ML20008D754
| ML20008D754 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 01/13/1969 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8007300624 | |
| Download: ML20008D754 (53) | |
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1 O f - i i I [! L l i l O i l i l 0010' O QY1-iB0 0 7 3 0 0 THIS DOCUMENT CONTAINS P00li 10ALITY PAGES t
y A TABLE OF CONTEIES Section Pye 1 EERODUCTION AITD
SUMMARY
1-1 1.1 UTIROIIICTION 1-1 1.2 DESIGN HIGRLIGHTS 1-1 1.2.1 SITE CHARAC ERISTICS 1-1 1.2.2 POWER LEVEL l-2 1.2 3 PEAK SPECIFIC P0k"T. IIVEL 1-2 l 1.2.h REACTOR BUILDING SYSTH4 1-2 1.2 5 ENGINEERED SAFEGUARDS 1-2 1.2.6 TLECTRICAL SYSTEMS AND 34ERGENCY POWER 1-h
.1.2 7 .ONCE-THRCUGH STEAM GENERAIORS 1-4
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) 1.2.8 SHARED C0!TONCTIS kTIII~0"dIER REACICR UNIT 1-h
'd l.2 9 PROCESS STEAM l-5 13 TABULAR CHARACTERISTICS 1-5
- 1. 4' PRINCIPAL ARCHITECTURAL AND ENGINEERING ~CHITERIATOR DESIGN 1-15 1.4.1 PIANT DESIGN 1-15 l 1.h.2 REACf0R 1-15 1.h.3 REACTOR C00IMTI AND AUTILIARY SYSTEMS 1-15 1.h.4 REACTOR ~ BUILDING 1-15 t 1.4.5 ENGINEERED SAFEGiJARDS 1-16 i
l.h.6 INSTRUMENTATION AND CCITfROL 1-16 f 1.h.7 ELECTRICAL SYSTE4S 1-16 l 1.4.8 RADICACTIVE WASTES 1-16 1 1.4 9 mm_rNG AND ACCESS CONTROL 1-16 l.4.10. FUEL HANDLING AND STORAGE l-17 1.4.11 PROCESS STEAM l-17 1-1
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i O U TABLE OF CONTriTS (Centd) Section Page 15 RESFARCH MID DEVELOP! CIT RETIRHEITS 1-17 151 XENON OSCILIATIONS 1-17 152 THEF14AL MiD hTDPAULIC FROGFAMS 1-17 153 FUEL RCD CLAD FAILURE 1-13 1 5.4 HIGH BURNUP FUEL TESTS 1-19 155 IIiTERNALS VDT VALVES 1-19 156 CCNTROL ROD DRIVE LDE TET 1-20 157 ONCE-TEROUGH STEAM GENERA'IOR TEST l-2C 158 SELF-PCWEPID DETECI'OR TESTS 1-21 159 3 LOWDOWN FORCES ON _ INTERNALS .1-21
, .O 1.6 CONSUMERS FCWER CCMPANY CCMPETENCE' TOM AND l OPERATE MIDLAND PIANT 1-21 17 IDENTIFICATION CF CONTRACTORS 1-22
1.8 CONCLUSION
S 1-22 P 003.CS 1-11 i _.
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I. l j ', LIST CF TABLES l l Table No. Title Page i I l-1 Egineered Safeguards 1-3 l 1-2 Cc=parisen of Design Parameters 1-9 i i I i I ( I r
004O'7 1-111
LIST OF FIGURES
- Figure No. Title 1-1 Site Plan 1-2 Equip =ent Location Reactor & Auxiliary Buildings Plan at Elev. 568'-0" 1-3 Equipment Location Reactor & Auxiliary Buildings Plan of Elev. 584'-0" 21 1-4 Equipment Location Reactor & Auxiliary Buildings Plan of Elev. 599'-0" 1-5 Equipment Location Reactor & Auxiliary Buildings Plan of Elev. 614'-0" 1-6 Equipment Location Reactor & Auxiliary Buildings Plan of Elev. 634'-6" 1-7 Equipment Location Reactor & Auxiliary Buildings Plan of Elev. 646'-0" 1-8 Station Arrange =ent 21 1-9 Equipment Location Reactor &. Auxiliary. Buildings p --Plan.of Elev. 659'-0"
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l-9a Equipment 1,ocation Reactor & A =414=ry Buildings 25 Plan of Elev. 673'-6" & 685'-0" 1-10 Equipment Location Auxiliary Building Section A-A
- 21 1-11 Equipment Location Reactor & Auxiliary Buildings Section B-B 1-12 Equipment Location Reactor Building Section C-C & D-D j 1-12a . Equipment Location Auxiliary Building Section E-E 1-12b Equipecut Location Auxiliary Building Section F-F 25 1-12c Equipment Location Auxiliary 3uilding
. Ear.tial P.Lans.&. Sections 1-13 Equipment Location Turbine Building - Unit 1 21 Plan of Elev. 614'-0"
.,. 1-14 Equipment Location Turbine Building - Unit 1 Plan of Elev. 634'-6" 00sCS . 1-iv Amerdnent No. 25 2/74
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i ' j LIST OF FIGURES (COSTD) I I Figure No. Title f f . 1-15 Equipment Location Turbine Building - Unit 1 i Plan of Elev. 659'-0" & 694'-0" i i i 1-16 Equip =ent Location Turbine Building - Unit 2 ! i Plan of Elev. 614'-0" i
- 1 1 F 21 1-17 Equipment Location Turbine Building - Unit 2 l
Plan of Elev. 634'-6" 1-18 Equipment Location Turbine Building - Unit 2 Plan of Elev. 659'-0" & 694'-0" i ! 25 1-19 Equipment Location Tisrbine Building - Unit 2 j Sections A-A, B-B, & C-C i
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;, 00109 1-iva Amendment No. 25 2/74 t-
l i > LIST C? AFFriDICFS ' t Iiu=ber Title 1A G1cssary cf Ter=s , i : l 13 Quality Assurance I-regra= l l
; 1C Principal Design Criteria !
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LIST OF FIG'JRES (At Rear of Appendices ) Fimire No. Title 1B-1 Consumers Power Co. Quality Acceptance - Field quality Acceptance Flow Chart 13-2 Consumers Pcwer Co. Quality Acceptance - Organization Chart 13-3 Project Quality Assurance Organization 13-h Quality Assurance Program i l l l i i l l 1 001i.1 l . I l-v1 Amendment No. 2 5/28/69
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This Preliminary Safety Analysis Report is sutritted in support of Ctnsurers Pcuer Ccepany's (CP Co) application for a constructica perri, and facility
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on the cap in the Metecrcicgy Appendix 2A, Figure 2A-1. Each of ,he tyc ceners ,ing units wil2 c.cerate initials./ at a core cover
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report is based upon the reference core design cf 2 i52 rit. It io expected
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.e . . 4. . . , ., n, structures engineered safeguards and certain h"pothetical 7 J accidents are evaluatG cr the expected ultimate core cutput of 2,552 1.'40.
Each nuclear stea: syste (USS) is a pressuriced water reactor type si:1mr to systens operating or under construction. It uses che ical shi and control rods for reactivity ccatrol and generates steam with a small a cunt of super-heat in ence-through stes: generaters. The NSS and initial fuel cores v1.. te supplied by Be Babecch & 'dilecx Cc pany (E4'4). e The plant censists of two units having a total cocbined capability of apprcxi-cately .,300 Ge and h,050, COO lb/hr of process stea=. 2e process steam will f be supplied to The Ecv Che 1 cal Ccapany (Dev) and the electricity will be sup-f plied to CP Co's system.
'V Construction is scheduled for cc=pletion of Unit 1 in time for icading fuel cc March 1, 1976, and Unit 2 one year later, and for turning the units over tc the Dispatcher by Septe ber, 1977 and 1978, respectively. To meet tnis sched-ule, constructicn activities will begin by July 1, 1969 ~he general arrange-cent or cajor equipment and structures, including the reactor, auxiliary and l turbine buildings, is shown en Figures 1-1 thrcush 1-18.
As the plant design progresses frc ecnceptual design tc fina; cetailed de-sign, the plant description and analyses will be subject tc change and refine-cent. This report presents descriptive caterial and analyses of a " reference design." Any significant changes to the criteria or designs wnich affect safety vill be prceptly crought to the attention of the AEC by suitable supplements. CP Co is fully respcnsible for the cceplete safety and acequacy of the plant and Cc:can.y cersonnel vill test > start up and operate the plant. Assistance in perforcing these functions will be rendered by Bechtel Corporation (Mchtel), Ea'J, and such cther consultants and suppliers as 2.ny be recuired. V f
- 1. c') T"'A TGi. .UT P _U a T T C.U.T. o' 1 . -O. .i dI"."" v C.U.< _0.AN."_*:"O.T S"u'T Co" O-~ The Midla.d Plant site is located along the south shcre of the Tittabavassee i River adjacent to Dov's =ain industrial ec: plex in Midland, Michigan.
c' 1.,. 00 3. W. A ene=ent no. 21
p Location of the plant i=ediately adjacent to Dov's cc= plex is essential to g facilitate the dud purpose concept. This site is characterized by an exclu-sicn distance (radius) of 1.26 miles to the north toward the city of Midland and a minimum distance of about 0 73 mile toward the other cardinal points; prcximity to the che=ical plant i=ediately adjacent to the north; freedc= frc= flooding; an abundant supply of cooling water; a flat terrain ecnducive to favorable vind speeds; a reliable network for e=ergency power; and favor- , able conditions of hydrology, seis=clogy, and geology. 2e site is in a seis=icly dor = ant regicn. A reliable heat sink is provided by the artificial cooling pond planned for the site. 1.2.2 PCWER E/EL
~ Initially lice.nsed power for each reactor core is proposed at 2,452 MWt. Core
. performance analyses in this repo-t are based en this initial pcVer level.
Operating confir=ation of reactor core parameters is expected to support an ulti= ate core power level of 2,552 MWt, and Midland Units 1 and 2 vill be de-signed to operate at this cutput. Postulated accidents that could release fission Troducts to the envircn=ent have been analyzed on the basis of 2,-552 MWt. An additional 16 MWt win be contributed to the cycle by the reacter coolant pumps. 1.2 3 PEAK SPECIFIC PcWER w/EL The peak specific power level in the fuel for initial operatica at 2,h52 MWt results in a maximum the:._al output of 16.83 kW per foot of fuel rod. This ]V}. value io ec=Inrable with other Teactors of this size and1herefore Tepresents no extrapolation of technology. This cc=parisen =ay be seen in the infor=a-tion presented in Table 1-2. 1.2. 4 - REAC'ICR WILDriG SYSTS4 The syste= designed to contain the marimum hypothetical accident (MEA) con-sists of the reacter building envelope and the engineered safeguards. The prestressed, post-tensioned concrete reactor building is of essentially the same design as the reactor buildings for the Palisades Plant, the Turkey Point Plant, the Point Beach Plant, the_0conee Nuclear Station, and he Rus-sellville Nuclear Unit. Several of the engineered safeguards are si=ilar to these plants and the Midland Plant presents no unce==on solutiens to the engi-neering problems or extrapolations of the present technology. 1.2 5 ENGDir.z.tud) SAFEGUARDS Engineered safeguards are employed to reduce the pctential Tadiation dose to the general public frc= the MHA (lk.2.2.h) below the guideline values of 10 CPR 100. The potential dose is reduced by i=ediate, autc=atic isolation of all Teact;.tr building Tluid Tenetraticus that 'are vt Tequired for 11:1 ting the consequences of the accident, thereby eliminating potential leakage paths. Long-term pctential releases following the accident are =inimized by rapidly reducing the reactor but M1ng pressure to near-at=cspheric within 24 hcurs, j j thereby reducing the driving pctential for fissicn product escape, and by j./ providing for the re= oval of -adicactive iodine frc= the reactor building atmosphere. 1-2 A=end ent No. 2 00i13 5/28/69
O V In addition, the engineered safeguards will prevent core meltdown should the
\=/ worst postulated loss-of-coolant accident occur. This is acco=plished by a large capacity injection, e=ergency core ecoling syste=, parts of which are continuously operated for normal purposes and are i==ediately available for energency duty. These systems, coupled with the ther=al, hydraulic, and blow-devn characteristics of this reactor, vill reliably prevent =etal-water reac-
, ions _and core melting (or core disfiguration into a gecretry which could prevent _further ecoling).
~ Equipment for the engineered safeguards of each nuclear unit, together with its nor=al operation codes, is as follows:
- a. High-pressure injection - nor= ally operates as part of the =akeup and purification system.
- b. Core flooding system - self-operating when e=ergency conditions require its use. No-external signal tr p ver source required for operation.
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- c. Low-pressure injection - nor= ally operates for shutdown cooling as part of the decay heat re= oval system.
- d. Reactor building spray system - normally shut down.
r _i e. ~ Reactor buflding cooling system - normally operating, except at
\h-- - reduced service water flow.
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- f. Reactor _ building isolation system - operates on test or accident signal.
Table 1-1 lists equip =ent supplied for the engineered safeguards. Table 1-1 Engineered Safecuards (For Each Nuclear Unit) Function Total Equipeent Installed
~
High-Pressure Injection- 3 pu=ps (Makeup) l 1 Storage Tank ,_ Core Flooding Syste= 2. Tanks ! Low-PPssure Injection 2 Pu=ps (Decay _ Heat Removal)
- 2 Heat Exchangers
- Reactor Building Spray .2 Pu=ps
] Syste= 2 Che=ical Additive Tanks
, ( N_ / Reactor Building Cooling 4 Air Recirculating and Syste= Cooling Units V 00:E. 8s./1 13 Amendment _No. 5
1.2.6 ELECTRICAL SYSTEMS AND DERGENCY PCWER a
; The Midland Plant Units 1 and 2 are part of the CP Co system. CP Co and The j Detroit Edison Company operate as a completely integrated electric power pool.
This pool is, at present, interconnected outside Michigan with the Hydro-Electric Power Commission of Ontario and will become interconnected to utili-ties in Ohio and Indiana in 1969 The Midland Plant units have the following redundant sources of electrical power:
- a. One unit will continue to supply cuxiliary power in the unlikely event of a trip separation from the tranccission system,
- b. The five (5) 345 kV lines which terminate at the Midland Plant switchyard provide start-up and standby power through the 3h5-138 kV step-down substation.
- c. Each of two emergency diesel generators supplies a separate emer-gency switchgear bus for each unit.
- d. The station battery is sized to provide a safe and orderly shutdown of each unit in the event that all a-c power is 1 cst.
These normal, standby, and emergency sources of auxiliary electrical power ,_ assure a safe and orderly-chutdown of-the -two-unit 71 ant. 'They-also assure
\. V the ability to maintain a safe shutdown conditica under all credible circum-stances.
1.2 7 ONCE-TERCUGH STEAM GENERATORS i The design of the steam generators lo based on extensive research, development, and experimental work on boiling heat transfer performed by B&W over the past 11 years. Each~ generator is a vertical shell-and-tube, counterflow heat ex-changer with reactor coolant on the tube side and steam on the shell side. Feedwater is pumped into the generator, heated to saturation by direct mixing with steam, converted to steam and superheated in a single pass through the generator. The basic design parameters, such as feed-wster heating, boiling length, superheat length, and perforrance characteristics, have been confirmed by testing a full-length, 7-tube unit and a 37-tube unit. Tests are continu-ing to provide additional data in these design areas for the 37-tube test unit. In addition, testing will continue with one 19-tube, full-length unit. With the once-through design, natural circulation flow is adequate to remove full decay teat without the use of reactor coolant pu:::ps. Even with total loss of pumps, no fuel rod will experience departure from nucleate boiling. 1.2.8 SHARED CCMPONENTS WITH Cf1'HER REACIOR UNIT l The _following systems and . facilities in nor=al operation will be shared between the two nuclear power units for the Midland Plant: 1 L 003.15 1-4
- . . - . _ - _ . . _ ~ , . _ . ._- , _ . _ . - - - -
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\'d a. Fuel Pool and Fuel Pool Cooling Syste
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- b. Service Water Syste:
; c. _3uilding Heating and Ventilating Systems
- d. Fire Protection Systen
- e. Process Water Makeup System
- f. Cooling Water-Pond
- g. Water Treatment Plant
- h. Auxiliary Building, Ad=inistration Building, Warehouss, Machine Shop, Control Room, and Laboratories i .
- 1. Radioactive Waste Treatment System 7 j. ~ Borated Wate Storage Tank
- k. Purification Demineraliser
- 1. Condensate Storage Tank
- m. (Deleted)
- n. .(releted)
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1.2.9 . PROCESS ST 27
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A feature of the Midland Plant design is the provision to furnish process steam as well as' electricity to.Dow, located adjacent to the nuclear plant site. The steam in normal plant operation is furnished as follows:
- a. 400,000 lb/hr at 675 psia '
- b. 13,650,000 lb/hr at 197 psia Two lines vill; transport the 197 psia steam to the site boundary. A single additional header transports the 675 psia steam to the site boundary. The radioactivity content of the steam co= plies with the limits set forth in 10 CFR, Part.20.
13 TAEULAR CHARACTERISTICS' Table.1-2 is -a co=parative ' list of important design and operating character-istics of the Midland Units.1 and 2, Rancho Seco Unit 1.(Sacramento Municipal Utility District),'Oconee Units-1, 2, and 3 (Duke Power Co pany), and Turkey Point Units 3 and E (Florida Pcwer and Light Company). The-design and oper-
- .ating parameters of the Rancho Seco, Oconee, and Turkey Point units are close
]Q to those of Midland-Units l'and 2. ' Rancho Seco and Oconee units each have-the
.sa=e rated core power as.the Midland. units,-and are near-duplicates in other
-g- ~f respects. The data in Table 1-2. represent infor:ation presented in available V station descriptions-and in Safety Analysis Reports submitted for licensing.
.003.16 i (1-5 Amendment No. 5
O The design of each of these stations is based en infer:atien developed frc (C/ opration of e-=~"al and prototype pressurized water reacters Over a number of years. The Midland unit design is based en this existing pcver reaeter technelegy and has not been extended beycnd the boundaries of kncun inferna-tien er operating experience. Tc.e similarities and differances of the features of -the reactor units listed in Table 1-2 are discussed in the folleving paragraphs. In each case, the ite: nu=ber used refers to the ite nu=bers used in the table. Ite: 1. ' Hydraulic and Ther al Desien parameters The rated pcVer of each Midland unit is the sa e as that for the Ccenee and Rancho Seco units. The slight variation in other parameters between the Mid-land units and the cther 3&*J units is due to the utilizaticn of canless fuel asse blies in place of the canned fuel asse:blies. The canless asse:bly allows a slightly larger fuel red which increases fuel leading and heat trans-fer m-face area. m - h tien cf the can vall results.in a sligh'.ly. lover pcver ' peaking factor and the =cre open lattice of the canless assembly in-creases coolant flew area. The reactor ecolant ficv rate, operating pressure, and operating ecolant temperature are the same for the Midland, Ocenae, and F.ancho Secc units. The conservatis= of design cf the Midland units is evi-denced by the DNER cf 171 ('4-3) at the overpower condition ec pared to essen-tially the sa=e value for the other B'M units and to a lever value for the other reactor presented. - y ( v) Ite= 2. Core Me*anical Design Parameters The table presents cos; arable technM cal design data for the canless fuel assembly for the Midland units, the canned fuel assembly for the Ocenee and Rancho Seco units, and the canless fuel assembly used for tha " ay ?cint units. The dimensions, materials, and technology for each of these reactors are similar. Differences between the W4 units and the Turkey Point units are related to differences in pcVer levels. The small differences in fuel red dimensions between the Midland units and the other B?ll reactors result frc= the utilitation of larger fuel rods and a larger fuel red pitch to =atch fuel assembly pitch of tr.e canless and canned fuel assemblies. The nu=ber of fuel rods per core is unchanged. The lesser number of centrol m d asse=blies in the. Midland units cc= pared to the other M4 units results frc: a reduction in the require ents for inserted centrol rod assemblies for equilibriu= xenon and transient xenen centrol. Burnable poiscn md asse:blies described for the Midland units result frc the higher first cycle burnup shown in the Preli=inary Nuclear Iesign rata, Ite: 3 4 Ite: 3 Tralizinary nuclear Desirn rata The core size, nu=ber of Tuel asse blies, and number of fuel cds are the sa=e for all of the B&W units and differ frem the Turkey Point units prinarily due to the difference in pcVer level. Fuel enrich ents differ between the 3&W w units pri=arily due to the different fuel cycle burnup requirements. Enrich- .Y) %'v' - tent increase for the higher first cycle burnup cf the Midland units is par-tially offset by the reduced a: cunt of structural steel in the canless 1-6 QQj,G
j] asse=bly. The excess reactivity require =ents for each reactor also vary with s V fuel cycle burnup; the higher burnup of Midland units is reflected in the higher initial excess reactivity. The Midland units have fewer control rod assemblies than do the Oconee and Rancho Seco units and more control rods than the Turkey Point units. The reduction in the number of control rod asse=blies and control rod worth for the Midland units is due to less control rod inser-tion in the core during operation for ec=pensation cf equilibriu= and tran-sient xenon reactivity changes. The tovable centrol rod worth for shutdown is not changed. The utilization of burnable poison as e part of the control balance allows for a reduction of the soluble poisen concentration to obtain moderator coefficients within a desired range. The Doppler coefficient for all cases shown is negative over the core life. Ite 4. Principal Desien Para =eters of the Reactor Coolant Syste= Most of the features in this section are directly related to =aterial proper-ties and the amount of heat produced in the reactor core. Note that the B&W units are identical. The parameters are scaled in proportion to the power of the reactor. The major difference is the number of coolant loops required to recove the heat produced. For the B&W units, only two loops are required since once-through steam gen-erators are used instead of the U-tubes-in-shell design. The greater cooling capacity of these steam generators per=:its a reduction in the nu=ber of cool-ing loops for an equivalent amount of. heat re=oved. .-(,/f3 Item 5 Reactor Ccolant System - Code Recuire=ents The B&W units are identical. Code requirements for the shell side of the t eam generator conform to the ASME III Class A Specification. 'This is con-sidered to be a contribution to the safety of the vessel. It enhances the
' integrity'because of-the more stringent ASME III Class A design, material, and quality-control require =ents.
Ite: 6. Principal Design Parameters of the Reactor Vessel The B&W units are identical. These vessel designs are characterized by a thinner ther=al shield and a relatively larger shell diameter. The larger diameter provides for additional water between the edge of the core and the vessel which leads to additional neutron attenuation. Item 7 PrinciTal Design Features of the Steam Generators The. steam generators in the B&W units are the sa=e. They are basically dif-ferent from the Turkey Point units since they are a once-through design incor-porating an integral superheat section. Item 8. Principal Design Parameters of the Reactor Coolant Pumps The B&W' designs are the same. In each specific tabular para =eter the rela-
,m tive number or site is in 3rupudion to the total a= cunt of heat removed frem / \ \ )
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i l l w the core. The B&W reactor pumps have higher head and horsepower requirements than the Turkey Point units have for approxhately the same flow because of differences in system pressure drops. Ite= 9 Princi p l Design Pars =eters of the Reactor Ccolant Piping The B&W designs are the sa=e. They utilize carben steel clad with stainless steel. Ite: 10. Reactor Building Parameters All reactor buildings are basically of the same design and construction. The differences are physical dimensions, a= cunt of concrete shielding needed and design incident pressures, which are a direct result of plant layout, engi-neered safeguards, system capacities, and site locatien. The reactor building design and shielding offer satisfactory protection to the surrounding popula-tion in case of accident and during no:::a1 operation of the generating units. Ite= 11. Engineered Safeguards
-Engineered safeguards are gene. W si=ilar.
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( m ,'< i ( d rJ N j! Table 1-2 Comparison of_Dee'a1 Parameters (per station unit easts unless noted) Midland Plant Hancho Seco Oconee Nuclear Station Turkey Point item Unit 1 or 2 Unit 1 Unit 1, 2, or 3 Unit 3 or b 1 Hydraulic and Thermal Desian Parameters Rated Heat Output (core), MWt 2,b52 ?,452 2.452 2,097 Rated Heat Output (core), stu/hr 8.369 a 106 t.369 m 10 6 8,369 x 10 6 7.157 a 10 6 N aimum Overpower, % lb 1. 14 12 System Pressure (nominal), psia 2,200 2,200 2,200 2,250 System Pressure (minimum steady state), psia 2,150 2,150 2,150 2,220 Power Distribut Sn Factorm Heat Generated on Fuel and Cladding, 3 97.3 97.3 97.3 97.4 Ft.h (nuclear) 1.70 1.85 1.85 1.75 Fq (nuclear) 3.03 3.15 1.15 3.12 ht Channel Factors . Fq (nuc. and meth.) 3.12 3.24 3.24 3.25 DNB Ratio at Rated Conditions 2,21 (W-3) 2.27 (W-3) 2.27 (W-3) 1.85 (W-3) 1.60 (BAW-168) 1.60 (BAWal68) Minimum DNS Ratio at Design overpower 1.71 (W-3I 1.73 (W-3) 1.73 (W-3) 1.30 (W-3) 1.38 (BAW-168) 1.38 (BAW-168) ' Coolant Flow Total Flow Rate, Ib/hr 131. 3 a 106 13g,.3 , 306 g3g,3 , go6 100.6 a 106 Effective Flow Rate for Heat Transrat. Ib/hr 124.2 a 106 gyo,9 , 306 120.9 x 106 93,5 go6 Effective Flow Area for Heat Transfet, it2 49.19 47.75 47.75 39.0 Average Velocity along Fuel Rods, it/sec ( y) Average Miss Velocity, Ib/hr-ft2 Coolant Temperature. F 15.2 2.52 a 106 15.70 2.53 m 106 15.70 2.53 x 10 D 13.9 2.35 m 10 Nominal Inlet 555 555 555 546.5 4 Naimum Inlet Jue to Instrumentation Error and Deadband 557 557 557 550.5 Average Rise in Vessel 47.8 47.8 47.8 54 Average Rise in Core 49.3 49.) 49.3 59 Average in Core $79.7 579.7 579.7 577 Average in Vesse! 578.9 578.9 578.9 574 Nominal outlet of Hot Channel 642.8 644.4 644.4 647 Average Film Coefficient, Btu /hr-ft2-F 5,000 5,000 5,000 5,500 Average Film Temperature Difference F 31 31 31 30 Heat Transfer at 100% Power Active Heat Transfer Surface Area, it 2 69,73h 48,578 48,578 42,460 Average Heat Flum, Btu /hr-ft2 163,725 167,620 167,620 164,200 N ximum Heat Flum, Stu/hr-ft2 510,300 543,000 543,000 533,600 Average Thermal output, kw/ft 5.h 5.4 5.4 5.3 C Naimum Thermal output, kw/ft 16.83 17.5 17.5 17.3 P Ptsalmum Clad Surf ace Temperature at
], Nominal Pressure, F 654 654 654 657 t . Fuel Central Temperature. F D- Nainum at 100% Power k,150 4,160 4,160 4,070 C'l N aimum at 184% overpower 4.650 4,400 4,400 4.210 The r m.s ! Out ptit , kw/ft at N aimum overpower 19.2 19.9 19.9 19.4 2 Core Mechanical Desian Parameters Fuel Assemblies Design CNA canless CRA can CRA can Rod Pitch, in. 0.568 RCC cantess 0.558 0.558 0.563
g x / l ) f I w/ y ,7 Q Table 1-2 (Cont 'd) e Midland Plant Hancho Seco Oconee Nuclear Station Turkey Point item Unit 1 or 2 Unit ! Unit 1, 2 or 3 Unit 3 or 4 Overall Dimensions, in. 8.522 x 8.522 8.522 s 8.522 8.522 s 8.522 8.426 s 8.426 . Fust Weight (as UOy), Ib 207,k86 201,520 201,520 179,000 1 Total Weight, Ib 268,155 283,200 283,200 226,200 Number of Crida per Assembly 8 8 8 8 Fuel Rods Number 36,816 36,816 36,816 32,028 Outside Diameter, In. 0. k 30 0.420 0.420 0.422 Diametral Cap, in, 0.007 0.006 0.006 0.0065 Clad Thickness, in. 0.0265 0.026 0.026 0.0243 Clad Material Zirealoy-b zgycatoy-4 Zircaloy-4 Zircaloy Fuel Pellets Ita t e rial UO2 sintered UO2 sintered UO2 sintered U02 sintered Density, t.of theoretica! 93.5 95 95 94-93 Diameter, in. 0. 370 0.362 0.362 0.3669 1.ength, in. 0.7 0.8 0.s 0.600 Control Rod Assemblies (CRA) Neutron Absorber $$ Cd-155 In-805 Ag 5% Cd-15% In-80% is 52 Cd-151 In-not Ag 51 Cd-15% In-80% Ag bladding Naterial 104 SS - cold worked 304 SS - cold worked 304 SS - cold worked 304 SS - cold wor *ed Dlad Thickness, in. 0.019 0.018 0.018 0.019 ilumber of desemblies 57 69 69 41 Number of Control Rods per Assembly 16 16 16 20 Burnable Poison Hod Assemblies (BPRA) 72 No No -- 7 Core Structure Core Barrel ID/0D, in. IhT/150 147/150 F' 147/150 133.5/137.25 O thermal Shield 1D/00, in. 155/159 155/159 155/159 34 3,of g4 7,5 3 Preliminary Nuclear Design Dats Structural Characteristics Fuel Weight (as UO y ), Ib 207.h86 201,520 201,520 179,000 Clad Weight, Ib L2,200 43,000 43,000 35,600 Cors Diameter, in. (equivalent) 128.9 128.9 128.9 119.5 Core Height, in. (active fuel) Ibb 144 144 144 Reflector Thickness and Composition i Top (water plus steel), in, ga 12 12 10 Bottom (water plus steel), in. 12 12 12 10 Side (water plus steel), in. 18 18 18 15 Id2 0/U (unit cell - cold) 2.85 2.97 2.97 3.48 Number of Fuel Assemblies 177 177 177 157 I C Fuel Rods / Fuel Assembly Performance Chardcteristics 208 208 208 204 Q i leading Technique 3 region 3 region 3 region i region m
'.. Fuel Discharge Burnup, HWD/HTU f'd Average First Cycle 13,*2L0 12.460 8,260 14,000 N Equilibrium Core Average 27.490 28,200 28,200 21,000 Feed Enrichments, w/o U-235 Nos. I and 3 Region i 2 . 30 2.29 2.24 2.28
, Region 2 2. 30 2.64 2.47 2.43 Region 3 2.64 2.90 2.77 2.73 Control Characteristics Effective Multiplication (beginning of itfe) Cold, Mn Power, Clean, No Burnable Poison 1.271 Nos. I and 1 No. 2 Hot, No Power, Clean, No burnable Poison 1.238 1.302 1. 312 1.255 1,275 Hot, Hsied Power, Xe Equilibrium, Bur m $le 1.247 1.258 1.201 1.225 l Poison 1.08h -- l i
V r n . \- Table 1-2 (Cont'd) Midland Plant Hancho Ecco Oconee Nuclear Station Turkey Point Item Lnit 1 or 2 Unit 1 Unit 1, 2 or 3 Unit. 3 or k Control Rod Assemblies . Material 55 cd-155 In-80% Ag 5% Cd-15 In-80% Ag 5% Cd-15% In-80% Ag 5% CJ-15% In-80% Ag . Number of Assemblies k9 69 69 41 Number of Absorber Rods per.CRA 16 - 16 lb 20 Total Rod Worth ,% 8.0 10.0 10.0 7.0 Baron Concentrations To Shut Reactor Down With Rods Inserted (clean), cold / hot ppm 810/L50 1290/1080 1290/1150' 2300/2500 Baron Worth (hot),% / ppm 1/100 1/100 1/100 1/130 Baron Worth (cold), % / ppm' 1/75 1/75 1/75 1/100 ICinetic Charackeristics Moderator Temperature Coefficient, /F 0 to -3.0 x 10~4 +1.0 x 10 to -3.0 x 10 +1.0 x 10 to -3,0 x 10 +).0 x 10 to -3.0 x 10 Moderator Pressure Coefticient, /psg + k.0 x 10-7 to + 3.0 x .g a-t- _g,g , 3g-6 ,, ,3,g , go -b _ ,g , 3g-b ,, 3,g , 3g 4 ,3,g , 3g-6 ,, 3,g , 3g 4 Hoderator Void Coefficient, /% void L.0 x 10 to -3.0 x 10'3 +l.0 x 10
- to -3,0 x 10 1.0 x 10' to -3.0 x 10'l +0.5 x 10'3 to -2.0 x 10' Doppler Coefficient, /F -1.1 x 10'D to-1.7 x 10'$ -1.1 x 10' to -1.7 x 10 -1.1 x 10 to -1.7 x 10 -1.0 x 10 to ~2.0 x 10' 4 Principal Deetan Parameters of the Reactor Coolant System I i bd System Heat Output, NWt 2,L68 2,468 2,468 2,097 h System Heat Output, Btu /hr 8,L23 m 106 3,423 x gg 6 8,423 x 10 7,15b x 10 6 Operating Pressure, psig 2,165 2,185 2,185 2,235 Reactor Inlet Tempe ra t u re , F 555 355 iS3 546.5 Reactor Outlet Temperature, F t,03 tO) i 003 600.6 Number of 1. oops 2 2 2 3 Design Pressure, psig 2,500 2,500 2,500 2,485
. Design Temperature, F 650 650 650 650 Hydrostatic Test Pressure (cold), psig 3.125 3.125 3,125 3.110 3
} Coolant Volume, including pressuriter, ft II .M 11,800 !!,800 9,800 ) Total Reactor Flow, gpm 352,000 352,000 352,000 2th,400 5 Reactor Coolant System Code bequirements Reactor Vessel ASME 111, Class A ASME !!!, Clan. A ASME 111, Class A ASME !!!, Class A Steam Generator h los Tube Side Shell Side ASME III, Class A ASI-li Ill, Class A ASME Ill, Class A ASME III, Class A ASME 111, Class A ASME Ill, Class A ASME Ill, ASME 111, Cl.ss Class A C i, ' y Pressurizer , ASME III, Class A ASME 111, Class A ASME 111 Class A ASME III, Class A ' Pressurizer Relief Tank ASME 111, Class C ASME III Class C ASME III, Class C ASME III, Class C
=
4 Pressurizer Safety Valves ASME Ill ASME !!! ASME Ill ASME 111 Reactor Coolant Piping USASI B31.1 USASI B31.1 USASI B31.1 USASI B31.1 Reactor Coolant Purp Casing ASE III, Cisss A ASME III, Class A ASME Ill, Class A 6 Principal Deslan Paramerers of the Reactor Vessel Materiel SA-533, Grade B, clad with SA-533, Grade b, clad with SA-5 33, Grade b, clad with SA- 302, Grade B, clad with a 18-8 Stainless Steet 18-8 Stainless Steel 18-8 Stainless Stect lype 304 austenitic SS
f% p m v) ( d (v) Table I-2 (Cont'J) Midland I'Innt Hencho Seco Oconee Nuclear Statto" l Turkey Point Item Unit i or 2 Unit 1 Unit 1, 2, or 3 Unit 3 or le Design Pressure, pelg 2,500 2,500 2,500 2,485 Design Temperatura, F 650 650 650 650 Operating Pressure, peig 2,185 2,185 2,185 2,235 InstJe Dianeler of Shell, in, 171 171 171 155.5 OutstJe Diametrr Across Nossles, In. 249 249 249 240/235-3/8 Overall Height of Vessel and Closure dead (over CBD naastes), it-in. 42-0 42-0 42-0 41-0 Minimus Clad TI.tckness, in. 1/8 1/8 1/8 5/32 7 Principal Desian Parameters of the Steam Generat rs Number of Units 2 2 2 3 Type Vertical, once-through Vertical, once-through Vertical, once-through Vertical, U-tube with inte-with integral super- with integral super- with integr al super - gral suis tur e separator, heater. heater. heater. Tube Material Inconal Incone! Inconel Inconel Shell Material Carbon steel Cart >on steel Carbon steel Carbon steel Tube Side Design Pressure, poig 2,500 2,500 2,500 2.485 Tube Side Design Temperature, F 650 650 650 650 Tube Side Design Flow, Ib/hr 65.66 s 10 6 65.66 m au" 65.66 a 10 6 11.5) a 10 Shell Side Design Pressure, psig 1,050 1,050 1.050 1,085 Shel! Side Design Temperature, F 600 600 600 600 Operating Pressure, Tube StJe, Nominal.psig 2,185 2,185 2,185 2,235 g Operating Pressure, Shell Side, M.salmum, e peig 910 110 910 1,005 g Maximus Molef ure at Outlet at Rated load, 1 15 F superheat tiyJenstatic Test Pressure (tut,e-side cold), 35 F superheat 35 F superheat 1/4 psig 3.125 3,125 1,125 1,110 8 Frincipal Deslan Parameters of the Reactor Coolant Pumps Number of Units 4 4 4 ) Type Vertical, single stage Vertical, single o. age Vertical, single stage Vertical, single stage. Radial flow with bottom suction and horizontal d i s c ha r ge . Design Pressure, reig 2,500 2,'300 2,500 2,485 C Design Tesperature, F 650 650 650 650 Q Operating Preset:re, Numinal, pelg Suction Temperature, F 2,185 555 2.185 555 2,185 555 2,235 546.5
-( Design Capacity, gpm 88,000 118,000 88,000 88,800 *b Design Total, Developed Hea.1, .t $10 170 310 256 h ityJrostatic Test Pressure (cold), psig .l.125 l 125 1,125 1,110 Motor type A-C Induct ion, sir gle A-C Induction, single A-C Induction, single A-C Induction, single speed speed speed specJ Hotor Rating (nominal), hp 9,000 9,000 9,000 5,500 9 Principal Design Parameters of the Reactor Coolant PipinA Material Carbon steel clad with SS Carbon steel clad with SS Carbon steel clad wit h SS Austentric SS Hot Leg (ID), in. 3h 36 36 29 Cold Leg (ID), in. 28 28 28 27 3/2
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Table 1-2 (Cont'd) Ite 4 Midland Plant 3 Rancho Ecco Unit I or 2 Oconee Nuclear Station Wrkey foint Unit 1 Unit 1, 2, or 3 Between rump and Steam Ceneraior (ID), in. Unit 3 or k 28 28
- 28 31 10 pesetor Fulldina Systen Parametete Type Steel-lined, prestressed, Steel-11aed, prestressed, Stest-lined, prestressed.
post-tensioned concrete, post-tensioned concrete, Steel-lined, prestressed, vertical cylinder with vertical cyttader with post-tensioned concrete, post-tensioned concrete, vertical cylinder with flat bottom and shallow flat bottom and shallow flat botts,e and shallow vertical cylinder with domed roof, flat bottom and shallow Design Pa? -dtsts domed roof. domed toef. Inside 1 4eter, it domed roof. 116 116 Height, ft 116 193 116 l Pree Volume, ft) # 658,0N P06 206 l 1,900,000 177 Reference incident kressure, pelg 1,900,000 t,550,000 59 Ref erence Incideht Energy (Eg), 8tu N'N** 306,700,000 59 56 Energy Required to Produce Incident 306,700,000 272,000,000 Pressure (E2 ), seu 326,000,000 Ratto: Eg/E2 3hl,906.000 341,806.000 0 941 300,000,000 j Ritto: (E2 - Eg /Eg 0.063 0.89T 0.e97 Concrete Thickness, f t 0.115 0.w? 0.115 0.103 Vertical limit 3-1/2 3-3/b Dome 3-3/4 3 3-1/% 3-1/2 Peactor Building I.eak Prevention taak-tight penetrations 3-1/4 3 anJ, Mitigation 1.eak-tight penetrations Leak-tight penetrations 7 and continuous steel and continuous steel and continuous steel Lesk-tight penetrations P liner. Automatic isola- liner. Automatic isota- and continuous steel 123 tion where required. tion where required. liner. Automatic isola- liner. Automatic isola-Caseous Effluent Pdrge tion where required. tion where required. Discharge vent line at Discherse vent above top top of Hesetor Pulld- Discharge vent above top Through particulate filters ing of Reactor Building of Reactor Building ( 200 ft shove grade) (*200 ft above grade) and monitors. Part of 11 Engineered Safeguards the main exhaust system. Q Safety Injection System Q No. of High Head Pumps 3 3 No. of 1.ow Head Pumps 2 3 3
. . Reactor Buildir.g Coolers 2 J
hv No. of Units k 2
}h Air Flow Cap'y. fach, at Accident b
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. Condition, cfm 21.000 b0,000
*orn Flooding System 54,000 80,000 No. of Tanks 2 g Tcrel Volume 6 ft) 2,820 2 2
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less-of-coclant accidents gLOCA) in which (a) the total energy contad-ad *~ the reactor cociant syste vater is assumed to be released into the reatter N building through a double-ended break of any one Of the pri ary ecolant pipes, (b) there is a simultanecus less of external electric pcVer, (c) heat is transferred frc: the reacter to the reactor building at:csphere by water supplied frc the d emergency core ecoling syste= (ECCS), (d) either the reactor building a - -=^4 culati~ a-d ecoling units function er the reacter building spray systen functions, and (e) the eng*-aa-a4 =afeguards including safety injection de nct c;erate until 25 to h0 ses;nds fcileving the accident. Selected penetrations are provided with either a seal vater injection systen or are continuously pressurdced with air at a pressure greater than building design pressure. Means are provided for pressure and leak rate testing of the reactor building syste including provisiens for leak rate testing cf piping and electrical penetrations that rely on gasketed seals er seed r ccepcunds. l . h .' 5 aul.xr -e.v SA7EGUARDS Engineered safeguards systems with redundant features are inecrporated in the plant design which, in conjunction with the reacter buildirs system, provide 1 high degree cf assurance that the release of fissicn procacts to the envi-ren=ent folleving any credible Icss-of-ccclant accident vill not exceed the reference deses set forth in 10 CTR, Part 100.
)
V 1.h.6 USTRUME! CATION A'iD COIEROL Inter 1ccks and autc=atic protective syste=s are provided along with ad inis-trative centrols to insure safe operation of the plant. A reacter protective syste= is provided to initiate reactor trip if the reacter approaches an operating 14-4t. An-engineered. safeguards actuation syste is Trevided to initiate these systers upon detection of lOCA. Sufficient redundancy is installed to ;e mit periodic testing of the reacter protective syste=s and so that failure or receval frc: service of any ene protective system cc=ponent er portion of the syste: vill not preclude -eactor trip or other safety action when required. l.h.7 11ECTRICAL SYSTD G Nor:al, standby and emergency scurces of auxiliar-y electrical power are pro-vided to assure safe and Orderly shutdevn of the plant and the ability to naintain a safe shutdevn ecndition under all credible circu= stances. 1.h.8 RADICACrrTE WASTES The radicactive vaste treat =ent syste is designed so that discharge of radic-activity to the environment is in acecr5ance with the requirements ;f '.0 CTR, Part 20. 1.h.9 SHIELDr:G AND ACCESS CONTROL m The plant is provided with a centraliced centrol rcce having adequate shield-ing to pemit occupancy during all credible accidN: situaticns. The 1-16 OO Q.7 Amendment No. 2 5/25/60
J radiation shielding in the plant, in ecmbination with plant radiation control (~))
\. procedures, insures that operating personnel do not receive radiation expc-sures in excess of the applicable limits of 10 GR, Part 20, during nornal operation and r".ntenance, 1.4.10 FUEL HA32 LING AND STORAGE Fuel handling and storage facilities are provided for the safe handling, s cr-age, and shiptent of fuel and will preclude accidental criticality.
1.4.11 PROCESS STEAM Process steam frem the plant will meet regulations as ts radicactivity con-tent, within the applicable limits of 10 GR, Part 20. 15 RESEARCH AND DEVELOPMENT REQUIREMENTS The research and develoiten+ progams that have been initiated to establish Tinal-design or to demonstrate the capability of the design for future opera-tion at a higher power level are su=narized as follows: 151 XENON OSCILIATIONS An analysis to evaluate the possibiJity of xenon oscillations throughout core life is under way. A modal analysis t detemine critical parameters has been completed, and the detailed spatial calculations are in progresa. If it is detemined that such oscillations may occur, appropriate design changes to
)
3 eliminate or control the oscillations will be incorporated. See also 3 2.2.2 3 l.S .2 ~ THERMAL ANDTIDRAULIC PROGRAMS 354 is conducting a continuous research and developtent program for heat transfer and fluid flow investigations applicable to the design of the Midland units. Two important aspects of this program are:
- a. Reactor Vessel Flev Distributien and Pressure Drop Tests A 1/6-scale .=odel of the vessel and . internals is under test to measure the flow distribution to the core, fluid mixing in the i vessel and core, and the distribution of pressure drop vithin the reactor vessel.
- b. Fuel Assembly Heat Transfer and Fluid Flow Test Oritical heat flux data have been obtained on single-channel ttibular and annular test sections with unifom and nenuniform heat fluxes, and en the multiple red fuel asse=blies with uni-form heat fluxes. These data have been obtained for a range of pressure, temperature, and mass velocities-enecmpassing the
"(
+s
) reactor design conditions. This verk is being extended to 00!'28 1-17
include cultiple rod fuel assemblies with nonuniform axial heat
.V generation. Additional mixing, flow distribution, and pressure-drop data vill be taken on models of various reactor flow calls and on partial fullocale fuel assemblies.
See also 3 3 2. 153 FUEL ROD CLAD FAILURE A study of clad failure mechanis=s associated with a loss-of-coolant accident is presently under way. This study has included identification of the poten-tial failure mechanis=s, a search of the literature to obtain applicable data, evaluation and application of existing data, and scoping tests to obtain data on potential failure me.chanis=s. The initial results of this study include the identification of the failure mechanisms, an evaluation of the infor=ation available in the literature concerning these mechanisms, and an evaluation of the effects of these mechanis=s on the reactor system design. The objective of-the study is to insure that there are no potential failure
=echanisms that might interfere with the ability of the emergency core cooling system to terminate the core te=perature transient and remove decay heat in tl.c went of a loss-of-coolant accident. These potential failure mechanisms include clad melting, zirconium-water reaction, eutectic for=ation between the Zircalcy-clad and the stainless steel spacer grids, the possibility of clad embrittlemert as a result of the quenching during core flooding, and clad per-foration or defor=ation accompanying its failure. In the case of clad melting s A)
( and circonium-water reaction, our present design limit for peak clad te=pera-ture precludes these as possible failure modes. Infer =ation available in the literature, along with experimental evidence from tests conducted by B&W, show that brittle fracture of the cladding will not occur as a result of q w iching following a loss-o#. coolant accident, and that eutectic for=ation between dis-si=ilar core =aterials will not interfere with the flow of emergency core coolant after the accident. B&W has undertaken a program to evaluate the effects of perforation and defor-
=ation of fuel rods during the te=perature transient following the loss-of-coolant accident. PreHMnary tests have been run on nine sa=ples of Zircaloy-4 cladding filled with ceramic pellets, and additic aal experiments are planneu to gain a clearer understanding of the effects of te=perature excursions on.Zircaloy-clad fuel elements. Current plans include perfor=ance of a three-phase program. In the first two phases which are experimental, single-rod excursions will be performed to better establish temperature-pressure relationships at the time of clad perforation. The single-rod tests of the first phase will also investigate the extent of defor=ation to be ex-pected under the varying conditions associated with simulated in-reactor te=perature excursions. 'These vi*ll include the effects of hydrogen concen-tration and oxide fi1=s. The second phase of .the progra= w.ill consist prin-cipally of multired tests to explore the effect of the restraining action of spacer grids and adjacent fuel rods and to deter =ine the randomization of the localized defor=ation in an assembly of fuel rods. In the third phase of the program, the data obtained from the two experimental phases wi.11 be applied g)
( to the analysis of the effects in a loss-of-coolant accident. v 0029 1-18
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9 (V) d. A valve assembly has been installed and removed re=ctely in a test stand to judge the adequacy of handling equipment,
- e. A valve asse=bly has been prototype tested ever an appropriate range of vibration frequencies and a=plitudes to verify the analytical results showing that the valve vill not unseat because of vibratien during nor=al operation.
See also 3 3.h. 1 5.6 cornoL RCD IRIVE LrrE TEST The test assembly for this program is a full-siced fuel assembly vith asso-ciated control rod and control red guide, adjacent internals, and centrol rod drive. The unit is being tested under conditiens of temperature, pressure, flov, and water che=istry specified for the full-sized reactor installation. 1This program wille=hrace_a prototype . phase in .which ..the imdt vill be sub-jected to cisalign=ent, . varying flow, and temperature. The second phase of this progra= is cne of life-testing where the unit vill be centinucusly cycled to cover the number of feet of travel and the number of trips anticipated for its life in the reactor. Both phases of the progra: vill confir: the opera-bility of the drive line in nor=al and =isaligned conditions, confirm the rod drop times and load-carrying characteristics of the actuator, indicate vibra-tion and fretting wear characteristics of the centrol rods and fuel asse=- fO 4 bliec, and da+-~da- the venr ^" -aa+--da+*aa ^ #' a'l the drive line ec=penente. Also, a ec=penent test program is being conducted using autoclave testing of selected cc=penents at reactor pressure and te=perature. The purpose of this . prcgram is to seek cut potential =aterial and/cr design proble=s prior to production unit testing. See also 3 3 3 1 and 3 3 3 4. 1 5.T ONCE-TEROUGH STEAM GEERATOR TEST Testing necessary to prove the adequacy of the ence-through steam generater design for service at the initial power level and to confir= the size and configuration of the units has been ec=pleted. Steady-state and lead-changing
- operations using ence-through stea= generator =cdels were perfor=ed to de=en-strate the ability of the unit to follev the transients and to de=cnstrate the interaction of the centrol syste= vith the water level, steam pressure, and
~
flows. The test equipnent consisted of one 37-tube full-length unit, ene 19-tube full-length unit, and a full-length T-tube unit. The tubes were fabri-cated in accordance with the production techniques anticipated for the
. full-sined unit.
The latter pcrtion of the progran included tests to determine the natural frequency of the tubes in the stea generator by subjecting them to artifi-cially induced vibrations from an external source. The buckling and vibration daracteristics verify the structural integrity of the tube design.
-(m \
t () Prinary and secondary blevdevn tests on the codels have de=cnstrated the in-tegrity of the unit under conditions of rapil depressurization and large tube-
-to-shell te=perature differentialt '~he results of these tests are being used in the development and verification of blevdo n analyses. .
007.d, f. 1-20
O 158 SELP-PCkHED DETECTG TESTS
\s,,_.-)
The test units for this proe m are the self-powered detectors described in 733 These units have been tested in the 3&4 Test Reacter at conditicus of te=perature and neutren flux anticipated in a central station reactor. These units are currently being tested in the Big Rcck Point Nuclear Plant where they are expceed to te=perature, neutron flux, and flew for ecnditions approxi-nating those in the Midland units. The results of these pregrams vill provide a detector syste= with predictable characteristics of perfernance and icngevity under in-ccre ecnditions. 159 EL0h'ICW PCRCES CN ETIERNAIS 3&*J has developed an analog ec=puter =odel to obtain detailed infor=ation en the forces i= posed en the reactor vessel internals curing the subeccled per-tien of blevdcun following a reactor ecolant syste rupture. The =cdel can be used to s1=ulate leaks at any location in the het er cold leg piping. Leak sizes up to a ec=plete shear of a nain ecolant pipe can be simulated. Resis-
~tance to flev between all regions is simulated, as vell as the inertia of the fluid in the connecting flev paths. Pressure in each region is calculated by using the equations for conservatica cf = ass and =c=entum and assu=ing an isentropic expansion of water in each region. The =inimu= pressure in each region is restricted to the saturation pressure corresponding to the tempera-ture in that region.
Test results have been obtained by the Phillips Petrolen: Cc=pany for the
-p/
('# blevdown of a vessel with and without s1=ulated reacter vessel internals. Additional blevdown testing has been conducted an:1 is still under way using the 1/h-scale LCFT vessel. Tests have been conducted with and without in-ternals in the vessel.
.The tests that have been cc=pleted, -together with those that are under way, vill provide an adequate a= cunt of test data to verify the B&'d analytical
=cdel.
1.6 CCNSUMERS P0kE CCMPA?TI CCMPETENCE TO MILD A?D OPERATE MIDIx e PIxtr Midland vill be the third nuclear plant to be built and operated by CP Co. In 1962 CP Co started operation of the Big Rock Point Nuclear Plant. The re-search and develop:ent progra= vas extended in Septe=ber 1967 for an indefi-nite period. The Big Rock Point Plant is centinuing a fuel develop:ent progra=. The generation at Big Rcck has been based upon fi:: power cen=it-
=ents for CP Co's electrical system.
In 1967 CF Co acquired a construction permit for its Palisades Plant near
. South. Haven, Michigan, en lake. Michigan. .The .reaetor at Palisades is of.a similar type to those which will be used at Midland. Palisades is expected to assu=e cen=ercial operation by May 1970. , OpcTators for the Midland Plant 1 rill receite experience at the Big Rock and (m)
Palisa:1es Plants. This training vill be supple =ented with cperating ic.struc-tion frc= the NSS supplier. 1 00132 i-
. -_ - .. . - - - - __. - - - _ . - - - - _ - _ _ ~ _ __ ._ _ _ . - _ _ . - _ _ _ _ _ _ _ _ _
4
, C? Co has had sufficient experience in the design and operation of nuclear
~, facilities to build and operate an additional nuclear generating plant.
i. 1.7 IDE:"TIFICATIO'I CF CO::TEACTORS
; CP Co is the sole applicant for the construction pe=it and crerating license , fer the :41dlend Plant. As owner c f the proposed facility, CP Co bears ulti-
- lte res.;cncibility for the design construction, and the safe operation of the plant.
i E&W has been selected to design and supply both nuclear cteam supply systens, core ficcding systems, feed-water centrol, reacter control and protection sycten and ciner related reactor auxiliary systems. E&W will alco provide instruction for plant dperators and administrators. Eechtel Corporation and its affiliate, Eechtel Company, have been employed to design and supply the balance of plant equipment, systems and structures. Eechtel vill perform the on-site construction of the entire plant with technical advice and concultation to be provided by E&W for the inctallation of the ::SJ. I 1 1.8 CO?iCLUSIONS 4 l On the basis of information presented in this Preliminary Safety Analysis i s Re po rt , the Midland Plant will be designed, constructed and operated without undue risk to the health and safety of the public. v ! l s.
. .m v (,;d. 5. 30 1-22 Amendment ?
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