letdown coolers is rejected to the componant cooling system. 9.1.1. 3 Lat h n centrol letiown flow is established by use of s block crifice which is sized for the normsl puritiestion rste. However, during a startup cr shutdown phase when the reseter ecolant system is at icw pressure, the desired letdown flow is
=sintainad by the supple ental use of a parallel control valve along with the crifice. Both flow paths are slso used when high letdown flow is reTaired, e.g. , resctor coolant toron concentration adjustment.
3.1.1. L P2rificat ion Daminarslizer The leticwn flew is passed through the purificati;n deminerslizer to remve resetor ecolant 1 purities other thsn toron. "he purifiestion letdown flow to aasintain the resetor coolant water quality is eTaal to one reactor coolant vol-ume per R hours. Ewh purification demineralizer is sized tor the letdown flow rste as ret 2 ired for borcn concentration control. Eefer to Table 11-3 for tne tuimum anticipatad e72ilibrium fission product secumulation in the resetor coolsnt. 3.1.1. 5 va k m N ps
*he maxeup pu:=ps sre desi.med to return the letdow . flaw the the resctor coolant system ar1 supply the sesl wster flow to the resctor coolant pu. ps. ihe r.ormal flow espacity is etaal to the normal makeup ficw plus the seal water flow to the resctor coolant puqs. The pumps are sized to meet these require =ents with j one pu: p in operstion.
, ?.l.1.5 Fasl Paturn cmlers The sesl return coolers are sized to ramove the hest sdded by the makeup pump I
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. -I i i D-B and concentrated to 25% solids by weight. Neutralization of the evaporator feed may te necessary in order to obtain this high concentration level. The distillate goes to the miseenaneous vaste monitor tank vnere a final check 1 cn activity is made before stcrage for reuse cr release to the circulating water discharge. The evapcrator bottoms are transferred to a concentrate I tank where they are stored prict to dru=ning for off-sit i'spos al. Before any liquii is released from a vaste tank to the ciredation water system, a sample is taken and its levsl of activity is determined. If it does not meet established limits, it is recyclei until it does. A final check is made on the vaste as it is discharged through an in-line radiation monitor. If its activity exceeds preset values, an alarm is annunciated and isolation valves automatically shut off the discharge. Some mode of re-cycling is then necessitated. A record win be kept of both the a=ount and level of activity of all discharged effluent. The valves in the inlet line to a vaste tara are closed whenever the tank is being discharged so that filling and discharging to the circulating water system cannot be done simultaneously. This insures batch processing. 11.1.2.2 Caseous Waste Systes The Gaseors Waste System processes potentially radicactive hydrogenated and aerated vaste gases. A diagram of the system is shown in yigure U-3. Sources of hydrcgenated, radioactive vaste gas include the reactor coolant drain tard, $egasifier, the make-up tara and the quench tank. Hydrogen and fission gases stripped in the degasifier ani vented frcet other tanks, flow to the vaste gas header and then to the vaste gas surce tans. A nitregen blanket in this tank automatically maintains a slight positive pressure in the system. The gases are then compressed into one of three 1 decay tanks, which are sized for 30-60 days holdup. A sample is removed from the tank and its activity level deter =ined. If it is sufficiently low, the gases are discharged to the vent througn a HEPA filter. .f it is hign, the gases are allowed to decay until future sampling shows ttat they are suitable for controlled release to the etwePe*o The nitrogen cover gas displaced from the receiver tarAs is also handlei by this system. Since it should have little activity, an eff;rt is made to keep it separate from the hydregen and fission gases. "'his is done by forming one of the compressers and tanks into a separate processing chain. 1
"'he cover gas, that has been ccepressed, can be reused or, after sampling, vented, dhould any significant contaninatica occur, then it cast te precessed sisilarly to the nor=al hyiregenated vaate. If only ene ecmpressor is availatie at ar.y time, all gases win be conected at a co= mon header.
All released gases must pass a radiation monitoring system and if their activity exceeds a set point, an alarm is annunciated and the isclation valves in the disenarge line vill automatically close. Aerated radioactive vaste gases from the miscellaneous and detergent vaste drain tanks will be processed separately frcan the hydrogenerated vaste gases to prevent the pessibility cf axplosive mixtures. These low level gases are
, l , simply collected and released to the vent.
11-n I nn ' D 0
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j u.1.2.3 Solid Waste Systes j Solid wastes are placed in ICC-approved containers appropriate for the vaste material. i.caded containers are monitored for radiation levels and stored in a special area prior to shipment to an cff-site disposal facility. Radio-active spent resins sluiced from iemineralisers are collected and stored in the spent resin tana until a quas.: tty sufficient for disposal is accumulated. The tana is sized for at least a one-year accumulation and is arranui with pumpout connections for resin transfer to a shipping cast for dj spessi. Au soft solid wastes, such as contaminated clothing rags, wiping t<wels, paper, gloves, and shoe coverings will be compressed into the containers by a baler. Hard solids such as wood, metal, glass, plastics, cencrete and ceramics win be put into the containers without compressing. Items too large for tne containers will be stored, packaged and disposed of as appropriate. 11.1.3 E31:3 EVALUATION The possibility of an accidental release of activity from the radvaste system is minimized by reuse of much of the liquid wastes. I.iquid and stored gaseous wastes are sampled prior to aischarge to the enviros.nent. Solid wastes are disposed of by licensed contractors in accordance with ICC regulations. All liquid radioactive vastes flow to storage tanas prior to discharge to the environment and cannot be discharged to the environment by gravity (i.e., the effluent must be pumped out). An actua<,or operated valves which control the discharge of radioactive materisi into the environment fail in the closed pcsition on loss of actuating force or signal. Radioactive gases are continuously monitored during discharge in compliance with requirements of 10 CFR 20. Standby units (pumps, ion exchangers, and compressors) permit continueus processing in the event of equipr'r failures c7 routine maintenance. 11.1.1 * *r AvD IMfE MIO E Functional operational tests and inspections of the radioactive vaste system will be made as required to insure performance consistent with the require-ments of 10 C71t 20. Routine surveillance will be conducted f.r detection of system leaks. Padiation detectors and monitors vi u te periodically checked for calibration. Alarm circuits and autcmatic features of flow diversion j for vaste liquid and gaseous effluents will be periodically tested. i l Each component is inspected and cleaned prior to instaustion into the system. I Demineralized water is used to flush au portions of the system. Pre-operational tests include es11bration of instruments, testing of automatic
, ".mtrols, and verification of alarm set points. Au flow paths are checked for capacity and mechanical crerability. All pumps are run to demonstrate head and capacity. I l
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D-B ( between 548' and 560'. considered. The water available as heat sink below Sh8' was not The design parameters used in this analysis and the results are given below. DESIGH PARAMETERS: 6 g,1, Volume of water available below elevation 560' = 10 x 10 6
= 83 x 10 lbs.
Surface area at elevation 560' = 165,000 ft2 Time period considered = 1h days Average Decay Heat Rate (for ik day period) = h3 x 10 Etu/hr. Total Decay Heat = 14.5 x 109 Btu Weather data for Toledo (from Marley Handbook) Max. dry bulb temperature F = 93 Max. wet bulb temperature F = 77 Latitude = hl*-36' N
, Wind Velocity . = 3 mph 3
( Max. expected heat gain by solar radiation (2h hr. ave m e) = 130 Btu /hr-ft 2 Minimum time after reactor shutdown the service = 19 hrs. water pumps are required Service water temperature at the start of decay = 80 F (max.) heat removal system CALCULATED DATA FOR COOLING POND: f Heat Removal Capability Water loss b ? Pond Temp. F Btu /hr. evaporation-gs,* g Based on 160,000 ft2 surface area 160 160 x 106 330
*1h0 60 x 10 120 0
130 h2.3 x 10 83 120 40.0 x 10 78 .
- Expected equilibrium pond water temperature, since the total solar heat
< gain plus decay heat at 75 hours after shutdown approximately equals the
( removal capability. 1h-62A Amendment No. 3
D-B Time to raise the pond te=perature to = 75 hours m 140 F based on water elevation 560' and *
)
assuming pond does not lose any heat during this time (includes heat enin by solar radiation) Total water evaporated in 2h hours = 1 7 in, water depth , (for 140 F water temperature) Approximate time to evaporate water from = 20 days elevation 560' to 557'-6 (assuming no in-leakage) FORMULAS USED
- 1. W = moisture evaporated lbs/hr-ft 2 = (2h0 + 3 7) (P,-P -
7000 Where 3P = saturaticn pressure of vapor at temperature t F, inches of mercury P = actual vapor pressure of air, in. Hg t = average water temperature, F
- 2. Total water evaporated = W lb x A ft2 x 1 eal. x Hr 3 Hr-ft2 8.25 lb - 60 min ,
x
= gal / min '
- 3. Heat removal rate = wha Btu /hr Where h = latent heat of water, Btu /lb A = surface area, ft 2
REFERENCES:
- 1. Cooling Tower Fundamentals and Application Principles, Published by The Marley Company, Kansas City, Missouri.
- 2. Perry's Chemical Engineers' Handbook; Perry, Chilton, Kirkpatrick, McGraw-Hill Publications, 4th Edition.
3 Kent's Mechanical Engineers' Handbook; J. Kenneth Salisbury, Wiley Handbook Series, 12th Edition. Amendment No. 3 14-62B
D-B 2.3 Describe the systems which vill assure that radioactive liquid vaste vill be diluted in circulating water discharge, or confirm that the radicactive liquid vaste released into the discharge canal vill not exceed the limits of 10 CFR 20 for unrestricted areas. Your description should include the folleving information:
- a. The instruments and equipment that will prevent the dis-charge of radioactive liquid vaste while circulating water is not being discharged.
- b. The minimum circulating water discharge rate (gpm) that will permit the discharge of radioactive liquid vaste.
- c. The maximum discharge rate (gpm) of radioactive liquid vaste.
- d. A drawing showing how the radioactive liquid vaste vill be mixed with circulating water discharge.
RESPONSE
Samples are taken and analyzed of all vaste before it is released from the station. If these show that the specific activity is sufficiently low, the liquids are passed from the radvaste system to the circulating water discharge canal. On the way out, they must pass at.least one radiation ,, monitor which controls two (2) downstream isolation valves in series. These valves are set to close if excessive radiation is detected. Further down-stream a third valve is located which is interlocked with the circulating water pumI(s). This ensures that dilution flow is provided before vastes are put int the canal. The minimum circulating water discharge rate vill be that provided by two (2) pumps or about 356,000 gpm. The maximum discharge rate for vaste vill be determined by either the- physical capacity of the system (1h0 gpm) or by the need for dilution to meet 10 CFR 20 limits. Processing through the radvaste system should induce all isotopes, except tritium to concentrations far below those allowed for release to unrestricted areas. For these, no dilution is needed. In the case of tritium, however, processing has no effect. Therefore, if it builds up to a point where credit for dilution is needed, the maximum discharge rate of the vaste vill be regulated accordingly. For prelisiaary calculational purposes, about 80% mixing is assumed in the canal for circulating water and vaste. A design for accomplishing this has not been developed yet. \ 2.3-1
D-B It is estimated that a maximum of about 150,000 ft3 of vaste are procesaed
; during a complete cycle. Normally, a large portion of this would be reused; but for this analysis, all of it will be assumed to be released from the station. With this assumption and the specific activities of the radionuclides in the processed vaste that have already been calculated the total isotopic releases to the lake are determined. These are also presented in Tatie 2.k-1.
Tritium is treated separately since, at present, there is no feasible way of removing it from the vaste. The only way of reducing its concentration is by dilution. This is done in the station by mixing the vaste with the circulating water before releasing it to the lake. The mixing ratio vill be controlled to ensure that the concentration in Lake Erie at the dis-charge canal vill be below the 10 CFR 20 li=it of 3 x 10-3 uc/ml. Based on the conservative assumption of 30% leakage through the fuel cladding, about 3000 curies of tritium vill be released per e> cle. .
- b. The following is a list of all major tanks that may contain radioactive liquids:
- 1. Pressurizer Quench Tank
- 2. Makeup Tank
- 3. Boric Acid Addition Tank (2)
- h. Borated Water Storage Tank 5 Primary Water Storage Tank
- 6. Reactor Coolant Drain Tank .
7 Clean Waste Receiver Tank (2)
- 8. Concentrate Storage Tank 9 Clean Waste Monitor Tank (2)
- 10. Miscellaneous Waste Drain Tank
- 11. Detergent Waste Drain Tank
- 12. Evaporator Storage Tank
- 13. Miscellaneous Liquid Waste Monitor Tank All of these, with the exceptions of the borated and pricary water storage tanks, are housed in Class I structures which will contain any liquid released by a rupture. The two exceptions are located in the open and are vulnerable to tornadoes and, in the case of the primary water storage tank, earthquakes. The only radionuclide of any significance in either of these should be the tritium which has been picked up through contact with the primary system. If a tritium level of 6 pc/ml is maintained in the primary system (@ 600 F) it is estimated that the upper limits in the borated and primary water storage tanks would be about h uc/ml and 9 pc/ml respectively. Assuming the possibility of only a single tank rupturing at one time, the vorst accident for this analysis would be that involving the primary water storage tank which has the higher activity level. In this case the contents of the tank are assumed to go down yard drains
-into the discharge canal and then to the lake. Taking no credit for dilution by the circulating water, this would result in a maximum accident tritium concentration in Lake Erie at the discharge canal of 9 pc/ml.
Applying the dilution factors derived in part (d) (see Table 2.h-2), it is evident that this level very quickly drops below 10 CFR 20 limits even 2 along the beach (i.e. , within 330 ft alcag the beach the level has
~'
dropped to less than 3 x 10-2 pc/ml). i
- 2.h-3
D-B
- c. The maximum average and instantaneous concentrations of radionuclides in m
Lake Erie at the discharge canal that could result from normal operations ") or an instantaneous rupture of a tank are shown in responses (a) and (b) above. The nearest public potable water intakes to the discharge canal serve Camp Perry, the Erie Industrial Park, and surrounding residences and are located approximately 2.8 miles (k.5 km) away. Table 2.k-2 shows that, at this distance, a dilution factor of about 6.3 x 10-9 can be expected with regards to station effluent. Applying this factor to any type of discharge that has been postulated gives radionuclide concentrations at the potable water intake that are negligible. Even for the case of the rupture of the primary water storage tank when it was assumed that vaste with up to 9 pc/ml of tritium could flow into thg lake, the concentration at the intakes would probably not exceed 6 x 10- pc/ml. This is many orders of magnitude below the permissible limits for release of tritium . to, unrestricted areas.
- d. The following derivations and calculations were done by Dr. C. K. Huang of the Scripps Institution of Oceanography.
Most mathematical models describing the distribution of conservative material in a plume emanating from a continuous fixed source in the atmosphere or ocean are based on the scsumptions that the turbulent field is homogeneous and stationary. The theoretical steady-plume models are deduced from the super-position of an infinite number of patch distributions in the presence of a mean current. If the flow field has a detectable mean ,,) velocity the diffusion in the direction of the current can be ignored. i./ Furthermore, if the material distribution within any individual disk-element in the plume is assumed Gaussian, which is in general approximately the case, then the concentration at any point in a plume is estimated by Gifford's (1959) two-dimensional model. In the lake, the mean concentration at any point downstream from the continuous point source is given by E (x* y' z) = ~
*P - +
- n(<y'#,* }"* O i, ts7 1r " s (y) where x, y, z are coordinates, x is in the direction of mean current, y is horizontal and perpendicular to the current direction, z is vertical; Q is the steady rate of discharge of conservative material from a point source in units /sec; d#b are the coordinate variences of the material distribution in cm 2; U is the mean current speed in em/sec. Note that the above diffusion model is anisotropic.
~ The peak concentration on the surface of the lake is q ' _.C,,x(x) = y .2
- sg (2)
In a stationary homogeneous turbulent f;
- id, after a long period of time the diffusivity is considered to approe.a asymptotically a constant.
\
Csanady (1964) and Okubo and Farlow (1967) studied the turbulent dit;usion j in the West Basin of Lake Erie and have shown the effective lateral eddy 2.k-h l l l
D-B
, k.2 Describe in more detail the analysis procedures that vill be used to determine that the nuclear steam supply system vill meet seismic Class I criteria (Setion 3.1.2.h, page 3-3 PSAR).
Include in this discussion the following: I
- a. A detailed description and sketch of the proposed math-
+ ematical model(s) of the system, including a discussion . of the degrees-of-freedom and methods of lumping masses and determining section properties,
- b. The mathematical model(s) to be used for the reactor vessel internals.
- c. A discussion of the analytical procedures to be used, including the methods of computing periods, mode shapes, design accelerations, displacements, shears and moments. .
- d. An explanation of which " actual earthquake records" are to be used in the time-history analyses and a comparison of the response spectra from these earthquakes and the
- spectra postulated for the site (PSAR Figures 111-5 and 111-6, pages 2C h7 and 2C h8).
- e. An explanation of how it will be determined that the LOCA and maximum earthquake time-histories are conser-vatively applied such that the maximum structural re-sponse is obtained. Are the LOCA and earthquake time-histories assumed to start at the same time? If so, vould it be possible to obtain greater response if the earthquake vere started at some increment of time, such as 10 seconds, either before or after the start of the LOCA?
. f. A listing of the damping values to be used.
RESPONSE
The answer to this question vill be deferred until Amendment No. h. e k.2-1
h: D-B'
. 4 4 N i
( ' 57- With regard to the shield-building and connected penetratis T rooms, provide: va -
- a. The manner in whish negative pressure vill be assured in the shield building base slab area.
- b. The type of connection that is being used for the ,
[ penetration room / shield building connection. ! ! c. A discussion of specific lines penetrating the steel containment and shield building structures which will not be enclosed within the containment system complex. Describe the nrnner in which potential leakage through these lines is assessed and factored into dose estimates for accident conditions. .
RESPONSE
The answer to this question is'being deferred until Amendment No. h. f i - 1 1 I 4 d 4 4 4 i I f 5 7-1
, y , , r , - . , , , , . . , . _ . - ,m _ _-.. - --...,. -
D-B f' '5.15 Describe the containment building vacuum breaker system and s provide the following informatien:
- a. Criterion used to establish the system's capacity,
- b. A failure analysis of the system.
- c. Would the vacuum breaker system be adequate for a pipe
_" break in the shield building annulus or penetration
, rooms?
, RESPONSE Response to this question will be supplied in a subsequent Amendment. I J a f a l .i s i 1 i a i i i .I [ 5 15-1
. , , . . . , . . - . _ , . - . . . . _ . - - - . . .. . . . ~ . - . , - -
D-B j ; 5.16 Discuss the criteria for design of the pressure vessel cavity i shield valls. Include the following information:
- a. Discuss design provisions that vill be made to reflood
, the- reactor pressure vessel cavity following a loss-of-coolant accident,
- b. Indicate any changes in design which result from the pressure vessel being supported from the inlet and outlet nozzles.
- c. Describe the maximum capability of the cavity to withstand a pressure transient and indicate largest rupture of primary system considered. Include a discussion of the loc. ding conditions considered and the a sociated stress levels in the steel and cenerete. -
- c. Discuss the heat removal requirements for the cavity shield vall.
RESPONSE
Response to this question is being deferred at this time because of lack of complete design development. The answer will be submitted with Amendment No. h. ~ ' ( 5.16-1
D-B _. 6.2 With regard to the calc 21ations of the sump coolant temperature following a loss-of-coolant accident, state the time after shut-down when (1) the heat removal capability of one decay heat ex-changer equals the _ reactor decay heat and (2) the capability of the two heat exchangere equals the decay heat.
RESPONSE
This answer is being deferred at this time pending re-run of the "COPATTA" program. The original program did not extend far enough in time after LOCA to answer this question. This answer will be submitted with Amendment No. h. 6
-t 6.2-1 i
. . ._ . - . . - .- ~. . -. - .- -
D-B
,- 7.18 The neutron detectors which are part of the F' ant Protection 1
~
System are temperature limited at 175 F. Provide the design ..'! criteria for detector cooling and temperature monitoring which allows the detectors to meet the requirements of IEEE 279, Paragraph 3(g).
- RESPONSE The answer to this question is being deferred until Amendment No. h.
e l l i 1 1
. J 4.
%%/
T.18-1
D-B 9.0 Auxiliary and Emergency Systems C:- 9.1 On page 9-1 of the PSAR, you have provided a listing of the fifteen codes and standards used in the design, fabrication, and testing of components, valves, and piping of the auxiliary and emergency systems. Provide a cross-reference identifying the components to which the various codes and standards apply.
RESPONSE
Response to this question will be supplied in a subsequeni Amendment. f 1 d
't 9 1-1
D-B 12.0 Safety Analyses 12.1 Reactivity Transients 12.1.1 We note that your design does not incorporate a pressurizer level trip. Provide the bases for your conclusion that this trip is not required. State the pressurizer level assumed in your analyses of the startup accident and the accident resulting from rod withdrawal at rated power. State the mass of steam dis-charged from the secondary system during these transients. In view of the fact that no pressurizer level trip is provided, give assurance that the integrity of the primary system would not be jeepardized if these transients were to occur with the pressurizer full. Indicate the discharge capability of the safety valves for , releasing liquid water.
RESPONSE
The response to this question vill be submitted in Amendment No. h. ( (- 12.1.1-1
D-B { 12.3 7 To illustrate the safety margin which exists due to the inherent design of the facility, identify the largest missile originating fran the turbine which would not penetrate the containment,
- control room, or spent fuel pool. To indicate the sensitivity ,
of the analysis to your assumptions regarding energy absorption by the turbine casing, present this information assuming casing energy absorptions of 0, 25, 50 and 100 percent of your best estimate ~ of absorption.
RESPONSE
The response to this question is deferred until a later date because of additional study work required on this subject. e e t I 12.3 7 D-B [s 12.6.h To provide a better understanding of the sequence of events in your analysis of the design basis accident; provide a chronology indicating the time after the pipe rupture occura that the following events occur: (1) core ficoding tanks start injection,
" (2) reactor building pressure reaches peak pressure, (3) primary system blowdown is completed, (h) core flooding tanks empty, (5) emergency injection starts, (6) containment spray starts, (7) air cooling units start, (8) recirculation mode starts, and (9) containment returns to atmospheric pressure, assuming (a) two air cooling units and one spray pump operate for containment heat removal, and (b) three air cooling units and two spray pumps operate.
RESPONSE
This answer is being deferred at this time pending a new and revised "COPATTA" run. The original COPATTA run was made with only the minimum safety feature systems operating to give the most conservative answer. This answer will be submitted with Amendment No. h. k 12.6.h-1
D-B 12.6.5 Provide a table indicating energy distribution prior to and
\ following the loss-of-coolant accident, the amount of energy
( generated and absorbed from the time of pipe rupture to the time of peak pressure, and the distribution of energy at the time of peak pressure. This listing should include at least the following for the break sizes indicated in the PSAR: (1) primary coolant internal energy, (2) core flood tank internal energy. (3) energy initially stored in the core, (4) energy stored in core internals, (5) energy stored in reactor vessel metal, (6) energy generate.d during shutdown and decay heat, (7) energy transferred to the steam. generators, (8) energy stored in piping, valves, and pumps, (9) energy in steam generator metal, (10) secondary coolant internal energy, (11) energy content of water vapor in the reactor building, (12) energy content of air in the reactor building, (13) energy content of water in the reactor building, (14) energy content of
- steel structures, (15) energy content of concrete internal structures and the reactor building walls and dome, (16) energy removed by the air handling units, and (17) energy removed by l
the containment spray. l
RESPONSE
This answer is being deferred at this time pending a new and revised COPATTA run. This will be answered in Amendment No. h. l 12.6.5-1
D-B g 12.6.6 Provide plots of the temperature of the steam-air mixture within the reactor building vs. time and of the reactor building sump water temperature vs. time assuming: (1) reactor building heat removal assuming (a) two air cooling units and one spray pump operate and (b) three air cooling units and two spray pumps operate. (2) emergency core cooling assuming (a) one independent pumped injection train operates and (b) the two independent pumped injection trains operate.
RESPONSE
This answer is being deferred at this time pending a new and revised . COPATTA run. This answer will be submitted with Amendment No. h. e f i f 1 12.6.6-1
D-B
'~'s
^
12.6.7 In order that we may assess the ability of the air handling k units to function as proposed, the following information is requested:
- a. A preliminary design of the heat exchanger surfaces and fan assembly including configuration and dimensions.
- b. Heat transfer performance of the unit for the spectrum of accident conditions including flow rates, temperatures, pressures and compositions for both steam-air flow and cooling water.
- c. An outline of the analytical procedures used in designing the heat transfer ' surface and in determining its per-fomance and the basis for these analytical procedures.
Include fin efficiencies and the potential for liquid . binding of finned surfaces. BESPONSE
- a. The preliminary design of the containment vessel air-coolers is shown in Figure 12.6.7-1.
- b. Heat transfer performance data of the containment air-coolers is given in Table 12.6.7-1.
( TABLE 12.6.7-1 . PERFORMANCE DATA-CONTAINMENT COOLERS (Capacities are for single components) I. NorffAL OPERATION
- Cooling capacity 1.8 x 106 Btu /Hr
- Air-flow 117,000 cfm
- Fan BHP 152 HP Entering Air Temp. 120 F Containment Pressure 1b.7 psia Relative Hunidity 70%
Density of Containment Atmos. .062 #/ft3
- Cooling Water Flow Sh0 gpm Entering Water Temp. 85 F i
12.6.7-1
l D-B l II. POST-LOCA OPERATION
),-
l Cooling capacity 75 x 10 6 Btu /Hr Air-flow 58,000 cfm l Fan BHP h5 HP Entering Air Temp. 26h F l Containment Pressure 50.7 psia Relative Humidity 100". Density of Containment Atmos. 0.132 #/ft3 - I Cooling Water Flow 1600 gpm l Entering Water Temp. 85 F
- c. The fan-cooler units will in all respects be similar to the units that are described in Section 6.3 of the Duke Power Co.
FSAR. 1 i J I l-
- Values shown are approximate estimates. Final Design vill establish accurate figures.
l I l l 12.6.7-2
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D-B 12.6.8 Examine your piping systems to determine if any leakage paths A from the containment building can exist in lines normally filled with liquid which may be emptied during a loss-of-coolant accident. For example, following tne cessation of high pressure injection, two check valves block the leakage of gases from the open primary system to the vented borated water storage tank. Indicate if these valves will seat with a slight backpressure of air or if leakage vill occur. Examine other valves in po-tentially open lines to determine if all valves seat tightly once water is removed.
RESPONSE
As shown in Figure 6-1, Amendment No. 3 of the PSAR, there is one spring assisted and two regular check valves in series, in the safety injection line, through which the leaking gases will have to pass before reaching - the motor operated valve. The motor operated valve on the outlet of borated water storage tank is automatically closed (see Section 7 PSAR) upon low level signal which also initiates the recirculation mode. It . is not anticipated that any gases can leak through the spring assisted check. Spring assisted check valves will be used in any other system which may have potential possibilities of being emptied during a loss-of-coolant accident. l ( l-l l l t ( 12.6.8-1
D-B 12.6.9 Provide your analysis of the potential hydrogen evolution to k containment during the period follcwing the post-loss-of-coolant accident. Clearly state your assumptions. Describe your plans regarding use of combustible gas control measures, including consideration of measures which do not entail the intentional release of fission products from the containment to the atmosphere.
RESPONSE
There are four (h) sources of hydrogen inside containment vessel during accident conditions. These are as follows: I. Hydrogen released by radiolytic decomposition of water. II. Chemical hydrogen released by Zirconium - water reaction. III. Evolut'.on of dissolved hydrogen contained in the reactor coolant. IV. Hydrogen released by zine-boric acid reaction. A discussion of each of these sources and the magnitude of the hydrogen release is as follows: I. Radiolytic Hvdrogen i The calculations for the radiolytic hydrogen generation following the LOCA were based on the following assumptions:
- a. The core has operated at 2772 MWt for an average irradiation period of 620 days.
- b. Hydrogen generation constant ("G" value) Core - 0.h5 molecules H2/100 eV absorbed energy.
- c. Contribution due to the noble gases in the containment is not significant by comparison to the other sources, therefore it is neglected.
Calculations show that the hydrogen generation rates vary from 0.5 cfm at 500 hours after the accident to 0.2 cfn at 2000 hours. Figure 12.6.9-1 gives a curve of total radiolytic hydrogen generation versus time after LOCA. II. Zr-Water Hydrogen The extent of zirconium-water reaction is controlled by the emergency core cooling system performance. The Davis-Besse ECCS system is designed to limit the hot spot fuel temperature to 2300 F or less during a LOCA, thereby limiting the zirconium-water reaction to less than 0.1 percent. However, for i conservatism one (1) percent zr-water reaction has been assumed. A one (1) percent reaction produces 3,320 cubic feet of hydrogen. 12.6.9-1
D-B III. Dissolved Hydrogen The reactor coolant system is operated normally containing 15 ko ec of dissolved hydrogen per liter of coolant. This hydrogen is partially released as the coolant is depressurized and cooled during the accident. The total amount of dissolved hydrogen contained in the coolant is only h72 sef, therefore, this is considered an insignificant source. IV. Zine-Borie Acid Hydrogen Tests at the Franklin Institute performed under LOCA conditions to study the hydrogen evolution due to chemical reaction between inorganic-zine paints and H B03 -N S 0 -NaOH solution have shown 3 223 , insignificant amounts of hydrogen generated. Since there are no chemical spray additives in the Davis-Besse centainment spray system, the hydrogen generation will be even smaller. As can be seen, the total hydrogen generated over a period of one week is approximately 56,000 cubic feet or about 2% of the containment free volume. It is recognized that a potential hydrogen problem may exist following a LOCA. Since there is no one tested and acceptable solution available, at this stage we are not in any position to specify a de, finite design of the hydrogen control measures for the Davis-Besse Station. We are studying the 5 following three methods as possible solutions. ,,j
- 1. A review of the contain=ent air recirculaticn system to provide best circulation in order to avoid any possibilities of creatien of hydrogen pockets.
- 2. Purging of containment atmosphere.
3 Catalytic recembiners. A definite design recommendation vill be made at a later stage.
.)
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D-B 12.7 Steam Line Failure ( 12.7.1 The analysis of the steam line failure accident in section 14.1.2.9 assumes the availability of offsite power. Provide an additional analysis assuming loss of offsite power coincident with the steam line failure. In addition, provide the following information:
- a. Describe the sequence of events throughout the accident assuming the condenser heat sink and primary coolant pumps are lost.
- b. Provide a discussion and an analysis to support your conclusion that a return to criticality vill not occur with a stuck rod. This analysis should consider the equilibrium core which will have no burnable poison rods. *
- c. In the analysis, discuss how the primary temperature and flow in each loop varies during the accident.
- d. On page lk-19 it. is indicated that operator action is required to switch the feedwater valve controller to manual to assure the startup and main control valves for the affected steam generat )r remain closed. Pro-vide a discussion of the time requirements for this
, operator action and evaluata potdntial carsequences
( if the feedvater valve controller is not twitched to manual.
RESPONSE
The answer to this question is deferred until Amendmant No. h. t 12.7 1-1 , l j
. i D-B 12.8.1 Provide the analysis to support your assumption that failure of a single steam generator tube vill not result in rupture of adjacent tubes.
RESPONSE
The results of tube leak tests are reported in Section 5, Page 33, of BAL10002 "Once-Through Steam Generator Research and Development Report", August 1969. h (e
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