ML20098C518
| ML20098C518 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/02/1995 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20098C501 | List: |
| References | |
| NUDOCS 9510100184 | |
| Download: ML20098C518 (15) | |
Text
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' LAR'95-0009 Page 12 INDEX DESIGN FEATURES SECTION PAGE 5.1 SITF _ Loc.47/dd 5-I
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5.3 REACTOR CORE Fuel Assemblies................................................
5-/ I Control Rods...................................................
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5.6 FUEL. STORAGE Criticality....................................................
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f DAVIS-BESSE. UNIT 1 XIV Amendment No. 38.170
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The Davis-Besse Nuclear Power Station, Unit 5.0 05 SIGN FEATURES Number 1, site is located on Lake Erie in Ottava County, Ohio, approximately six miles northeast F
1 from Oak Harbor, Ohio and 21 miles east from SITE b> Tech Toledo, Ohio.
The exclusion area boundary has a 5.1 minimum radius of 2400 feet from the center of 2
EXAL*SION AREk the p1 ant.
The a:: cluster. scea-!: :hr := higer: 5.1-1.
5.1.1
- LO'.! POPUL ATION 20ME t
5.1.2 The icw populatier. :;r.; i: :hsr. On Ff;;ra 5.1=2r 41TE-B00NOAAV-5.1.2 The UNRCSTRICTE: ARCA :34 5!TE SOL"".".Y fer r:di:::the liquid Offlu::::
i: :nr r. cr Figure 5.1-3.
The UNRCSTRIC'00 ARCA er,d SITE 000%0ARY for redie-
- tiv: g:::::: Offlu::t: 1: :h = : "f;;r: 5.!-b 5.2 CONTA!'NENT M GURATION
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5.2.
The containment structure is comprised of a steel containment vessel, having e shape of a right circular cylinder with a hemispherical dome and ellipisoda bottom, surrounded by a reinforced concrete shield building.
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5.2.1.1 CONTA ENT VESSEL f
a.
Nominal in de diameter = 130 feet.
b.
Nominal inside ight'= 285.5 feet.
0
- . Net free volume = 2.
4 x 10 cubic feet.
d.
Nominal thickness of vess walls - 1 1/2 inches.
e.
Nominal thickness of vessel d
- 13/16 inches.
f.
Nominal thickness of vessel botton 1 1/2 inches.
5.2.1.2 SHIELD BUILDING a.
Minimum annular space = 4.5 feet.
b.
Annulus nominal volume = 678.700 cubic feet.
l c.
Nominal outside height (measured from top of foun tica base to the top of the dome) = 279.5 feet.
d.
Nominal inside diameter = 139.25 feet.
e.
Cylinder wall minimum thickness = 2.5 feet.
f.
Dome minimum thickness = 2.0 feet.
g.
Dome inside radius = 125.29 feet.
Nehd OAVI5-BESSE, UNIT 1 5-1 Amendment No.170 e
f' LAR 95-0009 Page le t
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LAR 95-0009 Page 15 N i i L gk' , ft.P 4 ,;i,3.e y .V,T,+1".- ,,v { 1 u...- ? N':~.'.i i n 3 s' b l .i.{.3w Y f.. I -YS {s' f;d i r;1 2 /. e x .,7 m[.my 2.71 1 %p: Ye .k,3.. l i us u ct.,W,,,/ g r i S' 8 r-b p k,. ', h \\ l a< j '.4.t. l;m b.< N (i.Q... c. ... e q g i l z% '? ' ? 'f i' s,. I,. u* % l 2 l i .Q---- >'.iA2 - l _ T; s i DAVIS-BESSE NUCLEAR POWER STATON bl*k cI I j LOW POPULATON ZONE FIGURE 5.1-2 DAVIS-BESSE. UNIT 1 5-3 i 9
. $AR95-0009.- Page 16 DESIGN FEATURES - OES!C" "RESSURE ^_40 TEMPEP_a.TURE- .\\ i 25.2.2' The r:::ter centsi.nnt building is designed :nd shell b; =intained for :: e i=e= iate ~e' pressure Of 40 p:ig :nd : tsi.pr:tur ef 204*,. 5.3 REACTOR CORE FUEL ASSEMBLIES or EIE.LO. 5.3.1 'The reactor core shall contain 177 fuel assemblies. Each assembly shall consist of a matrix of zircaloy c a fuel Tods with.an initial composition l of natural or slightly enriched uranium dioxide (UO ) as fuel material. Limited- ,l 2 substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with'96AC-approved applications of fuel ro'd configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or_ analyses to comply with all fuel safety design bases'. A-limited number of lead. test assemblies that have not completed representative testing may be placed in non-limiting core regions. CONTROL RODS 5.3.2 The reactor core shall contain 53 safety and regulating control rod assemblies @ht:5 =y 5: Of th:r :t:nd:rd er ent=d:d lif: :::tr:1d este-*'ier) and 8 axial power shaping rod (APSR) assemblies. Stend rd centre! -ad: 55 !! centtia e na-ine! 134 i-d:: ef 2: Ort:r = tert:1. The extended 'ife centml md: 05:11 :: t:in : 7.=i=1 1% ind = sf e nrber e terial. The nominal values of absorber material for the safety and f regulating' control rods shall be 80 percent silver,15 percent indium, and 5 percent cadmium. Ste derd := tre! ed: = d ^.PS9: :h:115: :1:d with :t:in 10:: st::1 tdir;- The enteded lif: centr:1 red: :h:11 b; :1:d with Incenei. The ^PSR the!' :^ateia e n^ inal 53 inches of abe -ber.-teri:1 :t thei-1:= r ed:. The absorber material for the APSRs shall be 100 percent Inconel. 5.4 RE^.CTOR COOLa" SYSTE., De/cleci l b N PRESSURE AND TEMPERATURE j i 5.4.1 The re r coolant system is designed and shall be maintained: t e code requirements specified in Section 5.2 a. In accordance w l of the FSAR, with allogfor normal degradation pursuant to applicable Surveillance Requ nts. j i b. For a pressure of 7500 psig, and c. For a temperature of 650'F, except for the pressuriz d i [ _ pressurizer surge line which is 670*F., t!.eIe;{s.f 4 9 1 DAVIS-BESSE, UNIT 1 5-4 Amendment No. JJ.J9,723,J20,135,7664Y71, 179 4 4 v m
. -. ~ ~ LAR 95-0009 .Page 17 DESIGN FEATURES-1 VOLUME 5 A? he total water and : tete velure of the reacter :0:1:ht zy:te= 1: ::,;;0 200 cubic feet at : nemi :1 T, Of 525"c n 5.5 "ETE090 Loc! CAL 79urp tnea7;nu gg 5.5.! 'h: :t:Or:10gic:1 t:::r :h:l' be !cetted 2: ther en ef;ere 5.!-1. 5.6 FUEL STORAGE CRITICALITY S. lo.1 C 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with: a. A K,, equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 1% delta k/k for calculation uncertainty. b. A rectangular array of stainless steel cells spaced 12 31/32 inches on centers in one direction and 13 3/16 inches on centers in the other direction. Fuel assemblies stored in the spent fuel pool shall be placed in a stainless steel cell of 0.125 inches nominal thickness or in a failed fuel container. c. Fuel assemblies stored in the spent fuel pool in accordance with Technical Specification 3.9.13. 5.6.1.2 The new fuel storage racks are designed and shall be maintained with: equivalent to less than or equal to 0.05 when flooded with A K,, ted water, which includes a conservative allowance of 1% delta a. unbora k/k for uncertainties as described in Section 9.1 of the USAR. I b. A K,,, equivalent to less than or equal to 0.98 when immersed in a hydrogenous " mist" of such a density that provides optimum moderation (i.e., highest value of K.,,)for uncertainties as , which includes a conservative allowance of 1% delta k/k described in Section fi.1 of the USAR. c. A nominal 21 inch center-to-center distance between fuel assemblies placed in the storage racks. d. Fuel assemblies having a maximum initial enrichment of 5.0 weight percent uranium-235. DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below 9 feet above the top of the 4 fuel storage racks. 4 DAVIS-BESSE, UNIT 1 5-5 Amendment No. A9,13D,181 Correction Ltr dated 5/17/94 G I +
~ LAR 95-0009 Page 18 4 DESIGN FEATURES (' CAPACITY s-5.6.3 with a;The spent fuel storage pool is designed and shall be maintained storage capacity limited to no more than 735 fuel assemblies. C0e0NEurirvci te go venggerur,,ugy(gf g 5.7 S.7.' The ccepenent: id:ntified'i= Table 5.7-! tre det!gned and th:11-b:' maintained withda th: cyc!!c en tr:n:f eet 'f-it of Table 5.'-1. 1 I t 4 l DAVIS-BESSE, UNIT 1 5-6 Amendment No.13 l
w LAR 95-0009 Page 19 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS omponent or System Cycle or Transient Limit Design Cycle or Trans 1 1. Reactor Coolant 240 Heatup and Cooldown 70*F to 557*F System Cycles 557"F to 70*F 2. Rea tor Coolant 160 Step Load Reduction 100% to 8% RTP* Sy tem Cycles (Resulting from turbine trip) 4 3. Reactor olant 150 Step Load Reduction 100% to 8". RTP* System Cycles (Resulting from electrical load rejection) 4. Reactor Coolan 40 Reactor Trip Cycles Reactor Trip System (Resulting from loss of electric power to all RC pumps) 5. Reactor Coolant 10 Reactor Trip Cycles Reactor Trip System (Resulting from turbine rip without' automatic c trol action) 4 6. Reactor Coolant 40 React r Trip Cycles Reactor Trip System (Resul ing from rod withdra 1 accident) 7. Once Through 88 Reactor Tr Cycles Reactor Trip Steam Generator (Resulting m complete loss of all in feed-water) 8. Once Through 40 Reactor Trip Cycl Reactor Trip Steam Generator (Resulting from los of station power) f 9. Once Through 20 Reactor Trip Cycles Reactor Trip Steam Generator (Resulting from loss of feedwater to one steam generator)
- RATED THERMAL POWER fck cb DAVIS-BESSE, UNIT 1 5-7
LAR 95-0009 Page 20 TABLE 5.7-1 (Continued) Comoon tit or System Cycle or Transient Limit Design Cycle or Transien 10. One rough 10 Reactor Trip Cycles Reactor Trip Steam nerator (Resulting from stuck open turbine bypass valve) 11. Reactor Coolan 40 Rapid Depressurization 2200 psig to 300 psig System in one hour 12. Reactor Coolant Change of Flow Cycles Loss of one or more System RC pumps 13. Reactor Coolant 20 Hydro atic Test Pressurized to > System 3125 psig 14. Once Through 35 Hydrostatic sts Pressurized to > Steam Generator 1312.5 psig 15. Reactor Coolant 40 Test Transients High Pressure System Injection Test 16. Reactor Coolant 240 Test Transients Core Flooding Check System lve Test blebe L i i i 1 DAVIS-BESSE. UNIT 1 5-8 Amendment No.16 5
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- AY:3.EE!!E. UNIT 1 5.g heendment No.
170 l
i LRR 95-0009 . Pace 22 E ~~ m $5 5 a &l4 cl ws e en :T 4 ES E $G M7 n G* gt; WW 5& 1 2 r -l hE g m = I ~I E h b .v'.' ~, i a i ,e 5,.. j { 1 i e ,0
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==ri 1 E { 8 { 4n I Yr\\ l4 I if" t.- i . !i i i l I ? n t llF' 's i j i 4y4.......&:;f! 0, k j l,!g-N'h -mEgN-;/ l = b_ L n.i(- y L/1 i i 7,. v M l L i E a DAVIS-8 ESSE. WIT 1 5-10 Amendment"No. p,170 5
) Lpa 35-0009 Pegs 24, " LAR 95-0009 INDEX Page 23 DESIGN FEATtFES P&E SECTION 5-1 ..g_...__.-.--
- 511 Site Location M Deleted Preposed 5.3 Reactor Cqrg 5-1 Fuel Assemblies......................
5-1 Index Page IN Control Pixis... and Technical Specification 5.0 M Deleted Pages 5-1 and 5-2 M Deleted _.6 -Fuel Storace S 51 Criticality... 5-2 Drainage........................ Capacity.... 5-2 M Deleted 4 .._3.. 4 Amendment No. 28.I78 XIV DAVIS-BESSE. UNIT 1 March 9, 1992 'M'- w mewemm4
w.n-- - = - ~ ~. - - - - - k } 1AR 95-0009 Design Features ' Design Features 3 Page 25 5.0 5.0 5.0 DESIGN FEATURES 5 5.0 DESIGN FEATURES d 5.1 Site Location e age (cond nued The Davis-Besse Nuclear Power Station. Unit _Nurter 1. site is located on Lake Erie i 1 on t i on 1 6 in Ottawa County. Ohio, approximately six miles northeast from Oak Harbor. Ohio and h a s sha be placed f froe c er the 1n, in a stainless steel cell of 0.125 inches nominal thickness or in a failed fuel container. } 3.2 (Deleted)
- c. Fuel assemblies stored in the spent fuel pool in accordance with Technical Specification 3.9.13.
5.3 Reactor Core 5.6.1.2 The new fuel storage racks are designed and shall be 5.3.1 Fuel Assemblies maintained witk The reactor core shall contain 177 fuel assemblies. Each assembly shall
- a. AK equivalent to less than or equal to 0.95 Wien consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial flooded with unborated water. which includes a m
co@osition of natural or slighty enriched uranium dioxide (UD,) as fuel n material. Limited substitutions of zirconium alloy or stainless steel conservative allowance of 11 delta k/k for uncertainties filler rods for fuel rods. in accordance with approved applications of as described in Section 9.1 of the USAR. fuel rod configurations may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff b. AK equivalent to less than or equal to 0.98 when m approved codes and methods and shown by tests or analyses to comply with imersed in a flydrogenous " mist" of such a density that .-4 a u fuel safety design bases. A limited nuter of lead test assemblies provides optieuc moderation (i.e.. highest value of Km). '3 that have not coupleted representative testing may be placed in which includes a conservative allowance of 11 delta k/k i ^! non-limiting core regions. for uncertainties as described in Section 9.1 of the i N 5.3.2 Control Rods
- c. A nominal 21 inch center-to-center distance between fuel l
The reactor core shall contain 53 safety and regulating control rod assemblies and 8 axial power shaping rod (APSR) assemblies. The nominal assemblies placed in the storage racks. values of absorber material for the safety and regulating control rods shall be 80 percent silver.15 percent indha and 5 percent cadatum. The d. Fuel assemblies having a maximum initial enricnment of 'j absorber material for the APSRs shall be 100 percent inconel. 5.0 weight percent uranium-235. ld Ll 5.4 (Deleted) 5.6.2 Drainaoe !-i The spent fuel storage pool is designed and shall be maintained to l4 I 5 5 (Deleted) prevent inadvertent draining of the pool below 9 feet above the top of the fuel storage racks. 5.6 Fuel Storage 5.6.3 Caoacity 5.6.1 Criticality The spent fuel storage pool is designed and shall be maintained with 5.6.1.1 pent fuel storage racts are designed and shall be maintained a storage capacity limited to no more than 735 fuel assemblies.
- a. A 4, equivalent to less than or equal to 0.95 when flooded 5.7 (Deleted) witn unborated water, which includes a conservative allowance of II delta t/k for calculation uncertainty.
(continued) DAVIS-BESSE. UNIT 1 5-2 Amendment No. I6.38.86.I78. (6ttettf66 tit. difed 5/IU94 DAVIS-BESSE. UNIT 1 5-1 Ame%nent No. II.16.I9.I23.Z30 NttM 9.1992 I35.I66.I78.I73.I79.I81 7 ~ ' ' ' - - - - .. _ ~ _}}